Draft
Environmental Impact Statement
for a
Geologic Repository for the Disposal of
Spent Nuclear Fuel and High-Level
Radioactive Waste at Yucca Mountain,
Nye County, Nevada
Volume II
Appendixes A through L
U.S. Department of Energy
Office of Civilian Radioactive Waste Management
DOE/EIS-0250D
July 1999
From the collection of the
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ACRONYMS AND ABBREVIATIONS
To ensure a more reader-friendly document, the U.S. Department of Energy (DOE) limited the use of
acronyms and abbreviations in this environmental impact statement. In addition, acronyms and
abbreviations are defined the first time they are used in each chapter or appendix. The acronyms and
abbreviations used in the text of this document are listed below. Acronyms and abbreviations used in
tables and figures because of space limitations are listed in footnotes to the tables and figures.
BWR
CFR
DOE
EIS
EPF
FR
LCF
MTHM
NWPA
OCRWM
PM,o
PM25
PWR
UFSAR
use
boiling-water reactor
Code of Federal Regulations
U.S. Department of Energy (also called the Department)
environmental impact statement
energy partition factor
Federal Register
latent cancer fatality
metric tons of heavy metal
Nuclear Waste Policy Act, as amended
Office of Civilian Radioactive Waste Management
particulate matter with an aerodynamic diameter of 10 micrometers or less
particulate matter with an aerodynamic diameter of 2.5 micrometers or less
pressurized-water reactor
Updated Final Safety Analysis Report
United States Code
UNDERSTANDING SCIENTIFIC NOTATION
DOE has used scientific notation in this EIS to express numbers that are so large or so small that they can
be difficult to read or write. Scientific notation is based on the use of positive and negative powers of 10.
The number written in scientific notation is expressed as the product of a number between 1 and 10 and a
positive or negative power of 10. Examples include the following:
Positive Powers of 10
10' =x 1 == 10
10-= lOx 10= 100
and so on, therefore,
10"= 1,000,000 (or 1 million)
Negative Powers of 10
10"= 1/10 = 0.1
10"^= 1/100 = 0.01
and so on, therefore,
10"^ = 0.000001 (or 1 in 1 million)
Probability is expressed as a number between 0 and 1 (0 to 100 percent likelihood of the occurrence of an
event). The notation 3 x 10"^ can be read 0.000003, which means that there are three chances in
1,000,000 that the associated result (for example, a fatal cancer) will occur in the period covered by the
analysis.
^
Draft
Environmental Impact Statement
for a
Geologic Repository for the Disposal of
Spent Nuclear Fuel and High-Level
Radioactive Waste at Yucca Mountain,
Nye County, Nevada
Volume II
Appendixes A through L
U.S. Department of Energy
Office of Civilian Radioactive Waste Management
DOE/EIS-0250D
July 1999
) Printed on recycled paper with soy ink.
CONTENTS
Appendix
A Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and
Other Materials
B Federal Register Notices
C Interagency and Intergovernmental Interactions
D Distribution List
E Environmental Considerations for Alternative Design Concepts and Design Features for
the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
F Human Health Impacts Primer and Details for Estimating Health Impacts to Workers
from Yucca Mountain Repository Operations
G Air Quality
H Potential Repository Accident Scenarios: Analytical Methods and Results
I Environmental Consequences of Long-Term Repository Performance
J Transportation
K Long-Term Radiological Impact Analysis for the No-Action Alternative
L FloodplainAVetlands Assessment for the Proposed Yucca Mountain Geologic Repository
in
Appendix A
Inventory and Characteristics of
Spent Nuclear Fuel, High-level
Radioactive Waste, and Other
Materials
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
TABLE OF CONTENTS
Section Page
A.1 Introduction A-1
A.1.1 Inventory Data Summary A-2
A.1.1.1 Sources A-2
A.1.1.2 Present Storage and Generation Status A-6
A. 1.1.3 Final Waste Form A-6
A. 1.1.4 Waste Characteristics A-7
A.l. 1.4.1 Mass and Volume A-7
A.l. 1.4.2 Amount and Nature of Radioactivity A-7
A. 1.1.4.3 Chemical Composition A-9
A.l. 1.4.4 Thermal Output A-11
A.1. 1.4.5 Canister Data A-11
A.2 Materials A-12
A.2.1 Commercial Spent Nuclear Fuel A-12
A.2.1.1 Background A-12
A.2.1.2 Sources A-12
A.2.1.3 Present Status A-12
A.2.1.4 Final Spent Nuclear Fuel Form A-12
A.2.1.5 Spent Nuclear Fuel Characteristics A-14
A.2.1.5.1 Mass and Volume A-14
A.2.1.5. 2 Amount and Nature of Radioactivity A-16
A.2. 1.5.3 Chemical Composition A-16
A.2.1.5.4 Thermal Output A-19
A.2.1. 5.5 Physical Parameters A-21
A.2.2 DOE Spent Nuclear Fuel A-22
A.2.2.1 Background A-22
A.2.2.2 Sources A-22
A.2.2.3 Present Storage and Generation Status A-22
A.2.2.4 Final Spent Nuclear Fuel Form A-23
A.2.2.5 Spent Nuclear Fuel Characteristics A-25
A.2.2.5.1 Mass and Volume A-25
A.2.2.5.2 Amount and Nature of Radioactivity A-25
A.2.2.5. 3 Chemical Composition A-25
A.2.2.5.4 Thermal Output A-30
A.2.2.5.5 Quantity of Spent Nuclear Fuel Per Canister A-30
A.2.2.5.6 Spent Nuclear Fuel Canister Parameters A-30
A.2.3 High-Level Radioactive Waste A-30
A.2.3.1 Background A-33
A.2.3.2 Sources A-34
A.2.3.2.1 HanfordSite A-34
A.2.3.2.2 Idaho National Engineering and Environmental Laboratory A-34
A.2.3.2.3 Savannah River Site A-35
A.2.3.2.4 West Valley Demonstration Project A-35
A.2.3.3 Present Status A-35
A.2.3.3.1 HanfordSite A-35
A.2.3. 3.2 Idaho National Engineering and Environmental Laboratory A-35
A.2.3.3.3 Savannah River Site A-36
A.2.3. 3.4 West Valley Demonstration Project A-36
A.2.3.4 Final Waste Form A-36
A-iii
Inventory and Characteristics of Spent Nuclear Fuel, High-level Radioactive Waste, and Other Materials
Table Page
A-38 Parameters of nonstandard packages from Savannah River Site A-47
A-39 Parameters of nonstandard packages from West Valley Demonstration Project A-47
A-40 Estimated spent nuclear fuel quantities for disposition of 32 metric tons of
plutoniumin mixed-oxide fuel A-49
A-41 Assumed design parameters for typical mixed-oxide assembly A-49
A-42 Radionuclide activity for typical pressurized-water reactor spent mixed-oxide
assembly A-50
A-43 Radionuclide activity for high-bumup pressurized-water reactor spent mixed-oxide
assembly A-50
A-44 Elemental distribution of typical bum-up pressurized-water reactor spent mixed-
oxide assembly A-51
A-45 Elemental distribution of high bum-up pressurized-water reactor spent mixed-
oxide assembly A-51
A.A6 Mixed-oxide spent nuclear fuel thermal profile A-52
A^7 Number of canisters required for immobilized plutonium disposition A-53
A-48 Average total radioactivity of immobilized plutonium ceramic in a single canister
in 2010 A-54
A-49 Chemical composition of baseline ceramic immobilization form A-54
A-50 Thermal generation from immobilized plutonium ceramic in a single canister in
2010 A-55
A-51 Greater-Than-Class-C waste volume by generator source A-57
A-52 Commercial light-water reactor Greater-Than-Class-C waste radioactivity by
nuclide (projected to 2055) A-57
A-53 Sealed-source Greater-Than-Class-C waste radioactivity by nuclide (projected to
2035) A-58
A-54 Other generator Greater-Than-Class-C waste radioactivity by nuclide (projected to
2035) A-58
A-55 Typical chemical composition of Greater-Than-Class-C wastes A-58
A-56 Estimated Special-Performance-Assessment-Required low-level waste volume and
mass by generator source A-59
A-57 Hanford Special-Performance-Assessment-Required low-level waste radioactivity
by nuclide A-60
A-58 Idaho National Engineering and Environmental Laboratory (including Argonne
National Laboratory-West) Special-Performance-Assessment-Required low-level
waste radioactivity by nuclide A-60
A-59 Oak Ridge National Laboratory Special-Performance-Assessment-Required low-
level waste radioactivity by nuclide A-60
A-60 Radioactivity of naval Special-Performance-Assessment-Required waste A-61
A-61 Typical chemical composition of Special-Performance-Assessment-Required low-
level waste A-61
LIST OF FIGURES
Fieure Page
A-1 Locations of commercial and DOE sites and Yucca Mountain A-5
A-2 Proposed Action spent nuclear fuel and high-level radioactive waste inventory A-8
A-3 Inventory Module 2 volume A-9
A-4 Proposed Action radionuclide distribution by material type A-10
A-5 Thermal generation A-10
A-6 Typical thermal profiles over time A-21
A-vi
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
APPENDIX A. INVENTORY AND CHARACTERISTICS OF
SPENT NUCLEAR FUEL, HIGH-LEVEL RADIOACTIVE WASTE,
AND OTHER MATERIALS
A.1 Introduction
This appendix describes the inventory and characteristics of the spent nuclear fuel and high-level
radioactive waste that the U.S. Department of Energy (DOE) anticipates it would place in a monitored
geologic repository at Yucca Mountain. It includes information about other highly radioactive material
that DOE could dispose of in the proposed repository. It also provides information on the background
and sources of the material, present storage conditions, the final disposal forms, and the amounts and
characteristics of the material. The data provided in this appendix are the best available estimates of
projected inventories.
The Proposed Action inventory evaluated in this environmental impact statement (EIS) consists of 70,000
metric tons of heavy metal (MTHM), comprised of 63,000 MTHM of commercial spent nuclear fuel and
7,000 MTHM of DOE materials. The DOE materials consist of 2,333 MTHM of spent nuclear fuel and
8,315 canisters (4,667 MTHM) of solidified high-level radioactive waste. The inventory includes
approximately 50 metric tons (55 tons) of surplus weapons-usable plutonium as spent mixed-oxide fuel
and immobilized plutonium.
The Nuclear Waste Policy Act, as amended (also called the NWPA), prohibits the U.S. Nuclear
Regulatory Commission from approving the emplacement of more than 70,000 MTHM in the first
repository until a second repository is in operation [Section 1 14(d)]. However, in addition to the
Proposed Action, this EIS evaluates the cumulative impacts for two additional inventories (referred to as
Inventory Modules 1 and 2):
• • The Module 1 inventory consists of the Proposed Action inventory plus the remainder of the total
projected inventory of commercial spent nuclear fuel, high-level radioactive waste, and DOE spent
nuclear fuel. Emplacement of Inventory Module 1 wastes in the repository would raise the total
amount emplaced above 70,0(X) MTHM. As mentioned above, emplacement of more than 70,000
MTHM of spent nuclear fuel and high-level radioactive waste would require legislative action by
Congress unless a second licensed repository was in operation.
• Inventory Module 2 includes the Module 1 inventory plus the inventories of the candidate materials,
commercial Greater-Than-Class-C low-level radioactive waste and DOE Special-Performance-
Assessment-Required waste. There are several reasons to evaluate the potential for disposing of these
candidate materials in a monitored geologic repository in the near future. Because both materials
exceed Class C low-level radioactive limits for specific radionuclide concentrations as defined in
10 CFR Part 61, they are generally unsuitable for near-surface disposal. Also, the Nuclear Regulatory
Commission specifies in 10 CFR 61.55(a)(2)(iv) the disposal of Greater-Than-Class-C waste in a
repository unless the Commission approved disposal elsewhere. Further, during the scoping process
for this EIS, several commenters requested that DOE evaluate the disposal of other radioactive waste
types that might require isolation in a repository. Disposal of Greater-Than-Class-C and Special-
Performance-Assessment-Required wastes at the proposed Yucca Mountain Repository could require
a determination by the Nuclear Regulatory Commission that these wastes require permanent isolation.
In addition, the present 70,(XX)-MTHM limit on waste at the Yucca Mountain Repository could have
to be addressed either by legislation or by opening a second licensed repository.
A-1
Inventory and Characteristics of Spent Nuclear Fuel, High-level Radioactive Waste, and Other Materials
A.1.1 INVENTORY DATA SUMMARY
There are six general inventory categories, as follows:
• Commercial spent nuclear fuel
• DOE spent nuclear fuel
• High-level radioactive waste
• Surplus weapons-usable plutonium
• Commercial Greater-Than-Class-C waste
• DOE Special-Performance-Assessment-Required waste
This section summarizes the detailed inventory data in Section A.2. The data provide a basis for the
impact analysis in this EIS. Data are provided for the candidate materials included in the initial 70,000
MTHM for the Proposed Action and other inventory that is not currently proposed but might be
considered for repository disposal in the foreseeable future.
This summary provides general descriptive and historic information about each waste type, including the
following:
• Primary purpose and use of the data
• General comparison of the data between waste types
• Potential for change in inventory data
Table A-1 lists the inventory data that DOE used in the EIS analyses and their descriptions throughout the
document.
A.1.1 .1 Sources
Figure A-1 shows the locations of generators or sources of spent nuclear fuel and high-level radioactive
waste. Spent nuclear fuel is fuel that has been withdrawn from a nuclear reactor following irradiation.
The Proposed Action includes the disposal of 63,000 MTHM of commercial spent nuclear fuel in the
repository. More than 99 percent of the commercial spent nuclear fuel would come from commercial
nuclear reactor sites in 33 states (DOE 1995a, all). In addition, DOE manages an inventory of spent
nuclear fuel. The Proposed Action includes 2,333 MTHM of spent nuclear fiiel from four DOE locations:
the Savannah River Site in South Carolina, the Hanford Site in Washington, the Idaho National
Engineering and Environmental Laboratory, and Fort St. Vrain in Colorado.
High-level radioactive waste is the highly radioactive material resulting from the reprocessing or
treatment of spent nuclear fuel. The Proposed Action includes disposing of 8,315 canisters of high-level
radioactive waste in the repository. High-level radioactive waste is stored at the Savannah River Site, the
Hanford Site, the Idaho National Engineering and Environmental Laboratory, and the West Valley
Demonstration Project in New York.
The President has declared approximately 50 metric tons (55 tons) of plutonium to be surplus to national
security needs (DOE 1998a, page 1-1). This surplus weapons-usable plutonium includes purified
plutonium, nuclear weapons components, and plutonium residues. This inventory is included in the
Proposed Action, and the Department would dispose of it as either spent mixed oxide fuel from a
commercial nuclear reactor (that is, commercial spent nuclear fuel) or immobilized plutonium in a high-
level radioactive waste canister (that is, as high-level radioactive waste), or a combination of these two
inventory categories (DOE 1998a, page 1-3). Spent mixed-oxide fuel would come from one or more of
A-2
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-1. Use of Appendix A radioactivity inventory data in EIS chapters and appendixes (page 1 of 2).
Item'
Appendix A
EIS section
Number of commercial nuclear sites
Number of DOE sites
Mapped location of sites
Commercial SNF material
Commercial SNF dimensions
Commercial SNF cladding material
Percentage of conmiercial SNF with stainless-steel
cladding
MOX SNF part of commercial SNF Proposed Action
Number of sites with existing or planned ISFSIs
Amount of commercial SNF projected for each site
List of commercial SNF sites, state, operations period
DOE SNF storage locations
HLW includes immobilized Pu
HLW generators
HLW vitrification status
Weapons-usable Pu declared surplus
Two forms: MOX and immobilized Pu
Proposed Action inventory
Total projected inventory commercial SNF
Total projected inventory DOE SNF
Total projected inventory HLW
Total projected GTCC waste
Total projected SPAR waste
Table A-3
A.1.1
Figure A-1
A.2. 1.5.3
Table A-15
A.2.I.5.3
A.2. 1.5.3
A.2.4.5.1.1
Table A-4
Tables A-6 and A-7
Table A-3
Table A-17
A.2.4.5.2.1
A.2.3.2
A.2.3.4
A.2.4.1
A.2.4.1
A.1
Figure A-2
Figure A-2
Figure A-2
Table A-51
Table A-56
A.2.3.5.6
HLW canister dimensions
Thermal generation of 1 MTHM of commercial SNF at Table A- 14
time of emplacement
Commercial SNF, DOE SNF, and immobilized Pu A.2. 1 .5.2
contain fissile material A.2.2.5.2
A.2.4.5.2.2
Kr-85 (gas) is contained in fuel gap of commercial A.2. 1 .5 .2
SNF
Typical radionuclide inventory for commercial SNF Tables A-8 and A-9
1.1,2.2,2.2.2,2.4.1,2.4.2.3,
2.4.2.4,2.4.2.8,2.4.3,6.1,
7.0,7.2.1,7.3,1.1.3.1.1
1.1,2.2,2.2.2,2.4.1,2.4.2.3,
2.4.2.4,2.4.2.8,2.4.3,6.1,
7.0, 7.2.1, 7.3
Figure 1-1, Several Chapter 6,
7, App. J and K figures
1.1.1
1.1.1, Figure 1-3, H.2.1.4
1.1.2.1.1. 5.2.2, K.2.1.4.1
1.1.2.1.1.1.5.3, 5.2.2,5.5.1,
K.2.1.4.1
1.1.2.1.1
1.1.2.1.1
1.1.2.1.1, 6.1.1, K.2.1.6
Table 1-1
1.1.2.1.2, K.2.1.6
1.1.2.2
1.1.2.2, K.2.1.6
1.1.2.2
1.1.2.3
1.1.2.3
1.1.2.5,1.3.2,1.6.3.1,2.1,
Figure 2-3, 2.1.4, 2.2.2, 2.2.3,
5.1,5.2.2,5.6.3,6.1.1.1,7.0,
7.2, 8.1.2.1, Ll.3.1.1,
J.1.3.1.2, K.2.1.6
1.1.2.5,1.6.3.1,7.2,7.3,
8.1.2.1,1.1.3.1.1, K.2.1.6
1.1.2.5,1.6.3.1,6.1.1.1,7.2,
7.3, 8.1.2.1, J.1.3.1.2, K.2.1.6
1.1.2.5,1.6.3.1,7.2,7.3,
8.1.2.1, K.2.1.6
1.6.3.1,7.3,8.1.2.1,1.3.1.2.4,
J.1.3.1.3
1.6.3.1,7.3,8.1.2.1,1.3.1.2.4,
J.1.3.1.3
Figure 2-3
2.1.1.2
2.1.2.2.2
4.1,4.1.2.3.2
4.1.8.1, 6.1.3.2.1, H.2.1.4,
Table H-4, 1.3.1.1, 1.3.1.2.1,
J.1.5.2.1, K.2.1.6
A-3
Inventory and Characteristics of Spent Nuclear Fuel, High-level Radioactive Waste, and Other Materials
Table A-1. Use of Appendix A radioactivity inventory data in EIS chapters and appendixes (page 2 of 2).
Item" Appendix A EIS section
Amount of chromium per SNF assembly
Commercial SNF comprises at least 92% of
radioactivity in Proposed Action
DOE SNF has a variety of cladding
Commercial SNF has higher radionuclide content than
DOE SNF or HLW
Cs-137, actinide, and total curies contained in a rail
shipping cask for commercial SNF, HLW, DOE
SNF, and naval fuel
Radiological inventory of GTCC and SPAR waste
much less than commercial SNF or HLW
Average radionuclide inventory per package for SPAR
and GTCC waste
C-14 (gas) is contained in fuel gap of commercial SNF
Typical PWR bumup, initial enrichment, and average
cooling time
Typical BWR burnup, initial enrichment, and average
cooling time
N-reactor radionuclide inventory per canister is larger
than HLW radionuclide per canister.
21 PWR assemblies contain a higher radionuclide
content than 44 BWR assemblies
DOE would emplace twice as many PWR assemblies
as BWR
N-reactor fuel represents a large quantity of DOE SNF
Mass of N-reactor fuel per canister
Immobilized Pu disk dimensions
Number of immobilized Pu cans per HLW canister
DOE SNF radionuclide inventory
Assumed packaging method for GTCC and SPAR
Chemical makeup of waste inventory
MTU per assembly for PWR and BWR
Most HLW stored in underground vaults
A.2.1.5.3
5.1.2
A. 1.1. 4.2
5.2.2,5.2.3.3
A.2.2.5.3
5.2.2
Table A-2
. 6.1.2.1
Derived from Tables
A-8, A-27, and A-18
Derived from Tables
A-8, A-27, A-18, A-54,
and Section A.2.6.4
Derived from Table
A-54 and Section
A.2.6.4
Tables A-8 and A-9
A.2.1.5
A.2.1.5
Tables A-18 and A-27
Tables A-8 and A-9
A.2.1.5. 1
Table 6-2, Table J-17
8.2.7, 8.2.8, 8.4.1.1, F.3
8.3.1.1, Table 1-9
5.5,8.3.1.1,1.3.3,1.7
G.2.3.2,H.2.1.4,L1.4.2.5
G.2.3.2,H.2.1.4
H.2.1.1
H.2.1.1
H.2.1.1
Table A- 17
H.2.1.1
Table A- 17
H.2.1.1
A.2.4.5.2.1
L3
A.2.4.5.2.1
L3
Table A-18
1.3.1.1,1.3.1.2.1
A.2.5.4, A.2.6.4
L3. 1.2.4
Tables A- 12,
-13,
-19,
Table I- 10
-29,-30,-31,
-32,
-33,
and -34
Table A- 15
J.1.4.1.1
A.2.3.3
K.2. 1.5.2
Abbreviations: SNF = spent nuclear fuel; MOX = mixed oxide; ISFSI = independent spent fuel storage installation; HLW =
high-level radioactive waste; Pu = plutonium; GTCC = Greater-Than-Class-C; SPAR = Special-Performance-Assessment-
Required; MTHM = metric tons of heavy metal; Kr = krypton; Cs = cesium; PWR = pressurized-water reactor; BWR =
boiling-water reactor; MTU = metric tons of uranium.
the existing commercial reactor sites. Although the location of the plutonium immobilization facility has
not been decided, DOE (1998a, page 1-9) has identified the Savannah River Site as the preferred
alternative. For purposes of analysis, this EIS assumes that the high-level radioactive waste canisters,
which would contain immobilized plutonium and borosilicate glass, would come from the Savannah
River Site.
A-4
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
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A-5
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Greater-Than-Class-C waste is waste with concentrations of certain radionuclides that exceed the Class C
limits stated in 10 CFR Part 61, thereby making it unsuitable for near-surface disposal. Greater-Than-
Class-C waste is generated by a number of sources including commercial nuclear utilities, sealed
radioactive sources, and wastes from "other generators." These other generators include carbon- 14 users,
industrial research and development applications, fuel fabricators, university reactors, and others. These
wastes are currently stored at the commercial and DOE sites and exist in most states. They are included
in Inventory Module 2 of the EIS but are not part of the Proposed Action.
Special-Performance-Assessment-Required wastes are also Greater-Than-Class-C wastes managed by
DOE and are stored primarily at the Hanford Site, Idaho National Engineering and Environmental
Laboratory, West Valley Demonstration Project, and Oak Ridge National Laboratory in Tennessee.
These wastes are included in Inventory Module 2 of the EIS but are not part of the Proposed Action.
A.1.1.2 Present Storage and Generation Status
Commercial spent nuclear fuel is stored at reactor sites in either a spent fuel pool or in a dry storage
configuration generally referred to as an independent spent fuel storage installation. Through 1995,
approximately 32,000 MTHM of commercial spent nuclear fuel has been discharged from reactors (Heath
1998, Appendix C). DOE spent nuclear fuel is also stored either underwater in basins or in a dry storage
configuration similar to that used for commercial spent nuclear fuel.
As discussed in the next section, DOE would receive high-level radioactive waste at the repository in a
solidified form in stainless-steel canisters. Until shipment to the repository, the canisters would be stored
at the commercial and DOE sites. With the exception of the West Valley Demonstration Project, the
filled canisters would be stored in below-grade facilities. The West Valley canisters would be stored in
an above-ground shielded facility.
A.1.1.3 Final Waste Form
Other than drying or potential repackaging, processing is not necessary for commercial spent nuclear fuel.
Therefore, the final form would be spent nuclear fuel either as bare intact assemblies or in sealed
canisters. Bare intact fuel assemblies are those that do not have any disruption of their cladding and could
be shipped to the repository in an approved shipping container for repackaging in a waste package in the
Waste Handling Building. Other assemblies would be shipped to the repository in canisters that were
either intended or not intended for disposal. Canisters not intended for disposal would be opened and
repackaged in waste packages in the Waste Handling Building.
For most of the DOE spent nuclear fuel categories, the fuel would be shipped in disposable canisters
(canisters that can be shipped and are suitable for direct insertion into waste packages without being
opened) in casks licensed by the Nuclear Regulatory Commission. Uranium oxide fuels with intact
zirconium alloy cladding are similar to commercial spent nuclear fuel and could be shipped either in DOE
standard canisters or as bare intact assemblies. Uranium metal fuels from Hanford and aluminum-based
fuels from the Savannah River Site could require additional treatment or conditioning before shipment to
the repository. If treatment was required, these fuels would be packaged in DOE disposable canisters.
Category 14 sodium-bonded fuels are also expected to require treatment before disposal.
High-level radioactive waste shipped to the repository would be in stainless-steel canisters. The waste
would have undergone a solidification process that yielded a leach-resistant material, typically a glass
form called borosilicate glass. In this process, the high-level radioactive waste is mixed with glass-
forming materials, heated and converted to a durable glass waste form, and poured into stainless-steel
canisters (Picha 1997, Attachment 4, page 2). Depending on future decisions stemming from other EISs,
ceramic and metal waste matrices could be sent to the repository from Argonne National Laboratory-West
A-6
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
in Idaho. The ceramic and metal matrices would be different solidified mixtures that also would be in
stainless-steel canisters. These wastes would be the result of the proposed electrometallurgical treatment
of sodium bonded fuels.
As briefly described in Section A. 1.1.1, the surplus weapon-usable plutonium would probably be sent to
the repository in two different waste forms — spent mixed-oxide fuel assemblies or an immobilized
plutonium ceramic form in a high-level radioactive waste canister and surrounded by high-level
radioactive waste. The spent mixed-oxide fuel assemblies would be very similar to conventional low-
enriched uranium assemblies and DOE would treat them as such. The immobilized plutonium would be
placed in small cans, inserted in the high-level radioactive waste canisters, and covered with molten
borosilicate glass (can-in-canister technique). The canisters containing immobilized plutonium and high-
level radioactive waste would be externally identical to the normal high-level radioactive waste canisters.
A.1 .1 .4 Waste Characteristics
A.1 .1 .4.1 Mass and Volume
As discussed in Section A.1, the Proposed Action includes 70,000 MTHM in the forms of commercial
spent nuclear fuel, DOE spent nuclear fuel, high-level radioactive waste, and surplus weapons-usable
plutonium. Figure A-2 shows percentages of MTHM included in the Proposed Action and the relative
amounts of the totals of the individual waste types included in the Proposed Action. As stated above, the
remaining portion of the wastes is included in Inventory Module 1. Because Greater-Than-Class-C and
Special-Performance-Assessment-Required wastes are measured in terms of volume. Figure A-3 shows
the relative volume of the wastes in Inventory Module 2 compared to the inventory in Module 1.
The No-Action Alternative (see Chapter 7 and Appendix K) used this information to estimate the mass
and volume of the spent nuclear fuel and high-level radioactive waste at commercial and DOE sites in
five regions of the contiguous United States.
The mass and volume data for commercial spent nuclear fiiel is the result of several years of annual
tracking and projections by DOE, which anticipates few changes in the overall mass and volume
projections for this waste type. The data projections for DOE spent nuclear fuel are fairly stable because
most of the projected inventory already exists, as opposed to having a large amount projected for future
generation. Mass and volume data for high-level radioactive waste estimates are not as reliable. Most
high-level radioactive waste currently exists as a form other than solidified borosilicate glass. The
solidification processes at the Savannah River Site and West Valley Demonstration Project are under
way; therefore, the resulting mass and volume are known. However, the processes at the Idaho National
Engineering and Environmental Laboratory and the Hanford Site have not started. Therefore, there is
some uncertainty about the mass and volume that would result from those processing operations. For this
analysis, DOE assumed that the high-level radioactive waste from the Hanford Site and the Idaho
National Engineering and Environmental Laboratory would represent 65 and 6 percent of the total high-
level radioactive waste inventory, respectively, in terms of the number of canisters.
A.1 .1 .4.2 Amount and Nature of Radioactivity
The primary purpose of presenting these data is to quantify the isotopic inventory expected in the
projected waste types. These data were used for accident scenario analyses associated with
transportation, handling, and repository operations. The data were also used to develop the source term
associated with accident scenarios and long-term effects for the Proposed Action and the No-Action
Alternative.
A-7
Inventory and Characteristics of Spent Nuclear Fuel, High-level Radioactive Waste, and Other Materials
c
T3 ^
a) o
(0 o
o o
2g
Q.
A-8
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Special-Performance-
Assessment-Required waste
5.3%
(4,000 cubic meters)
Greater-Than-Class-C waste
2.7%
(2,000 cubic meters)
To convert cubic meters
to cubic yards, multiply
by 1 .3079.
Module 2 relative volumes
(76,000 cubic meters)
Sources: Dirkmaat (19983. all); DOE (1994. all): DOE (1997b. page 1-8):
Heath (1998, Appendixes B and C); Picha (1997. Anachment 1.
page 1); Picha (1998a, Attachment 1|: Picha (1998b. all).
Figure A-3. Inventory Module 2 volume.
In a comparison of the relative amounts of radioactivity in a particular waste type, radionuclides of
concern depend on the analysis being performed. For example, cesium- 1 37 is the primary radionuclide of
concern when reviewing preclosure impacts and shielding requirements. For postclosure impacts, the
repository performance assessment evaluated nine radionuclides (see Appendix I) and identified
technetium-99 and neptunium-237 as the nuclides that provide the greatest impacts. Plutonium-238 and
-239 are shown in Chapter 7 to contribute the most to doses for the No-Action Alternative. Table A-2
presents the inventory of each of these radionuclides included in the Proposed Action. Figure A-4 shows
that at least 92 percent of the total inventory of each of these radionuclides is in commercial spent nuclear
fuel.
Table A-2. Selected nuclide inventory for the Proposed Action (curies).
Commercial
DOE
High-level
Surplus
spent nuclear fuel
spent nuclear fuel
radioactive waste
Plutonium
Totals
Cesium- 137
4.0x10'
1.7x10*
1.7x10*
NA"
4.3x10'
Technetium-99
9.2x10'
2.9x10"
2.1x10"
NA
9.7x10'
Neptunium-237
2.8x10"
1.1x10^
4.5x10^
NA
3.0x10"
Plutonium-238
2.1x10*
5.6x10*
3.0x10*
7.6x10"
2,2x10*
Plutonium -239
2.3x10^
3.8x10'
4.4x10"
1.0x10*
2.5x10'
a. NA = not applicable.
A.I .1 .4.3 Chemical Composition
The appendix presents data for the chemical composition of the primary waste types. For commercial
spent nuclear fuel, the elemental composition of typical pressurized-water and boiling-water reactor fuel
is provided on a per-assembly basis. Data are also provided on the number of stainless-steel clad
assemblies in the current inventory.
For DOE spent nuclear fuel and high-level radioactive waste, this appendix contains tables that describe
the composition of the total inventory of the spent nuclear fuel (by representative category) or high-level
radioactive waste (by site).
The chemical composition data were used primarily in the repository performance assessment (see
Chapter 5 and Appendix I) to evaluate the relative amounts of materials that would need further study.
A-9
Inventory and Characteristics of Spent Nuclear Fuel, High-level Radioactive Waste, and Other Materials
100
5 75
Cs»-137
Tc'-99
Np9-237
Radionuclide
Pu-238
Pu-239
a. Pu = surplus weapons-usable plutonlum;
Included In Proposed Action as spent nuclear
fuel and high-level radioactive waste.
b. HLW = high-level radioactive waste.
c. DSNF = DOE spent nuclear fuel.
d. CSNF = commercial spent
nuclear fuel
e. Cs = Cesium
f. Tc = Technetium
g. Np = Neptunium
Figure A-4. Proposed Action radionuclide distribution by material type.
■4 A AAA
9,000-
8,000 -
7,000 -
6,000 -
1 5,000 -
4,000 -
3,000 -
2,000 -
1,000 -
n _
8,800-
6,200
4,300
3,300
a. PWR = pressi
b. BWR = boilinc
21 PWRs^ ' 4
jrized-water reactor,
-water reactor.
4 BWRs'' '
Naval
Codisposal
Figure A-5. Thermal generation (watts per waste package),
A-10
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
As a result of an initial screening, the repository performance assessment evaluated the long-term impacts
of molybdenum, uranium, and chromium in the repository.
A.1 .1 .4.4 Thermal Output
Thermal generation data associated with each material type are provided in this appendix. These data
were used to develop the thermal loads associated with the repository design. Chapter 2 describes the
thermal load scenarios. The thermal data demonstrate that the EIS analysis can make simplifying
assumptions that the thermal output of the commercial spent nuclear fuel waste packages, particularly the
pressurized-water reactor assemblies, would bound the thermal output of all other waste packages (see
Figure A-5).
The data presented in the thermal output sections of this appendix for each waste type are presented as
watts per assembly or MTHM for commercial spent nuclear fuel, and watts per canister for DOE spent
nuclear fuel or high-level radioactive waste. Figure A-5 normalizes these data into a common, watts-per-
waste-package comparison. The following waste packages are compared: one containing 21 typical
pressurized-water reactor assemblies, one containing 44 typical boiling-water reactor assemblies, a co-
disposal waste package containing five high-level radioactive waste canisters and one DOE spent nuclear
fuel canister, and a waste package containing one dual-purpose canister of naval spent nuclear fuel (also a
DOE spent fuel). Another potential waste package containing four multi-canister overpacks of DOE
uranium metal fuels is not included in Figure A-5 because its estimated maximum thermal generation is
only 72 watts per waste package.
Figure A-5 uses conservative assumptions to illustrate the bounding nature of the thermal data for
commercial spent nuclear fuel. The commercial spent nuclear fuel data represent typical assemblies that
are assumed to have cooled for nearly 30 years. The naval spent nuclear fuel data are a best estimate of
the thermal generation of 5 -year old spent nuclear fuel. The thermal data selected for the high-level
radioactive waste are conservatively represented by the canisters from the Savannah River Site and are
combined with the highest values of thermal output from all projected DOE spent nuclear fuel categories.
A.1 .1.4.5 Canister Data
Typically, DOE spent nuclear fuel and high-level radioactive waste would be sent to the repository in
disposable canisters. The design specifications for DOE spent nuclear fuel canisters are in DOE (1998c,
all). These canisters are generally of two diameters — 46 and 61 centimeters (18 and 24 inches). They
also would be designed for two different lengths, nominally 3 and 4.6 meters (10 and 15 feet), to enable
co-disposal with high-level radioactive waste canisters. Certain DOE spent nuclear fuel categories
require specific disposal canister designs. Naval fuels would be sent to the repository in Navy dual-
purpose canisters, which are described in Dirkmaat (1997a, Attachment, pages 86 to 88) and USN (1996,
pages 3-1 to 3-11). N-Reactor fuels from the Hanford Site would be sent to the repository in
multicanister overpacks 64 centimeters (25.3 inches) in diameter, which are described in Parsons (1999,
all).
High-level radioactive waste would be sent to the repository in stainless-steel canisters, 61 centimeters
(25 inches) in diameter and either 3 or 4.6 meters (10 or 15 feet) in length, depending on the DOE site.
The canister design specifications are contained in Marra, Harbour, and Plodinec (1995, all) and WVNS
(1996, WQR-2.2, all) for the operating vitrification processes at Savannah River Site and West Valley
Demonstration Project, respectively. The other sites would use canister designs similar to those currently
in use (Picha 1997, all).
These data were for analysis of the No- Action Alternative (see Chapter 7 and Appendix K) to determine
the time required to breach the canisters after they are exposed to weather elements.
A-11
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
A.2 Materials
This section describes the characteristics of the materials DOE has considered for disposal in the
proposed Yucca Mountain Repository. All candidate materials would have to meet approved acceptance
criteria.
A.2.1 COMMERCIAL SPENT NUCLEAR FUEL
A.2.1.1 Background
Spent nuclear fuel is fuel that has been withdrawn from a nuclear reactor following irradiation. Spent
nuclear fuel from light-water reactors (pressurized-water and boiling-water reactors) would be the
primary source of radioactivity and thermal load in the proposed monitored geologic repository. Spent
nuclear fuels from civilian research reactors (General Atomics, Aerotest, etc.) account for less than 0.001
percent of the projected total in the Proposed Action (DOE 1995a, all). The fuels addressed in this
section are those discharged from commercial light-water reactors.
Section A.2.2 discusses the spent nuclear fuel from the Fort St. Vrain reactor in Colorado as part of DOE
spent nuclear fuels, as are the fuels from Shippingport, Three Mile Island-2, and other fuels from
commercial facilities that DOE is managing at its facilities.
A.2.1 .2 Sources
The sources of commercial spent nuclear fuel are the commercial nuclear powerplants throughout the
country. Table A-3 lists the individual reactors, reactor type, state, and actual or projected years of
operation. The operation period is subject to change if a utility pursues extension of the operating license
or shuts down early.
A.2.1. 3 Present Status
Nuclear power reactors store spent nuclear fuel in spent fuel pools under U.S. Nuclear Regulatory
Commission licenses, but they can use a combination of storage options: (1) in-pool storage and
(2) above-grade dry storage in an independent spent fuel storage installation. When a reactor is refueled,
spent fuel is transferred to the spent fuel pool, where it typically remains until the available pool capacity
is reached. When in-pool storage capacity has been fully used, utilities have turned to dry cask storage in
an independent spent fuel storage installation to expand their onsite spent fuel storage capacities, hi 1990,
the Nuclear Regulatory Commission amended its regulations to authorize licensees to store spent nuclear
fuel at reactor sites in approved storage casks (Raddatz and Waters 1996, all).
Commercial nuclear utilities currently use three Nuclear Regulatory Commission-approved general dry
storage system design types — metal storage casks and metal canisters housed in concrete casks and
concrete vaults — for use in licensed independent spent fuel storage installations. Raddatz and
Waters (1996, all) contains detailed information on models currently approved by the Commission.
Table A-4 lists existing and planned independent spent fuel storage installations in the United States.
A.2.1. 4 Final Spent Nuclear Fuel Form
The final form of commercial spent nuclear fuel to be disposed of in the proposed repository would be the
current reactor fuel assemblies. The repository would receive bare spent nuclear fuel assemblies, spent
nuclear fuel packaged in canisters not intended for disposal, and spent nuclear fuel packaged in canisters
intended for disposal.
A-12
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-3. Commercial nuclear power reactors in the United States and their projected years of
operation.^
Unit name
Reactor
type"
State
Operations
period'
Unit name
Reactor
type"
State
Operations
periocf
Arkansas Nuclear One 1
PWR
AR
1974-2014
Millstone 3
PWR
CT
1986-2025
Arkansas Nuclear One 2
PWR
AR
1978-2018
Monticello
BWR
MN
1971-2010
Beaver Valley 1
PWR
PA
1976-2016
Nine Mile Point 1
BWR
NY
1969-2009
Beaver Valley 2
PWR
PA
1978-2018
Nine Mile Point 2
BWR
NY
1987-2026
Big Rock Point
BWR
MI
1963-1997
North Anna 1
PWR
VA
1978-2018
Btaidwood 1
PWR
IL
1987-2026
North Anna 2
PWR
VA
1980-2020
Braidwood 2
PWR
IL
1988-2027
Oconee 1
PWR
SC
1973-2013
Browns Ferry 1
BWR
AL
1973-2013
Oconee 2
PWR
SC
1973-2013
Browns Ferry 2
BWR
AL
1974-2014
Oconee 3
PWR
SC
1974-2014
Browns Ferry 3
BWR
AL
1976-2016
Oyster Creek
BWR
NJ
1969-2009
Brunswick 1
BWR
NO
1976-2016
Palisades
PWR
MI
1972-2007
Brunswick 2
BWR
NO
1974-2014
Palo Verde 1
PWR
AZ
1985-2024
Byron 1
PWR
IL
1985-2024
Palo Verde 2
PWR
AZ
1986-2025
Byron 2
PWR
IL
1987-2026
Palo Verde 3
PWR
AZ
1987-2027
Callaway
PWR
MO
1984-2024
Peach Bottom 2
BWR
PA
1973-2013
Calvert Cliffs 1
PWR
MD
1974-2014
Peach Bottom 3
BWR
PA
1974-2014
Calvert CUffs 2
PWR
MD
1976-2016
Peiry 1
BWR
OH
1986-2026
Catawba 1
PWR
SC
1985-2024
Pilgrim 1
BWR
MA
1972-2012
Catawba 2
PWR
SC
1986-2026
Point Beach 1
PWR
Wl
1970-2010
Clinton
BWR
IL
1987-2026
Point Beach 2
PWR
Wl
1973-2013
Comanche Peak 1
PWR
TX
1990-2030
Prairie Island 1
PWR
MN
1974-2013
Comanche Peak 2
PWR
TX
1993-2033
Prairie Island 2
PWR
MN
1974-2014
Cooper Station
BWR
ME
1974-2014
Quad Cities 1
BWR
IL
1972-2012
Crystal River 3
PWR
PL
1977-2016
Quad Cities 2
BWR
IL
1972-2012
D. C. Cook 1
PWR
MI
1974-2014
Rancho Seco
PWR
CA
1974-1989
D. C. Cook 2
PWR
MI
1977-2017
River Bend 1
BWR
LA
1985-2025
Davis-Besse
PWR
OH
1977-2017
Salem 1
PWR
NJ
1976-2016
Diablo Canyon 1
PWR
CA
1984-2021
Salem 2
PWR
NJ
1981-2020
Diablo Canyon 2
PWR
CA
1985-2025
San Onofre 1
PWR
CA
1967-1992
Dresden 1
BWR
IL
1959-1978
San Onofre 2
PWR
CA
1982-2013
Dresden 2
BWR
IL
1969-2006
San Onofre 3
PWR
CA
1983-2013
Dresden 3
BWR
DL
1971-2011
Seabrook 1
PWR
NH
1990-2026
Duane Arnold 1
BWR
L\
1974-2014
Sequoyah 1
PWR
TN
1980-2020
Edwin I. Hatch 1
BWR
GA
1974-2014
Sequoyah 2
PWR
TN
1981-2021
Edwin I. Hatch 2
BWR
GA
1978-2018
Shearon Harris
PWR
NC
1987-2026
Fermi 2
BWR
MI
1985-2025
Shoreham
BWR
NY
1989^
Fort Calhoun 1
PWR
NE
1973-2013
South Texas Project 1
PWR
TX
1988-2016
Ginna
PWR
NY
1969-2009
South Texas Project 2
PWR
TX
1989-2023
Grand Gulf 1
BWR
MS
1984-2022
St. Lucie 1
PWR
FL
1976-2016
Haddam Neck
PWR
CT
1968-1996
St. Lucie 2
PWR
FL
1983-2023
Hope Creek
BWR
NJ
1986-2026
Summer 1
PWR
SC
1982-2022
Humboldt Bay
BWR
CA
1962-1976
Surry 1
PWR
VA
1972-2012
H.B. Robinsoo 2
PWR
SC
1970-2010
Surry 2
PWR
VA
1973-2013
Indian Point 1
PWR
NY
1%2-1974
Susquehanna 1
BWR
PA
1982-2022
Indian Point 2
PWR
NY
1973-2013
Susquehanna 2
BWR
PA
1984-2024
Indian Point 3
PWR
NY
1976-2015
Three Mile Island 1
PWR
PA
1974-2014
James A. HtzPatrick/
BWR
NY
1974-2014
Trojan
PWR
OR
1975-1992
Nine Mile Point
Turkey Point 3
PWR
FL
1972-2012
Joseph M. Farley 1
PWR
AL
1977-2017
Turkey Point 4
PWR
FL
1973-2013
Joseph M. Farley 2
PWR
AL
1981-2021
Vermont Yankee
BWR
VT
1973-2012
Kewaunee
PWR
Wl
1973-2013
Vogtie 1
PWR
GA
1987-2027
LaCrosse
BWR
WI
1%7-1987
VogUe2
PWR
GA
1989-2029
LaSalle 1
BWR
IL
1970-2022
Washington Public
BWR
WA
1984-2023
LaSalle 2
BWR
IL
1970-2023
Power Supply System 2
Limerick 1
BWR
PA
1985-2024
Waterford 3
PWR
LA
1985-2024
Limerick 2
BWR
PA
1989-2029
Watts Bar 1
PWR
TN
19%-2035
Maine Yankee
PWR
ME
1972-1996
Wolf Creek
PWR
KS
1985-2025
McGuire 1
PWR
NC
1981-2021
Yankee-Rowe
PWR
MA
1963-1991
McGuire 2
PWR
NC
1983-2023
Zion 1
PWR
IL
1973-1997
Millstone 1
BWR
CT
1970-2010
Zion2
PWR
IL
1974-1996
Millstone 2
PWR
CT
1975-2015
a. Source: DOE (1997a, Appendix C).
b. PWR = pressurized-water reactor; BWR = boiling-water reactor.
c. As defmed by current shutdown or full operation through license period (as of 1997).
d. Shoreham is no longer a Ucensed plant and has transferred all fuel to Limerick.
A- 13
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-4. Sites with existing or planned independent spent fuel storage installations."
Reactor
Status
Reactor
Status
Prairie Island
Existing
Rancho Seco
Planned
Point Beach
Existing
Trojan
Planned
Palisades
Existing
Washington Public Power Supply System
Planned
Surry
Existing
Big Rock Point
Planned
Calvert Cliffs
Existing
Oyster Creek
Planned
Arkansas Nuclear
Existing
Duane Arnold
Planned
H. B. Robinson
Existing
McGuire
Planned
Oconee
Exisdng
Yankee Rowe
Planned
Davis-Besse
Existing
Maine Yankee
Planned
North Anna
Planned
Peach Bottom
Planned
James A. FitzPatrick/Nine Mile Point
Planned
Palo Verde
Planned
Dresden
Planned
Humboldt Bay
Planned
Susquehanna
Planned
a. Sources: Raddatzand Waters (1996, all); Cole (1998a. all).
A.2.1 .5 Spent Nuclear Fuel Characteristics
There are 22 classes of nuclear fuel assemblies, with 127 individual fuel types in those classes. Seventeen
of the classes are for pressurized-water reactor fuels and 5 are for boiling-water reactors (DOE 1992,
Appendix 2A). For this EIS, the typical assemblies chosen for analysis represent an assembly type being
used in the more recently built reactors. This results in physical characteristics that might be slightly
higher than average (size, uranium per assembly, etc.), but that, however, provide a realistic estimate for
EIS analyses. Specifically chosen to represent the typical fuel types were the Westinghouse 17 x 17
LOPAR fuel assembly for the pressurized-water reactor and the General Electric BWR/4-6, 8x8 fuel
assembly for the boiling-water reactor. Table A-5 lists the fissile content and performance parameters
selected to define the radiological characteristics of these typical fuel assemblies.
Table A-5. Typical spent nuclear fuel
parameters."
Fuel type""
Bumup
(MWdMTHM)"^
Initial enrichment
(percent of U-235
by weight)
Age
(years)
Typical PWR
Typical BWR
39,560
32,240
3.69
3.00
25.9
27.2
a. Source: TRW (1998, page 3-15).
b. PWR = pressurized-water reactor; BWR = boiling-water reactor.
c. MWd/MTHM = megawatt-days per metric ton of heavy metal; to convert
meuic tons to tons, multiply by 1.1023.
A.2.1 .5.1 Mass and Volume
As discussed in Section A.l, the Proposed Action includes 63,(XX) MTHM of commercial spent nuclear
fuel. For the No-Action Alternative (continued storage) analysis, Table A-6 lists the distribution of this
expected inventory by reactor site. The historic and projected spent nuclear fuel discharge and storage
information in Table A-6 is consistent with the annual projections provided by the Energy Information
Administration (DOE 1997a, page 32). The "1995 Actual" data presented in Table A-6 represents the
amount of spent nuclear fuel stored at a particular site regardless of the reactor from which it was
discharged. For analysis purposes, the table lists spent nuclear fuel currently stored at the General
Electric Morris, Illinois, facility to be at Dresden, because these facilities are located near each other.
For analyses associated with the Proposed Action, the projected spent nuclear fuel from pressurized-water
reactors comprises 65 percent of the 63,000 metric tons of heavy metal (TRW 1997, page A-2). The
A-14
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Fuel
1995
1996-
Equivaleni
t
Fuel
1995
1996-
Equivalent
Site
type"
actual
201 r
Total"
assemblies
i Site
type"
actual
201 r
Total" assemblies
Arkansas Nuclear One
PWR
643
466
1,109
2,526
Monticello
BWR
147
280
426
2,324
Beaver Valley
PWR
437
581
1,018
2,206
North Anna
PWR
570
1,098
613
1,184
2,571
Big Rock Point
BWR
44
14
58
439
Oconee
PWR
767
1,865
4,028
Braidwood
PWR
318
711
1,029
2,424
Oyster Creek
BWR
374
325
699
3,824
Browns Ferry
BWR
840
1,092
1,932
10,402
Palisades
PWR
338
247
585
1,473
Brunswick
Both
448
448
896
4,410
Palo Verde
PWR
556
1,118
1,674
4,082
Byron
PWR
404
664
1,068
2,515
Peach Bottom
BWR
908
645
1,554
8,413
Callaway
PWR
280
422
702
1,609
Perry
BWR
178
274
452
2,470
Calvert Cliffs
PWR
641
501
1,142
2,982
Pilgrim
BWR
326
201
527
2,853
Catawba
PWR
465
683
1,148
2,677
Point Beach
PWR
529
347
876
2,270
Clinton
BWR
174
303
477
2,588
Prairie Island
PWR
518
348
866
2,315
Comanche Peak
PWR
176
821
998
2,202
Quad Cities
BWR
813
464
1,277
6,953
Cooper
BWR
175
277
452
2,435
Rancho Seco
PWR
228
e
228
493
Crystal River
PWR
280
232
512
1,102
River Bend
BWR
176
356
531
2,889
D. C. Cook
PWR
777
656
1,433
3,253
Salem/Hope Creek
Both
793
866
1,659
7,154
Davis-Besse
PWR
243
262
505
1,076
San Onofre
PWR
722
701
1,423
3,582
Diablo Canyon
PWR
463
664
1,126
2,512
Seabrook
PWR
133
292
425
918
Dresden
BWR
1,557
590
2,146
11,602
Sequoyah
PWR
452
570
1,023
2,218
Duane Arnold
BWR
258
208
467
2,545
Shearon Harris
Both
498
252
750
2,499
Edwin I. Hatch
BWR
755
692
1,446
7,862
South Texas Projeci
t PWR
290
722
1,012
1,871
Fermi
BWR
155
368
523
2,898
St. Lucie
PWR
601
419
1,020
2,701
Fort Calhoun
PWR
222
157
379
1,054
Summer
PWR
225
301
526
1,177
Ginna
PWR
282
180
463
1,234
Surry
PWR
660
534
1,194
2,604
Grand Gulf
BWR
349
506
856
4,771
Susquehanna
BWR
628
648
1,276
7,172
H. B. Robinson
PWR
145
239
384
903
Three Mile Island
PWR
311
236
548
1,180
Haddam Neck
PWR
355
65
420
1,017
Trojan
PWR
359
-
359
780
Humboldt Bay
BWR
29
--
29
390
Turkey Point
PWR
616
458
1,074
2,355
Indian Point
PWR
678
486
1,164
2,649
Vermont Yankee
BWR
387
222
609
3,299
James A. FitzPatrick/
BWR
882
930
1,812
9,830
Vogtle
PWR
335
745
1,080
2,364
Nine Mile Point
Washington Public
BWR
243
338
581
3,223
Joseph M. Farley
PWR
644
530
1,174
2,555
Power Supply
Kewaunee
PWR
282
169
451
1,172
System 2
La Crosse
BWR
38
-
38
333
Waterford
PWR
253
247
500
1,217
La Salle
BWR
465
487
952
5,189
Watts Bar
PWR
-
251
251
544
Limerick
BWR
432
711
1,143
6,203
Wolf Creek
PWR
226
404
630
1,360
Maine Yankee
PWR
454
82
536
1,421
Yankee-Rowe
PWR
127
-
127
533
McGuire
PWR
714
725
1,439
3,257
Zion
PWR
841
211
1,052
2,302
Millstone
Both
959
749
1,709
6,447
Totals
31^6
31,074
63,000
218,700
a. Source: Heath (1998, Appendixes B and C).
b. PWR = pressurized-water reactor; BWR = boiling-water reactor.
c. Projected.
d. To convert metric tons to tons, multiply by 1.1023.
e. " = no spent nuclear fuel production.
balance consists of spent nuclear fuel from boiling-water reactors. Using the nominal volume for the
spent nuclear fuel assemblies described in Section A.2. 1.5.5, the estimated volume of spent nuclear fuel in
the Proposed Action, exclusive of packaging, is 29,000 cubic meters.
Section A.l also discusses the additional inventory modules evaluated in this EIS. Inventory Modules 1
and 2 both include the maximum expected discharge inventory of commercial spent nuclear fuel.
Table A-7 lists historic and projected amounts of spent nuclear fuel discharged from commercial reactors
through 2046. The estimated unpackaged volume of spent nuclear fuel for these modules is
approximately 47,000 cubic meters. For conservatism, these data were derived from the Energy
Information Administration "high case" assumptions. The high case assumes that all currently operating
nuclear units would renew their operating licenses for an additional 10 years (DOE 1997a, page 32).
A-15
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-7. Inventory
Modules 1 and 2 spent nuclear fuel inventory (MTHM)."
Fuel
1995
Equivalent
Fuel
1995
1996-
Equivalent
Site
type"
actual
1996-2046^
Total''
assembhes
Site
type"
actual
2046'
Total"
assemblies
Arkansas Nuclear One
PWR
643
1,007
1,650
3,757
Monticello
BWR
147
390
537
2,924
Beaver Valley
PWR
437
1,395
1,832
3,970
North Anna
PWR
570
1,384
1,955
4,246
Big Rock Point
BWR
44
14
58
439
Oconee
PWR
1,098
1,576
2,674
5,774
Braidwood
PWR
318
1,969
2,287
5,385
Oyster Creek
BWR
374
.470
844
4,619
Browns Ferry
BWR
840
2,508
3,348
18,024
Palisades
PWR
338
395
733
1,845
Brunswick
Both
448
992
1,440
7,355
Palo Verde
PWR
556
3,017
3,573
8,712
Byron
PWR
404
1,777
2,181
5,139
Peach Bottom
BWR
908
1,404
2,312
12,523
Callaway
PWR
280
1,008
1,288
2,953
Perry
BWR
178
732
910
4,974
Calvert Cliffs
PWR
641
1,069
1,710
4,466
Pilgrim
BWR
326
444
770
4,170
Catawba
PWR
465
1,752
2,217
5,168
Point Beach
PWR
529
614
1,143
2,961
Clinton
BWR
174
910
1,084
5,876
Prairie Island
PWR
518
692
1,210
3,234
Comanche Peak
PWR
176
2,459
2,635
5,816
Quad Cities
BWR
813
1,020
1,834
9,982
Cook
PWR
777
1,379
2,155
4,892
Rancho Seco
PWR
228
e
228
493
Cooper
BWR
175
587
762
4,106
River Bend
BWR
176
956
1,132
6,153
Crystal River
PWR
280
525
805
1,734
Salem/Hope Creek
Both
793
2,452
3,245
11,584
Davis-Besse
PWR
243
582
825
1,757
San Onofre
PWR
722
1,321
2,043
5,144
Diablo Canyon
PWR
463
1,725
2,187
4,878
Seabrook
PWR
133
831
964
2,083
Dresden
BWR
1,557
984
2,541
13,740
Sequoyah
PWR
452
1,393
1,845
4,001
Duane Arnold
BWR
258
434
692
3,776
Shearon Harris
Both
498
707
1,205
3,535
Fermi
BWR
155
1,005
1,160
6,429
South Texas Project
PWR
290
2,029
2,319
4,286
Fort Calhoun
PWR
222
312
534
1,485
St. Lucie
PWR
601
1,010
1,611
4,265
Ginna
PWR
282
283
565
1,507
Summer
PWR
225
732
958
2,141
Grand Gulf
BWR
349
1,261
1,610
8,976
Surry
PWR
660
1,029
1,689
3,682
H. B. Robinson
PWR
145
364
509
1,197
Susquehanna
BWR
628
1,745
2,373
13,338
Haddam Neck
PWR
355
65
420
1,017
Three Mile Island
PWR
311
513
825
1,777
Hatch
BWR
755
1,517
2,272
12,347
Trojan
PWR
359
..
359
780
Humboldt Bay
BWR
29
--
29
390
Turkey Point
PWR
616
905
1,520
3,334
Indian Point
PWR
678
1,005
1,683
3,787
Vermont Yankee
BWR
387
434
822
4,451
James A. FitzPatrick/
BWR
882
2,018
2,900
15,732
Vogtle
PWR
335
2,122
2,458
5,378
Nine Mile Point
Washington Public
BWR
243
924
1,167
6,476
Joseph M. Farley
PWR
644
1,225
1,869
4,070
Power Supply
Kewaunee
PWR
282
330
612
1,591
System 2
La Crosse
BWR
38
--
38
333
Waterford
PWR
253
685
938
2,282
La Salle
BWR
465
1,398
1,863
10,152
Watts Bar
PWR
—
893
893
1,937
Limerick
BWR
432
1,958
2,390
12,967
Wolf Creek
PWR
226
1,052
1,278
2,759
Maine Yankee
PWR
454
82
536
1,421
Yankee-Rowe
PWR
127
—
127
533
McGuire
PWR
714
1,813
2,527
5,720
Zion
PWR
841
211
1,052
2,302
Millstone
Both
959
1,695
2,655
8,930
Totals
31,926
73,488
105,414
359,963
a. Source: Heath (1998, Appendixes B and C).
b. PWR = pressurized-water reactor; BWR = boiling-water reactor.
c. Projected.
d. To convert metric tons to tons, multiply by 1 .1023.
e. - = no spent nuclear fuel production.
A.2.1 .5.2 Amount and Nature of Radioactivity
DOE derived radionuclide inventories for the typical pressurized-water reactor and boiling-water reactor
fuel assemblies from the Light-Water Reactor Radiological Database (DOE 1992, page 1.1-1). The
inventories are presented at the average decay years for each of the typical assemblies. Tables A-8 and
A-9 list the inventories of the nuclides of interest for the typical assemblies for both reactor types.
Table A- 10 combines the typical inventories (curies per MTHM) with the projected totals (63,000
MTHM and 105,(XX) MTHM) to provide a total projected radionuclide inventory for the Proposed Action
and additional modules.
A.2.1 .5.3 Chemical Composition
Commercial spent nuclear fuel consists of the uranium oxide fuel itself (including actinides, fission
products, etc.), the cladding, and the assembly hardware.
A-16
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-8. Radionuclide activity
for typical pressurized-water reactor fuel assemblies."''
Curies per
Curies per
Curies per
Isotope
assembly
Isotope
assembly
Isotope
assembly
Hydrogen-3
9.8x10'
Cesium- 134
1.6x10'
Neptunium-237
2.3x10'
Carbon- 14
6.4x10'
Cesium- 135
2.5x10'
Piutonium-238
1.7x10'
Chlorine-36
5.4x10-'
Cesium- 137
3.1x10"
Plutonium-239
1.8x10'
Cobalt-60
1.5x10^
Samarium- 151
1.9x10^
Plutonium-240
2.7x10'
Nickel-59
1.3
Lead-210
2.2x10-'
Plutonium-241
2.0x10"
Nickel-63
1.8x10^
Radium-226
9.3x10-'
Plutonium-242
9.9x10'
Selenium-79
2.3x10'
Radium-228
1.3x10"'
Americium-241
1.7x10'
Krypton-85
9.3x10^
Actinium-227
7.8x10-*
Americium-242/242m
1.1x10'
Strontium-90
2.1x10"
Thorium-229
1.7x10-'
Americium-243
1.3x10'
Zirconium-93
1.2
Thorium-230
1.5x10-"
Curium-242
8.7
Niobium-93m
8.2x10'
Thorium-232
1.9x10'°
Curium-243
8.3
Niobium-94
5.8x10'
Protactinium-23 1
1.6x10-^
Curium-244
7.0x10'
Technetium-99
7.1
Uranium-232
1.9x10-^
Curium-245
1.8x10'
Rhodium- 102
1.2x10'
Uranium-233
3.3x10-'
Curium-246
3.8x10'
Ruthenium- 106
4.8x10"'
Uranium-234
6.6x10'
Curium-247
1.3x10'
Palladium- 107
6.3x10-^
Uranium-235
8.4x10'
Curium-248
3.9x10'
Tin- 126
4.4x10'
Uranium-236
1.4x10'
Califomium-252
3.1x10'
Iodine- 129
1.8x10-^
Uranium-238
1.5x10'
a. Source: DOE (1992
:, page 1.1-1).
b. Bumup = 39,560 MWd/MTHM, enrichment = 3.69 percent, decay time = 25.9
years.
Table A-9. Radionuclide activity
for typical boiling-
water reactor fuel assemblies/''
Curies per
Curies per
Curies per
Isotope
assembly
Isotope
assembly
Isotope
assembly
Hydrogen-3
3.4x10'
Cesium- 134
3.4
Neptunium-237
7.3x10'
Carbon- 14
3.0x10'
Cesium- 135
l.OxlO'
Plutonium-238
5.5x10'
Chlorine-36
2.2x10'
Cesium- 137
1.1x10"
Plutonium-239
6.3x10'
Cobalt-60
3.7x10'
Samarium- 1 5 1
6.6x10'
Plutonium-240
9.5x10'
Nickel-59
3.5x10'
Lead-210
9.4x10-'
Plutonium-241
7.5x10'
Nickel-63
4.6x10'
Radium-226
3.7x10-'
Plutonium-242
4.0x10'
Selenium-79
7.9x10"^
Radium-228
4.7x10"
Americium-241
6.8x10'
ICrypton-85
2.9x10^
Actinium-227
3.1x10-*
Americium-242/242m
4.6
Strontium-90
7.1x10'
Thorium-229
6.1x10-'
Americium-243
4.9
Zirconium-93
4.8x10'
Thorium-230
5.8x10-'
Curium-242
3.8
Niobium-93m
3.5x10"'
Thorium-232
6.9x10"
Curium-243
3.1
Niobium-94
1.9x10-^
Protactinium-23 1
6.0x10*
Curium-244
2.5x10'
Technetium-99
2.5
Uranium-232
5.5x10'
Curium-245
6.3x10'
Rhodium- 102
2.8x10-"
Uranium-233
1.1x10-'
Curium-246
1.3x10'
Ruthenium- 106
6.7x10"
Uranium-234
2.4x10'
Curium-247
4.3x10-'
Palladium- 107
2.4x10-'
Uranium-235
3.0x10'
Curium-248
1.2x10'
Tin- 126
1.5x10'
Uranium-236
4.8x10-^
Califomium-252
6.0x10'
Iodine- 129
6.3x10'
Uranium-238
6.2x10'
a. Source: DOE (1992, page 1.1-1).
b. Bumup = 32,240 MWd/MTHM, enrichment = 3.00 percent, decay time = 27.2 years.
Typical pressurized-water and boiling-water reactor fuels consist of uranium dioxide with a zirconium
alloy cladding. Some assemblies, however, are clad in stainless-steel 304. Specifically, 2,187
assemblies, or 727 MTHM (1.15 percent of the MTHM included in the Proposed Action) are
stainless-steel clad (Cole 1998b, all). These assemblies have been discharged from Haddam Neck,
Yankee-Rowe, Indian Point, San Onofre, and LaCrosse. Table A-11 lists the number of assemblies
discharged, MTHM, and storage sites for each plant.
Tables A- 12 and A- 13 list the postirradiation elemental distributions for typical fuels. The data in these
tables include the fuel, cladding material, and assembly hardware.
A-17
Inventory and Characteristics of Spent Nuclear Fuel, High-level Radioactive Waste, and Other Materials
Table A- 10. Total projected radionuclide inventories.
Pressurized-water reactor
Boil
ling-water reactor
Grand tota
Proposed
Curies per
Total
curies
Curies per
Total
curies
Is (curies)
Proposed
Additional
Proposed
Additional
Additional
Isotope
mthm''
Action
modules
MTHM
Action
modules
Action
modules
Hydrogen-3
2.1x10^
8.6x10*
1,4x10'
1.7x10^
3.8x10*
6,4x10*
1.2x10'
2.1x10'
Carbon- 14
1.4
5.7x10"
9.5x10"
1.5
3.4x10"
5.7x10"
9.1x10"
1.5x10'
Chlorine-36
1.2x10"^
4.7x10^
7,9x10^
1,1x10"^
2.5x10^
4.1x10^
7.2x10^
1,2x10'
Cobalt-60
3.2x10^
1.3x10'
2.2x10'
1,9x10^
4.2x10*
7.0x10*
1,7x10'
2.9x10'
Nickel-59
2.8
l.lxlO'
1.9x10'
1.8
4.0x10"
6.6x10"
1,5x10'
2.6x10'
Nickel-63
3.8x10^
1.6x10'
2,6x10'
2.3x10^
5.1x10*
8.6x10*
2,1x10'
3.5x10'
Selenium-79
4.9x10'
2.0x10"
3,3x10"
4.0x10"'
8.9x10'
1,5x10"
2,9x10"
4.8x10"
Krypton-85
2.0x10^
8.2x10'
1,4x10*
1.5x10'
3.3x10'
5.5x10'
1.1x10*
1.9x10*
Strontium-90
4.6x10"
1.9x10'
3,1x10'
3.6x10"
8.0x10*
1.3x10'
2.7x10'
4.5x10'
Zirconium-93
2.5
l.OxlO'
1,7x10'
2.4
5.4x10"
9.0x10"
1.6x10'
2.6x10'
Niobium-93m
1.8
7.3x10"
1,2x10'
1.8
3.9x10"
6.6x10"
l.lxlO'
1.9x10'
Niobium-94
1.3
5.1x10"
8.6x10"
9.8x10"^
2.2x10'
3.6x10'
5.3x10"
8,9x10"
Technetium-99
1.5x10'
6.3x10^
1,1x10*
1.3x10'
2.9x10'
4,8x10'
9.2x10'
1,5x10*
Rhodium- 102
2.6x10-^
1.1x10^
1,8x10^
1.4x10'
3.2x10'
5,3x10'
1,4x10^
2,3x10^
Ruthenium- 106
1.0x10'^
4.2x10^
7,0x10^
3.4x10"'
7.5x10'
1,3x10^
5,0x10^
8.3x10^
Palladium- 107
1.4x10'
5.6x10'
9,4x10'
1.2x10'
2.7x10'
4,5x10'
8,3x10'
1,4x10*
Tin- 126
9.4x10'
3.8x10"
6,4x10"
7.9x10"'
1.7x10°
2,9x10"
5,6x10"
9,3x10*
Iodine-129
3.8x10"^
1.5x10'
2,6x10'
3.2x10"^
7.0x10^
1,2x10'
2,2x10'
3,8x10'
Cesium- 134
3.5x10'
1.4x10*
2,4x10*
1.7x10'
3.8x10'
6,4x10'
1,8x10*
3,0x10*
Cesium- 135
5.5x10'
2.3x10"
3,8x10"
5.1x10"'
1.1x10"
1,9x10*
3.4x10"
5,6x10"
Cesium- 137
6.7x10*
2.8x10'
4.6x10'
5.4x10"
1.2x10'
2,0x10'
4,0x10'
6,6x10'
Samarium-151
4.0x10^
1.6x10'
2,7x10'
3.4x10^
7.4x10°
1.2x10'
2,4x10'
4,0x10'
Lead-210
4.8x10-^
2.0x10^
3,3x10"^
4.8x10'
1.1x10"^
1.8x10"^
3.0x10"^
5,1x10^
Radium-226
2.0x10*
8.2x10"^
1,4x10"'
1.9x10*
4.2x10"^
7.0x10"^
1,2x10"'
2.1x10'
Radium-228
2.8x10"'
1.1x10"'
1.9x10'
2.4x10"'"
5.3x10*
8.9x10*
1,7x10"'
2.8x10"'
Actinium-227
1.7x10"'
6.9x10"'
1.2
1.6x10"'
3.5x10"'
5.8x10'
1,0
1.7
Thorium-229
3.8x10"^
1.5x10"^
2.6x10"^
3.1x10'
6.9x10"'
1.2x10"^
2,2x10"^
3.7x10"^
Thorium-230
3.3x10"
1.4x10'
2.3x10'
3.0x10"
6,6x10
1.1x10'
2,0x10'
3.4x10'
Thorium-232
4.1x10'°
1.7x10"'
2.8x10"'
3.5x10"'°
7,8x10*
1.3x10"'
2,5x10"'
4.1x10'
Protactinium-231
3.4x10-'
1.4
2.3
3.1x10"'
6,8x10"'
1.1
2,1
3.5
Uranium-232
4.0x10^
1.6x10'
2.7x10'
2.8x10"^
6,2x10^
1.0x10'
2,3x10'
3.8x10'
Uranium-233
7.1x10'
2.9
4.9
5.4x10"'
1,2
2.0
4,1
6.9
Uranium-234
1,4
5.8x10"
9.7x10"
1.2
2,7x10"
4.5x10"
8,5x10*
1.4x10'
Uranium-235
1.8x10"^
7.4x10^
1.2x10'
1.5x10"^
3,4x10^
5.6x10^
1,1x10'
1,8x10'
Uranium-236
3.0x10'
1.2x10"
2.1x10"
2,4x10"'
5,4x10'
9.0x10'
1,8x10"
3,0x10"
Uranium-238
3.1x10'
1.3x10"
2.2x10"
3.2x10'
7,0x10'
1.2x10"
2,0x10"
3,3x10"
Neptunium-237
4.9x10"'
2.0x10"
3.4x10"
3.7x10"'
8,2x10'
1.4x10"
2,8x10"
4,7x10"
Plutonium-238
3.6x10^
1.5x10*
2.5x10*
2.8x10'
6,1x10'
1.0x10*
2.1x10*
3.5x10*
Plutonium-239
3.9x10^
1.6x10'
2.7x10'
3.2x10^
7,1x10*
1.2x10'
2.3x10'
3.9x10'
Plutonium-240
5.8x10^
2.4x10'
4.0x10'
4.9x10^
1,1x10'
1.8x10'
3.4x10'
5,8x10'
Plutonium-241
4.4x10"
1.8x10'
3.0x10'
3,8x10"
8,4x10*
1,4x10'
2.6x10'
4,4x10'
Plutonium-242
2.1
8.7x10"
1.5x10'
2,0
4,5x10"
7,5x10"
1.3x10'
2,2x10'
Americium-241
3.7x10'
1.5x10*
2.5x10*
3,5x10'
7,7x10'
1.3x10*
2.3x10*
3,8x10*
Americium-242/242m
2.3x10'
9.3x10'
1.6x10*
2,3x10'
5,2x10'
8,7x10'
1,4x10*
2,4x10*
Americium-243
2.7x10'
1.1x10*
1.9x10*
2,5x10'
5,5x10'
9,2x10'
1,7x10*
2,8x10*
Curium-242
1.9x10'
7.7x10'
1,3x10*
1.9x10'
4.3x10'
7,1x10'
1,2x10*
2,0x10*
Curium-243
1.8x10'
7.3x10'
1.2x10*
1.6x10'
3,5x10'
5,8x10'
1,1x10*
1,8x10*
Curium-244
1.5x10'
6.2x10'
1.0x10*
1,3x10'
2,8x10'
4,7x10'
9,0x10'
1,5x10*
Curium-245
3.9x10'
1.6x10"
2,7x10"
3,2x10'
7.1x10'
1,2x10"
2,3x10"
3,8x10"
Curium-246
8.2x10"^
3.4x10'
5,6x10'
6,5x10^
1,4x10'
2.4x10'
4.8x10'
8,0x10'
Curium-247
2.9x10^
1.2x10^
2,0x10"^
2.2x10"'
4.8x10"'
8,1x10"'
1,6x10^
2.8x10"^
Curium-248
8.3x10"^
3.4x10"^
5,7x10"^
6.1x10'
1,4x10"^
2,3x10"^
4,8x10"^
8,0x10"^
Califomium-252
6.7x10"*
2,8x10'
4,6x10"'
3.1x10*
6,8x10"
1.1x10"'
3,4x10"'
5,7x10"'
a. Source: Compilation of Tables A-8 and A-9,
b, MTHM = metric tons of heavy metal.
A-18
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-11. Stainless-steel-clad spent nuclear fuel inventory.*
Discharging reactor
Storage location Assemblies
MTHM*"
Yankee-Rowe
Yankee-Rowe
76
21
San Onofre 1
San Onofre
395
144
San Onofre 1
Morris, Illinois
270
99
Indian Point 1
Indian Point
160
31
LaCrosse
LaCrosse
333
38
Haddam Neck
Haddam Neck
871
360
Haddam Neck
Morris, Illinois
82
34
Totals
2,187
727
a. Source: Cole (1998b, all).
b. MTHM = metric tons of heavy metal.
Table A-12. Elemental distribution of
typical pressurized-water reactor fue
."
Grams per
Grams per
Element
assembly*" Percent total'
Element
assembly""
Percent total*^
Aluminum
47
0.01
Oxygen
62,000
9.35
Americium
600
0.09
Palladium
790
0.12
Barium
1,200
0.18
Phosphorus
85
0.01
Cadmium
77
0.01
Plutonium
4,600
0.69
Carbon
77
0.01
Praseodymium
610
0.09
Cerium
1,300
0.20
Rhodium
230
0.04
Cesium
1,100
0.17
Rubidium
200
0.03
Chromium
4,300
0.65
Ruthenium
1,200
0.18
Cobalt
38
0.01
Samarium
470
0.07
Europium
72
O.OI
Silicon
170
0.03
Gadolinium
81
0.0 1
Silver
40
0.01
Iodine
130
0.02
Strontium
330
0.05
Iron
12,000
1.85
Technetium
420
0.06
Krypton
190
0.03
Tellurium
270
0.04
Lanthanum
670
O.IO
Tin
1,900
0.29
Manganese
330
0.05
Titanium
51
0.01
Molybdenum
2,000
0.31
Uranium
440,000
65.78
Neodymium
2,200
0.33
Xenon
2,900
0.43
Neptunium
330
0.05
Yttrium
250
0.04
Nickel
5,000
0.75
Zirconium
120,000
17.77
Niobium
330
0.05
Nitrogen
49
0.01
Totals
668,637
99.99
a. Source: DOE (1992, page 1.1-1).
b. To convert grams to ounces, multiply by 0.035274.
c. Table only includes elements that constitute at least 0.01 percent of the total; therefore, the total of the percentage column is
slightly less than 100 percent.
A.2.1 .5.4 Thermal Output
Heat generation rates are available as a function of spent fuel type, enrichment, bumup, and decay time in
the Light-Water Reactor Radiological Database, which is an integral part of the Characteristics Potential
Repository Wastes (DOE 1992, page 1.1-1). Table A-14 lists the thermal profiles for the typical
pressurized-water reactor and boiling-water reactor assemblies from the Light-Water Reactor
Radiological Database. For the EIS analysis, the typical thermal profile, applied across the proposed
inventory, yields a good approximation of the expected thermal load in the repository. Figure A-6 shows
these profiles as a function of time.
A-19
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-13. Elemental distribution of typical boiling-water reactor fuel/
Grams per
Percent
Grams per
Percent
Element
assembly"'
totaf Element
assembly''
total"
Aluminum
31
0.01 Nitrogen
25
0.01
Americium
220
0.07 Oxygen
25,000
7.82
Barium
390
0.12 Palladium
270
0.09
Cadmium
27
0.01 Plutonium
1,500
0.48
Carbon
36
0.01 Praseodymium
200
0.06
Cerium
430
0.14 Rhodium
79
0.03
Cesium
390
0.12 Rubidium
64
0.02
Chromium
1,900
0.60 Ruthenium
410
0.13
Cobalt
26
0.01 Samarium
160
0.05
Europium
24
0.01 Silicon
80
0.03
Gadolinium
310
0.10 Strontium
110
0.03
Iodine
43
0.01 Technetium
140
0.04
Iron
5,100
1.63 Tellurium
91
0.03
Krypton
62
0.02 Tin
1,600
0.50
Lanthanum
220
0.07 Titanium
83
0.03
Manganese
160
0.05 Uranium
170,000
55.35
Molybdenum
630
0.20 Xenon
950
0.30
Neodymium
730
0.23 Yttrium
81
0.03
Neptunium
97
0.03 Zirconium
96,000
30.52
Nickel
3,000
0.94
Niobium
29
0.01 Totals
310,698
99.94
a. Source: DOE (1992, page 1.1-1)
b. To convert grains to ounces, multiply by 0.035274.
c. Table only includes elements that contribute at least 0.01 percent of the total; therefore, the total of the percentage
column is slightly less than 100 percent.
Table A-14. Typical assembly thermal profiles.^
Years after _
Pressurized-water reactor Boiling-water reactor
discharge
w/mthm"
W/assembly' W/MTHM W/assembly''
1
10,500
4,800 8,400
1,500
3
3,700
1,700 3,000
550
5
2,200
1,000 1,800
340
10
1,500
670 1,200
220
26
990
450 820
150
30
920
420 770
140
50
670
310 570
100
100
370
170 320
58
300
160
73 140
26
500
120
53 100
19
1,000
66
31 58
11
2,000
35
16 30
5
5,000
22
10 19
3
10,000
16
8 13
3
a. Source : DOE ( 1 992, page 1.1-1).
b. W/MTHM = waUs per metric ton of heavy metal; to convert metric tons to tons, multiply by 1 . 1023.
c. W/assembly = watts per assembly; assumes 0.46 MTHM per assembly.
d. Assumes 0. 18 MTHM per assembly.
A-20
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Boiling-water reactor
Pressurized-water reactor
10,000
Source: DOE (1992, page 1.M).
Figure A-6. Typical thermal profiles over time.
A.2.1 .5.5 Physical Parameters
Table A- 1 5 lists reference characteristics of typical pressurized-water and boiling- water reactor fuel
assemblies. These data are from the Integrated Data Base Report (DOE 1997b, page 1-8) and reflect
characteristics of unirradiated assemblies.
Table A-15. Reference characteristics for unirradiated typical fuel assemblies."
Boiling-water reactor Pressurized-water reactor
Characteristics
Overall assembly length (meters)
Cross section (centimeters)
Fuel rod length (meters)
Active fuel height (meters)
Fuel rod outer diameter (centimeters)
Fuel rod array
Fuel rods per assembly
Assembly total weight (kilograms)
Uranium per assembly (kilograms)
Uranium oxide per assembly (kilograms)
Zirconium alloy per assembly (kilograms)
Hardware per assembly (kilograms)
Nominal volume per assembly (cubic meters)
4.5
14x14
4.1
3.8
1.3
8x8
63
320
180
210
100'
8.6'
0.086«
4.1
21x21
3.9
3.7
0.95
17x17
264
660
460
520
110"
26'
0.1 9«
a. Source: DOE (1997b, page 1-8).
b. To convert meters to feet, multiply by 3.2808; to convert centimeters to inches, multiply by 0.3937; to convert kilograms to
pounds, multiply by 2.2046; to convert cubic meters to cubic feet, multiply by 35.314.
c. Includes zirconium alloy fuel rod spacers and fuel channels.
d. Includes zirconium alloy control rod guide thimbles.
e. Includes stainless-steel tie plates, Inconel springs, and plenum springs.
f. Includes stainless-steel nozzles and Inconel-718 grids.
g. Based on overall outside dimension; includes spacing between the stacked fuel rods of the assembly.
For additional details, the Light-Water Reactor Assembly Database contains individual physical
descriptions of the fuel assemblies and fuel pins. The Light- Water Reactor Nonfuel Assembly Hardware
Database contains physical and radiological descriptions of nonfuel assembly hardware. These databases
are integral parts of the Characteristics of Potential Repository Wastes (DOE 1992, Section 2.8).
A-21
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
A.2.2 DOE SPENT NUCLEAR FUEL
A.2.2.1 Background
At present, DOE stores most of its spent nuclear fuel at three primary locations: the Hanford Site in
Washington State, the Idaho National Engineering and Environmental Laboratory iii Idaho, and the
Savannah River Site in South Carolina. Some DOE spent nuclear fuel is stored at the Fort St. Vrain dry
storage facility in Colorado. Much smaller quantities remain at other locations (LMIT 1997, all). DOE
issued the Record of Decision - Department of Energy Programmatic Spent Nuclear Fuel Management
and Idaho National Engineering Laboratory Environmental Restoration and Waste Management
Programs Final Environmental Impact Statement on June 1, 1995 (DOE 1995b, all) and amended it in
March 1996 (DOE 1996, all). The Record of Decision and its amendment specify three primary locations
as storage sites for DOE spent nuclear fuel. With the exception of Fort St. Vrain, which will retain its
spent nuclear fuel in dry storage, DOE will ship all its spent nuclear fuel from other sites to one of the
three primary sites for storage and preparation for ultimate disposition.
During the last four decades, DOE and its predecessor agencies have generated more than 200 varieties of
spent nuclear fuel from weapons production, nuclear propulsion, and research missions. A method
described by Fillmore (1998, all) allows grouping of these many varieties of spent nuclear fuel into
16 categories for the repository Total System Performance Assessment. The grouping method uses
regulatory requirements to identify the parameters that would affect the performance of DOE spent
nuclear fuel in the repository and meet analysis needs for the repository License Application. Three fuel
parameters (fuel matrix, fuel compound, and cladding condition) would influence repository performance
behavior. The grouping methodology presents the characteristics of a select number of fuel types in a
category that either bound or represent a particular characteristic of the whole category. Table A-16 lists
these spent nuclear fuel categories.
Table A-16 includes sodium-bonded fuel (Category 14); however, DOE is considering a proposal to treat
and manage sodium-bonded spent nuclear fuel for disposal. Alternatives being considered include
processing and converting some or all of its sodium-bonded fuel to a high-level radioactive waste form
before shipment. Section A.2.3, which covers data associated with high-level radioactive waste, includes
data on waste produced from potential future treatment of Category 14 spent nuclear fuel (Dirkmaat
1997b, page 7).
A.2.2.2 Sources
The DOE National Spent Fuel Program maintains a spent nuclear fuel data base (LMIT 1997, all). Table
A-16 provides a brief description of each of the fuel categories and a typical fuel. Section A.2.2.5.3
provides more detail on the chemical makeup of each category.
A.2.2.3 Present Storage and Generation Status
Table A-17 lists storage locations and inventory information on DOE spent nuclear fuels. During the
preparation of the Department of Energy Programmatic Spent Nuclear Fuel Management and Idaho
National Engineering Laboratory Environmental Restoration and Waste Management Programs Final
Environmental Impact Statement (DOE 1995c, all), DOE evaluated and categorized all the materials
listed in the table as spent nuclear fuel, in accordance with the definition in the Nuclear Waste Policy Act,
as amended.
A-22
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-16. DOE spent nuclear fuel categories
a,b
DOE SNF category Typically from
Description of fuel
1 . Uranium metal N-Reactor
2. Uranium-zirconium HWCTR
3.
Uranium-
molybdenum
Fermi
4.
Uranium oxide, intact
Commercial
PWR
5.
Uranium oxide, failed/
declad/aluminum
clad
TMI core debris
6. Uranium-aluminide
7. Uranium-silicide
8. Thorium/uranium
carbide, high-integrity
9. Thorium/uranium
carbide, low-integrity
10. Plutonium/uranium
carbide, nongraphite
11. Mixed oxide
12. Uranium/thorium
oxide
13. Uranium-zirconium
hydride
14. Sodium-bonded
15. Naval fuel
16. Miscellaneous
ATR
FRRMTR
Fort St. Vrain
Peach Bottom
FFTF carbide
FFTF oxide
Shippingport
LWBR
TRIGA
EBR-II driver
and blanket,
Fermi-I blanket
Surface ship/
submarine
Not specified
Uranium metal fuel compounds with aluminum or zirconium
alloy cladding
Uranium alloy fuel compounds with zirconium alloy
cladding
Uranium-molybdenum alloy fiiel compounds with zirconium
alloy cladding
Uranium oxide fuel compounds with zirconium alloy or
stainless-steel cladding in fair to good condition
Uranium oxide fuel compounds: (1) without cladding;
(2) clad with zirconium alloy, Hastelloy, nickel-chromium,
or stainless steel in poor or unknown condition; or
(3) nondegraded aluminum clad
Uranium-aluminum alloy fuel compounds with aluminum
cladding
Uranium silicide fuel compounds with aluminum cladding
Thorium/uranium carbide fuel compounds with graphite
cladding in good condition
Thorium/uranium carbide fuel compounds with graphite
cladding in unknown condition
Uranium carbide or plutonium-uranium carbide fuel
compounds with or without stainless-steel cladding
Plutonium/uranium oxide fuel compounds in zirconium
alloy, stainless-steel, or unknown cladding
Uranium/thorium oxide fuel compounds with zirconium
alloy or stainless-steel cladding
Uranium-zirconium hydride fuel compounds with or without
Incalloy, stainless-steel, or aluminum cladding
Uranium and uranium-plutonium metallic alloy with
predominantly stainless-steel cladding
Uranium-based with zirconium alloy cladding
Various fuel compounds with or without zirconium alloy,
aluminum, Hastelloy, tantalum, niobium, stainless-steel or
unknown cladding
a. Source: Fillmore (1998, all).
b. Abbreviations: SNF = spent nuclear fuel; HWCTR = heavy-water cooled test reactor; PWR = pressurized-water reactor;
TMI = Three Mile Island; ATR = Advanced Test Reactor; FRR MTR = foreign research reactor - material test reactor;
FPTF = Fast Flux Test Facility; LWBR = light-water breeder reactor; TRIGA = Training Research Isotopes - General
Atomic; EBR-II = Experimental Breeder Reactor II.
A.2.2.4 Final Spent Nuclear Fuel Form
For all spent nuclear fuel categories except 14, the expected final spent nuclear fuel form does not differ
from the current or planned storage form. Before its disposal in the repository, candidate material would
be in compliance with approved acceptance criteria.
DOE has prepared an EIS at the Savannah River Site (DOE 1998d, all) to evaluate potential treatment
alternatives for spent nuclear fuel and its ultimate disposal in the repository. The products of any
proposed treatment of the Savannah River Site aluminum-based fuels are adequately represented by the
A-23
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-17. National Spent Nuclear Fuel Database projection of DOE spent nuclear fuel locations and
inventories to 2035."''
Fuel category and name
Storage
Site
No. of
units"
Mass
(kilograms)''
Volume
(cubic meters)'
Fissile mass
(kilograms)
Equivalent
uranium mass
(kilograms)
MTHM
1.
Uranium metal'
INEEL
Hanford
SRS
Totals
85
100,000
350
100,435
4,500
2,160,000
120,000
2,284,500
0.7
200
18
218.7
13
25,000
110
25, /25
1,700
2,100,000
17,000
2,118,700
1.7
2100
17
2119
2.
Uraniimi-zirconium
INEEL
69
120
0.7
34
40
0.04
3.
Uranium-molybdenum
INEEL
29,000
4,600
0.3
970
3,800
3.8
4.
Uranium oxide, intact
INEEL
Hanford
Totals
14,000
87
14,087
150,000
44,000
194,000
41
11
52
2,200
240
2,440
80,000
18,000
98,000
80
18
99
5.
Uranium oxide,
failed/declad/aluminum clad
INEEL
Hanford
SRS
Totals
2,000
13
7,600
9,613
340,000
270
58,000
398,270
140
4.2
96
240.2
2,200
4
2,600
4,804
83,000
160
3,200
86.360
84
0.2
3.2
87
6.
Uranium-aluminide
SRS
18,000
130,000
150
6,000
8,800
8.7
7.
Uranium-silicide
SRS
7,400
47,000
53
1,200
12,000
12
8.
Thorium/uranium carbide, high-
integrity
FSV
INEEL
Totals
1,500
1,600
3,100
190,000
130,000
320,000
130
82
2/2
640
350
990
820
440
7,260
15
9.9
25
9.
Thorium/uranium carbide, low-
integrity
INEEL
810
55,000
17
180
210
1.7
10.
Plutonium/uranium carbide,
nongraphite
INEEL
Hanford
Totals
130
2
132
140
330
470
0
0.1
0.1
10
11
21
73
64
137
0.08
0.07
0.2
11.
Mixed oxide
INEEL
Hanford
Totals
2,000
620
2,620
6,100
1 10,000
116.100
2.4
33
35.1
240
2,400
2.640
2,000
8,000
10,000
2.1
10
12
12.
Uranium/thorium oxide
INEEL
260
120,000
18
810
810
50
13.
Uranium-zirconium hydride
INEEL
Hanford
Totals
9,800
190
9,990
33,000
660
33,660
8.1
33
8.3
460
7
467
2,000
36
2.036
2
0.04
2
15.
Naval fuel«*
INEEL
300
4,400,000
888
64,000
65,000
65
16. Miscellaneous
Grand totals
INEEL
Hanford
SRS
Totals
1,500
73
8,800
10,373
210,000
33,000
1,700
9,200
43,900
8,150,000
11
0.2
8.2
19.4
1,900
360
30
550
940
110,000
5,500
130
2,900
8,530
2,420,000
7.7
0.2
2.9
11
2,500
a. Source: Dirkmaat (1998a, all); individual values and totals rounded to two significant figures.
b. Abbreviations: SNF = spent nuclear fuel; INEEL = Idaho National Engineering and Environmental Laboratory; SRS = Savannah
River Site; FSV = Fort St. Vrain.
c. Unit is defined as an assembly, bundle of elements, can of material, etc., depending on the particular spent nuclear fuel category.
d. To convert kilograms to pounds, multiply by 2.2046; to convert metric tons to tons, multiply by 1.1023.
e. To convert cubic meters to cubic yards, multiply by 1.3079.
f. N-Reactor fuel is stored in aluminum or stainless-steel cans at the K-East and K-West Basins. The mass listed in this table does not
include the storage cans.
g. Information supplied by the Navy (Dirkmaat 1997a, Attachment, page 2).
h. A naval fuel unit consists of a naval dual-purpose canister that contains multiple assemblies.
properties of the present aluminum-based fuel (Categories 6, 7, and part of 5) for this Yucca Mountain
EIS. They are bounded by the same total radionuclide inventory, heat generation rates, dissolution rates,
and number of canisters. No additional data about the products will be required to ensure that they are
represented in the EIS inventory.
A-24
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
A.2.2.5 Spent Nuclear Fuel Characteristics
A.2.2.5.1 Mass and Volume
Table A-17 lists total volume, mass, and MTHM for each DOE spent nuclear fuel category from the
National Spent Nuclear Fuel Database (LMIT 1997, all).
A.2.2.5.2 Amount and Nature of Radioactivity
0RIGEN2 (Oak Ridge Isotope Generation), an accepted computer code for calculating spent nuclear fuel
radionuclide inventories, was used to generate activity data for radionuclides in the DOE spent nuclear
fuel inventory. The inventory came from the 1997 version of the National Spent Nuclear Fuel Database
(LMIT 1997, all).
Table A- 18 lists the activities expressed in terms of curies per handling unit for the radionuclides of
interest (uranium, fission products and actinides). The table lists activity estimates decayed to 2030 for
all categories except 15. A handling unit for DOE is a spent nuclear fuel canister, while for Category 15
naval fuels, it is a naval dual-purpose canister.
The activity for naval spent nuclear fuel is provided for typical submarine (15a) or surface ship (15b)
spent nuclear fuels. Dirkmaat (1997a, Attachment, pages 3 to 5) provided these activities for 5 years after
shutdown, which would be the minimum cooling time before naval fuel would reach the repository. The
power history assumed operations at power for a full core life. The assumptions about the power history
and minimum cooling time conservatively bound the activity for naval fuel that would be emplaced in a
monitored geologic repository. In addition, 0RIGEN2 was used to calculate the activity associated with
activation products in the cladding, which are listed in Table A- 18. For completeness, the data also
include the activity that would be present in the activated corrosion products deposited on the fuel.
A.2.2.5.3 Chemical Composition
This section discusses the chemical compositions of each of the 16 categories of DOE spent nuclear fuel
(Dirkmaat 1998a, all).
•
Category 1: Uranium metal. The fuel in this category consists primarily of uranium metal.
N-reactor fuel represents the category because its mass is so large that the performance of the rest of
the fuel in the category, even if greatly different from N-Reactor fuel, would not change the overall
category performance. The fuel is composed of uranium metal about 1.25 percent enriched in
uranium-235, and is clad with a zirconium alloy. Approximately 50 percent of the fuel elements are
believed to have failed cladding. This fuel typically has low bumup. Other contributors to this
category include the Single Pass Reactor fuel at Hanford and declad Experimental Breeder Reactor-II
blanket material at the Savannah River Site.
Category 2: Uranium-zirconium. The fuel in this category consists primarily of a uranium- (91-
percent) zirconium alloy. The Heavy Water Components Test Reactor fuel is the representative fuel
because it is the largest part of the inventory. This fuel is approximately 85-percent enriched in
uranium-235 and is clad with a zirconium alloy.
Category 3: Uranium molybdenum. The fuel in this category consists of uranium- (10 percent)-
molybdenum alloy and 25-percent enriched in uranium-235, and is clad with a zirconium alloy.
Fermi driver core 1 and 2 are the only fuels in the category. The fuel is currently in an aluminum
container. The proposed disposition would include the aluminum container.
A-25
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
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A-26
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
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A-27
•
•
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Category 4: Uranium oxide, intact. The fuel in this category consists of uranium oxide that has
been formed into pellets or plates and clad with a corrosion-resistant material. Commercial fuel is the
representative fuel for this category because it is a large part of the inventory. The fuel is made of
uranium oxide, some of which is highly enriched in uranium-235 and some of which is low enriched
in uranium-235. The fuel elements are clad with a zirconium alloy.
Category 5: Uranium oxide, failed/declad/aluminum clad. The fuel in this category is
chemically similar to the fuels in Category 4, except accident or destructive examination has
disrupted it. The failed fuel from Three Mile Island Reactor 2 represents this category because it
comprises 96 percent of the total MTHM of the category. The Three Mile Island Reactor 2 fuel is
melted uranium oxide. The accident greatly disrupted the cladding. Other fuel in this category is
declad or has a large amount of cladding damage. Approximately 4 percent consists of intact
aluminum clad fuel included in this category because the aluminum cladding is less corrosion
resistant than Category 4 cladding material.
Category 6: Uranium-aluminide. This category consists of fuel with a uranium-aluminum
compound dispersed in a continuous aluminum metal phase. The fuel is clad with an aluminum alloy.
T'lie uranium-235 enrichment varies from 10 to 93 percent.
Category 7: Uranium-silicide. The fuel in this category is a uranium-silicide compound dispersed
in a continuous aluminum metal phase. The fuel is clad with an aluminum alloy. The uranium-235
enrichment varies from 8 to 93 percent, but most are less than 20 percent.
Category 8: Thorium/uranium carbide, high-integrity. This category consists of fuels with
thorium carbide or uranium carbide formed into particles with a high-integrity coating. Fort St. Vrain
Reactor fuel represents the category because it makes up 95 percent of the mass of the category. This
fuel is uranium carbide and thorium carbide formed into particles and coated with layers of pyrolytic
carbon and silicon carbide. The particles are bonded in a carbonaceous matrix material and emplaced
in a graphite block. The fuel was made with uranium enriched to 93 percent in uranium-235. The
thorium was used to generate fissile uranium-233 during irradiation. Some fuel does not have a
silicon carbide coating, but its effect on the category is very small. Less than 1 percent of the fuel
particles are breached.
Category 9: Thorium/uranium carbide, low-integrity. This category consists of fuels with
uranium carbide or thorium carbide made into particles with a coating of an earlier design than that
described for Category 8. Peach Bottom Unit 1, Core 1 is the only fuel in this category. This fuel is
chemically similar to Category 8 fuel except 60 percent of the particle coating is breached. Peach
Bottom Unit 1, Core 2 is included in Category 8 because its fuel particles are basically intact and are
more rugged than the Peach Bottom Unit I, Core 1 particles.
Category 10: Plutonium/uranium carbide, nongraphite. This category consists of fuel that
contains uranium carbide. Much of it also contains plutonium carbide. Fast Flux Test Facility
carbide assemblies represent this category because they make up 70 percent of the category and
contain both uranium and plutonium. The Fast Flux Test Facility carbide fuel was constructed from
uncoated uranium and plutonium carbide spheres that were loaded directly into the fuel pins, or
pressed into pellets that were loaded into the pins. The pins are clad with stainless steel.
Category 1 1: Mixed oxide. This category consists of fuels constructed of both uranium oxide and
plutonium oxide. The Fast Flux Test Facility mixed-oxide test assembly is the representative fuel
because it comprises more than 80 percent of the category. The fuels are a combination of uranium
oxide and plutonium oxide pressed into pellets and clad with stainless steel or a zirconium alloy. The
A-28
•
Inventory and Characteristics of Spent Nuclear Fuel, High-level Radioactive Waste, and Other Materials
uranium-235 enrichment is low, but the fissile contribution of the plutonium raises the effective
enrichment to 15 percent.
Category 12: Uranium/thorium oxide. This category consists of fuels constructed of uranium
oxide and thorium oxide. Shippingport light-water breeder reactor fuel is the representative fuel
because it comprises more than 75 percent of the inventory. The Shippingport light-water breeder
reactor fuel is made of uranium-233, and the irradiation of the thorium produces more uranium-233.
The mixture is pressed into pellets and clad with a zirconium alloy.
Category 13: Urariium-zirconium hydride. This category consists of fuels made of
uranium-zirconium hydride. Training Research Isotopes-General Atomic fuels comprise more than
90 percent of the mass of this category. The fuel is made of uranium-zirconium hydride formed into
rods and clad primarily with stainless steel or aluminum. The uranium is enriched as high as
90 percent in uranium-235, but most is less than 20 percent enriched.
Category 14: Sodium-bonded. For purposes of analysis in this EIS, it is assumed that all
Category 14 fuels would be treated during the proposed electrometallurgical treatment that would
result in high-level radioactive waste. The chemical composition of the resulting high-level
radioactive waste is described in Section A.2.3. Category 14 is included here for completeness.
Category 15: Naval fuel. Naval nuclear fuel is highly robust and designed to operate in a high-
temperature, high-pressure environment for many years. This fuel is highly enriched (93 to 97
percent) in uranium-235. In addition, to ensure that the design will be capable of withstanding battle
shock loads, the naval fuel material is surrounded by large amounts of zirconium alloy (Beckett 1998,
Attachment 2).
DOE plans to emplace approximately 3(X) canisters of naval spent nuclear fuel in the Yucca Mountain
repository. There are several different designs for naval nuclear fuel, but all designs employ similar
materials and mechanical arrangements. The total weight of the fuel assemblies in a canister of a
typical submarine spent reactor fuel, which is representative of the chemical composition of naval
spent nuclear fuel, would be 1 1,000 to 13,000 kilograms (24,000 to 29,000 pounds). Of this total,
less than 500 kilograms (1,100 pounds) would be uranium. Approximately 1,000 to 2,000 kilograms
(2,200 to 4,400 pounds) of the total weight of these fuel assemblies is from hafnium in the poison
devices (primarily control rods) permanently affixed to the fuel assemblies (Beckett 1998,
Attachment 2).
There would be approximately 9,0(X) to 12,0(X) kilograms (20,000 to 26,500 pounds) of zirconium
alloy in the fuel structure in the typical canister. The typical chemical composition of zirconium alloy
is approximately 98 percent zirconium, 1.5 percent tin, 0.2 percent iron, and O.I percent chromium
(Beckett 1998, Attachment 2).
The small remainder of the fuel mass in a typical canister of naval submarine spent nuclear fuel [less
than 500 kilograms (1,100 pounds)] would consist of small amounts of such metals and nonmetals as
fission products and oxides (Beckett 1998, Attachment 2).
Category 16: Miscellaneous. This category consists of the fuels that do not fit into the previous
15 categories. The largest amount of this fuel, as measured in MTHM, is uranium metal or alloy.
The other two primary contributors are uranium alloy and uranium-thorium alloy. These three fuel
types make up more than 80 percent of the MTHM in the category. It is conservative to treat the total
category as uranium metal. Other chemical compounds included in this category include uranium
A-29
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
oxide, uranium nitride, uranium alloys, plutonium oxide, plutonium nitride, plutonium alloys, and
thorium oxide.
Table A- 19 lists the primary materials of construction and chemical composition for each category.
A.2.2.5.4 Thermal Output
Table A-20 lists the maximum heat generation per handling unit for each spent nuclear fuel category
(Dirkmaat 1997a, Attachment, pages 74 to 77; Dirkxnaat 1998b, all). The category 15 (naval fuel)
thermal data used the best estimate radionuclide content from Dirkmaat (1997a, Attachment, pages 74 to
77) at a minimum cooling time of 5 years.
A.2.2.5.5 Quantity of Spent Nuclear Fuel Per Canister
Table A-21 lists the projected number of canisters required for each site and category. The amount of
fuel per canister would vary widely among categories and would depend on a variety of parameters. The
average mass of submarine spent nuclear fuel in a short naval dual-purpose canister would be
approximately 13 metric tons (14 tons) with an associated volume of 2.7 cubic meters (95 cubic feet).
Surface ship spent nuclear fuel in a long naval dual-purpose canister would have an average mass of
approximately 18 metric tons (20 tons) and a volume of 3.5 cubic meters (124 cubic feet) (Dirkmaat
1997a, Attachment, pages 86 to 88).
A.2.2.5.6 Spent Nuclear Fuel Canister Parameters
The Idaho National Engineering and Environmental Laboratory would use a combination of 46- and
61 -centimeter (18- and 24-inch)-diameter stainless-steel canisters for spent nuclear fuel disposition. The
Savannah River Site would use 18-inch canisters, and Hanford would use 64-centimeter (25.3-inch)
multicanister overpacks and 18-inch canisters. Table A-21 lists the specific number of canisters per site.
Detailed canister design specifications for the standard 18- and 24-inch canisters are contained in DOE
(1998c, all). Specifications for the Hanford multicanister overpacks are in Parsons (1999, all).
There are two conceptual dual-purpose canister designs for naval fuel: one with a length of 539
centimeters (212 inches) and one with a length of 475 centimeters (187 inches). Both canisters would
have a maximum diameter of 169 centimeters (67 inches) (Dirkmaat 1997a, Attachment, pages 86 to 88).
Table A-22 summarizes the preliminary design information.
For both designs, the shield plug, shear ring, and outer seal plate would be welded to the canister shell
after the fuel baskets were loaded in the canister. The shield plug, shear ring, and welds, along with the
canister shell and bottom plug, would form the containment boundary for the disposable container. The
shell, inner cover, and outer cover material for the two canisters would be low-carbon austenitic stainless
steel or stabilized austenitic stainless steel. Shield plug material for either canister would be stainless
steel or another high-density material sheathed in stainless steel (Dirkmaat 1997a, Attachment, pages 86
to 88).
A.2.3 HIGH-LEVEL RADIOACTIVE WASTE
High-level radioactive waste is the highly radioactive material resulting from the reprocessing of spent
nuclear fuel. DOE stores high-level radioactive waste at the Hanford Site, the Savannah River Site, and
the Idaho National Engineering and Environmental Laboratory. Between 1966 and 1972, commercial
chemical reprocessing operations at the Nuclear Fuel Services plant near West Valley, New York,
generated a small amount of high-level radioactive waste at a site presently owned by the New York State
A-30
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
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A-31
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-20. Maximum heat generation for DOE spent nuclear fuel
(watts per handling unit)."'*'
Maximum heat
Category and fuel type
generation
1.
Uranium metal
18
2.
Uranium zirconium
90
3.
Uranium molybdenum
4
4.
Intact uranium oxide
1,000
5.
Failed/declad/aluminum clad uranium oxide
800
6.
Uranium aluminide
480
7.
Uranium silicide
1,400
8.
High-integrity thorium/uranium carbide
250
9.
Low-integrity thorium/uranium carbide
37
10.
Nongraphite plutonium/uranium carbide
1,800
11.
Mixed oxide
1,800
12.
Thorium/uranium oxide
120
13.
Uranium zirconium hydride
100
14.
Sodium-bonded
N/A'
15.
Naval fuel
4,250
16.
Miscellaneous
1,000
a.
Sources: Dirkmaat (1997a, Attachment, pages 74 to 77; Dirkmaat 1998b, all).
b.
Handling unit is a canister or naval dual purpose canister.
c.
N/A = not applicable. Assumed to be treated and therefore part of high-level
radioactive waste inventory (see Section A. 2. 2.1).
Table A-21.
Required number of canisters for disposal of DOE
spent nuclear fuel."''
Hanford INEEL
SRS
Naval
Category
18-inch
25.3-inch 18-inch 24-inch
18-inch
Short DPC" Long DPC
1
440 6
9
2
8
3
70
4
14
20 179 16
5
1
406
425
6
750
7
225
8
503"
9
60
10
2
3
11
324
43
12
24 47
13
3
97
14'
15
200 100
16
5
39
2
Totals
349
460 1,438 63
1,411
200 100
a. Sources: Dirkmaat (1997b, Attachment, page 2); Dirkmaat (1998a, all).
b. INEEL = Idaho National Engineering and Environmental Laboratory; SRS = Savannah River Site.
c. Naval dual-purpose canister.
d. Includes 334 canisters from Fort St. Vrain.
e. Assumed to be treated and therefore part of high-level radioactive waste inventory (see Section A.2.2.1).
Energy Research and Development Authority. These operations ceased after 1972. In 1980, Congress
passed the West Valley Demonstration Project Act, which authorizes DOE to conduct, with the Research
and Development Authority, a demonstration of solidification of high-level radioactive waste for disposal
and the decontamination and decommissioning of demonstration facilities(DOE 1992, Chapter 3). This
A-32
169
169
475
539
27
27
45
45
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-22. Preliminary naval dual-purpose canister design parameters/
Parameter Short canister Long canister
Maximum outside diameter (centimeters)'"''^
Maximum outer length (centimeters)
Minimum loaded weight (metric tons)''
Maximum loaded weight (metric tons)
a. Source: Dirkmaat (1997a, Attachment, pages 86 to 88).
b. To convert centimeters to inches, multiply by 0.3937.
c. Right circular cyhnder.
d. To convert metric tons to tons, multiply by 1.1023.
section addresses defense high-level radioactive waste generated at the DOE sites (Hanford Site, Idaho
National Engineering and Environmental Laboratory, and Savannah River Site) and commercial high-
level radioactive waste generated at the West Valley Demonstration Project.
A.2.3.1 Background
In 1985, DOE published a report in response to Section 8 of the Nuclear Waste Policy Act (of 1982) that
required the Secretary of Energy to recommend to the President whether defense high-level radioactive
waste should be disposed of in a geologic repository along with commercial spent nuclear fuel. That
report. An Evaluation of Commercial Repository Capacity for the Disposal of Defense High-Level Waste
(DOE 1985, all), provided the basis, in part, for the President's determination that defense high-level
radioactive waste should be disposed of in a geologic repository. Given that determination, DOE decided
to allocate 10 percent of the capacity of the first repository for the disposal of DOE spent nuclear fuel
(2,333 MTHM) and high-level radioactive waste (4,667 MTHM) (Dreyfuss 1995, all; Lytle 1995, all).
Calculating the MTHM quantity for spent nuclear fuel is straightforward. It is determined by the actual
heavy metal content of the spent fuel. However, an equivalence method for determining the MTHM in
defense high-level radioactive waste is necessary because almost all of its heavy metal has been removed.
A number of alternative methods for determining MTHM equivalence for high-level radioactive waste
have been considered over the years. Foiu- of those methods are described in the following paragraphs.
Historical Method. Table 1-1 of the 1985 DOE report provided a method to estimate the MTHM
equivalence for high-level radioactive waste based on comparing the radioactive (curie) equivalence of
commercial high-level radioactive waste and defense high-level radioactive waste. The method relies on
the relative curie content of a hypothetical (in the early 1980s) canister of defense high-level radioactive
waste from the Savannah River, Hanford, or Idaho site, and a hypothetical canister of vitrified waste from
reprocessing of high-bumup commercial spent nuclear fuel. Based on commercial high-level radioactive
waste containing 2.3 MTHM per canister (heavy metal has not been removed from commercial waste)
and defense high-level radioactive waste estimated to contain approximately 22 percent of the
radioactivity of a canister of commercial high-level radioactive waste, defense high-level radioactive
waste was estimated to contain the equivalent of 0.5 MTHM per canister. Since 1985, DOE has used this
0.5 MTHM equivalence per canister of defense high-level radioactive waste in its consideration of the
potential impacts of the disposal of defense high-level radioactive waste, including the analysis presented
in this EIS. With this method, less than 50 percent of the total inventory of high-level radioactive waste
could be disposed of in the repository within the 4,667 MTHM allocation for high-level radioactive
waste. There has been no determination of which waste would be shipped to the repository, or the order
of shipments.
Spent Nuclear Fuel Reprocessed Method. Another method of determining MTHM equivalence,
based on the quantity of spent nuclear fuel reprocessed, would be to consider the MTHM in the high-level
radioactive waste to be the same as the MTHM in the spent nuclear fuel before it was reprocessed. Using
A-33
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
this method, less than 5 percent of the total inventory of high-level radioactive waste could be disposed of
in the repository within the 4,667 MTHM allocation for high-level radioactive waste.
Total Radioactivity Mettiod. Another method, the total radioactivity method, would establish
equivalence based on a comparison of radioactivity inventory (curies) of defense high-level radioactive
waste to that of a standard MTHM of commercial spent nuclear fuel. For this equivalence method the
standard spent nuclear fuel characteristics are based on pressurized-water reactor fuel with uranium-235
enrichment of 3. 11 percent and 39.65 gigawatt-days per MTHM bumup. Using this method, 100 percent
of the total inventory of high-level radioactive waste inventory could be disposed of in the repository
within the 4,667 MTHM allocation for high-level radioactive waste.
Radiotoxicity Method. Yet another method, the radiotoxicity method, uses a comparison of the relative
radiotoxicity of defense high-level radioactive waste to that of a standard MTHM of commercial spent
nuclear fuel, and is thus considered an extension of the total radioactivity method. Radiotoxicity
compares the inventory of specific radionuclides to a regulatory release limit for that radionuclide, and
uses these relationships to develop an overall radiotoxicity index. For this equivalence, the standard spent
nuclear fuel characteristics are based on pressurized-water reactor fuel with uranium-235 enrichment of
3.11 percent, 39.65 gigawatt-days per MTHM bumup. Using this method, 100 percent of the total
inventory of high-level radioactive waste could be disposed of in the repository within the 4,667 MTHM
allocation for high-level radioactive waste.
A recent report (Knecht et al. 1999, all) describes four equivalence calculation methods and notes that,
under the Total Radioactivity Method or the Radiotoxicity Method, all DOE high-level radioactive waste
could be disposed of under the Proposed Action. Using different equivalence methods would shift the
proportion of high-level radioactive waste that could be disposed of between the Proposed Action and
Inventory Module 1 analyzed in Chapter 8, but would not change the cumulative impacts analyzed in this
EIS. Regardless of the equivalence method used, the EIS analyzes the impacts from disposal of the entire
inventory of high-level radioactive waste in inventory Module 1.
A.2.3.2 Sources
A.2.3.2.1 HanfordSite
The Hanford high-level radioactive waste materials discussed in this EIS are those in the Tank Waste
Remediation System Disposal Program and include tank waste, strontium capsules, and cesium capsules
(Picha 1997, Table RL-1). DOE has not declared other miscellaneous materials or waste at Hanford,
either existing or forecasted, to be candidate high-level radioactive waste streams. Before shipment to the
repository, DOE would vitrify the high-level radioactive waste into a borosilicate glass matrix and pour it
into stainless-steel canisters.
A.2.3.2.2 Idaho National Engineering and Environmental Laboratory
The Idaho National Engineering and Environmental Laboratory has proposed three different high-level
radioactive waste stream matrices for disposal at the proposed Yucca Mountain Repository — glass,
ceramic, and metal. The glass matrix waste stream would come from the Idaho Nuclear Technology and
Engineering Center and would consist of wastes generated from the treatment of irradiated nuclear fuels.
The Argonne National Laboratory-West proposed electrometallurgical treatment of DOE sodium-bonded
fuels would generate both ceramic and metallic high-level radioactive waste matrices. DOE is preparing
an EIS [DOE/EIS-0287 (Notice of Intent, 62 FR 49209, September 19, 1997)] to support decisions on
managing the high-level radioactive waste at the Idaho Nuclear Technology and Engineering Center.
DOE is preparing a separate EIS on managing sodium-bonded spent nuclear fuel at Argonne National
A-34
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Laboratory-West and elsewhere, under which electrometallurgical treatment as well as alternative
terminologies are being considered [DOEyEIS-0306 (Notice of Intent, 64 FR 8553, February 22, 1999)].
A.2.3.2.3 Savannah River Site
Savannah River Site high-level radioactive waste consists of wastes generated from the treatment of
irradiated nuclear fuels. These wastes include various chemicals, radionuclides, and fission products that
DOE maintains in liquid, sludge, and saltcake forms. The Defense Waste Processing Facility at the
Savannah River Site mixes the high-level radioactive waste with glass-forming materials, converts it to a
durable borosilicate glass waste form, pours it into stainless-steel canisters, and seals the canisters with
welded closure plugs (Picha 1997, Attachment 4, page 2).
Another source of high-level radioactive waste at the Savannah River Site is the immobilized plutonium
addressed in Section A.2.4.
A.2.3.2.4 West Valley Demonstration Project
The West Valley Demonstration Project is responsible for solidifying high-level radioactive waste that
remains from the commercial spent nuclear fuel reprocessing plant operated by Nuclear Fuel Services.
The Project mixes the high-level radioactive waste with glass-forming materials, converts it to a durable
borosilicate glass waste form, pours it into stainless-steel canisters, and seals the canisters with welded
closure plugs.
A.2.3.3 Present Status
A.2.3.3.1 HanfordSite
The Hanford Site stores high-level radioactive waste in underground carbon-steel tanks. This analysis
assumed that before vitrification, strontium and cesium capsules currently stored in water basins at
Hanford would be blended with the liquid high-level radioactive waste. To date, Hanford has
immobilized no high-level radioactive waste. Before shipping waste to a repository, DOE would vitrify it
into an acceptable glass form. DOE has scheduled vitrification to begin in 2007 with an estimated
completion in 2028.
A.2.3.3.2 Idaho National Engineering and Environmental Laboratory
Most of the high-level radioactive waste at the Idaho Nuclear Technology and Engineering Center
(formerly the Idaho Chemical Processing Plant) is in calcined solids (calcine) stored at the Idaho National
Engineering and Environmental Laboratory. The calcine, an interim waste form, is in stainless-steel bins
in concrete vaults. Before shipment to a repository, DOE proposes to immobilize the high-level
radioactive waste in a vitrified (glass) waste form. The Idaho Nuclear Technology and Engineering
Center proposes to implement its vitrification program in 2020 and complete it in 2035 (LMIT 1998,
pages A-39 to A-42).
As discussed in Section A.2.2.1, DOE is evaluating treatment of sodium-bonded fuels at Argonne
National Laboratory-West. If electrometallurgical treatment were to be chosen, DOE would stabilize the
high-level radioactive waste generated from the treatment of its sodium-bonded fuel in the Fuel
Conditioning Facility and Hot Fuel Examination Facility into ceramic and metal waste forms in the same
facilities. The Radioactive Scrap and Waste Facility at Argonne National Laboratory-West would
provide interim storage for these waste forms. There are several technologies being considered for waste
A-35
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
treatment (for example, electrometallurgical treatment, melt and dilute, Purex). If a decision was made to
implement this proposal, DOE would begin stabilization in 2000.
A.2.3.3.3 Savannah River Site
DOE stores high-level radioactive waste in underground tanks in the F- and H- Areas at the Savannah
River Site. High-level radioactive waste that has been converted to a borosilicate glass form is stored in
the Glass Waste Storage Building in the S-Area. DOE projects completion of the vitrification of the
stored high-level radioactive waste by 2022 (Davis and Wells 1997, all).
A.2.3.3.4 West Valley Demonstration Project
High-level radioactive waste is stored in underground tanks at the West Valley site. High-level
radioactive waste that has been converted into a borosilicate glass waste form is stored in the converted
Chemical Process Cell in the Process Building, referred to as the hiterim High-Level Radioactive Waste
Storage Facility. West Valley plans to complete its vitrification program by the Fall of 2(X)2 (DOE 1992,
Chapter 3).
A.2.3.4 Final Waste Form
The final waste form for high-level radioactive waste from the Hanford Site, Savannah River Site, Idaho
Nuclear Technology and Engineering Center, and West Valley Demonstration Project would be a vitrified
glass matrix in a stainless-steel canister.
The waste forms from Argonne National Laboratory-West could be ceramic and metallic waste matrices
depending on decisions to be based on an ongoing EIS. These could be in stainless-steel canisters similar
to those used for Savannah River Site and Idaho Nuclear Technology and Engineering Center glass
wastes.
A.2.3.5 Waste Characteristics
A.2.3.5.1 Mass and Volume
Hanford Site. The estimated volume of borosilicate glass generated by high-level radioactive waste
disposal actions at Hanford will be 15,700 cubic meters (554,000 cubic feet); the estimated mass of the
glass is 44,000 metric tons (48,500 tons) (Picha 1998a, Attachment 1). The volume calculation assumes
that strontium and cesium compounds from capsules currently stored in water basins would be blended
with tank wastes before vitrification with no increase in product volume. This volume of glass would
require 14,500 canisters, nominally 4.5 meters (15 feet) long with a 0.61-meter (2-foot) diameter (Picha
1998a, Attachment 1).
Idaho National Engineering and Environmental Laboratory. Table A-23 lists the volumes, masses,
densities, and estimated number of canisters for the three proposed waste streams.
Savannah River Site. Based on Revision 8 of the High-Level Waste System Plan (Davis and Wells
1997, all), the Savannah River Site would generate an estimated 5,978 canisters of high-level radioactive
waste (Picha 1997, Attachment 1). The canisters have a nominal outside diameter of 0.61 meter (2 feet)
and a nominal height of 3 meters (10 feet). They would contain a total of approximately 4,240 cubic
meters (150,(X)0 cubic feet) of glass. The estimated total mass of high-level radioactive waste for
repository disposal would be 11, 600 metric tons (12,800 tons) (Picha 1997, Attachment 1). Section
A.2.4.5.2.1 addresses the additional high-level radioactive waste canisters that DOE would generate at the
A-36
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-23. Physical characteristics of high-level radioactive waste at the Idaho National Engineering
and Environmental Laboratory."''
Physical quantities
INTEC glass matrix
ANL-W ceramic matrix
ANL-W metal matrix
Volume (cubic meters)"
743
60.0
1.2
Mass (kilograms)''
1,860,000
144,000
9,000
Density (kilograms per cubic meter)
2,500
2,400
7,750
Number of canisters [range]'
1,190
96 [80 - 125]
6 [2 - 10]
a. Sources: Picha (1997, Attachment 1); Goff (1998a, all); Goff (1998b, all).
b. INTEC = Idaho Nuclear Technology and Engineering Center; ANL-W = Argonne National Laboratory- West.
c. To convert cubic meters to cubic yards, multiply by 1 .3079.
d. To convert kilograms to pounds, multiply by 2.2046.
e. Canister would be nominally 3 meters (10 feet) by 0.6 meter (2 feet). Canisters would be filled to approximately 0.625
cubic meter (22 cubic feet).
Savannah River Site as a result of immobilizing surplus plutonium. As discussed in that section,
77 additional canisters would be required if the assumed 18 metric tons (20 tons) of plutonium is
iinmobilized. If the entire 50 metric tons (55 tons) of surplus plutonium was immobilized, 210
additional high-level radioactive waste canisters would be required.
West Valley Demonstration Project. The West Valley Demonstration Project will generate between
260 and 300 canisters of high-level radioactive waste. The canisters have a nominal outside diameter of
0.61 meter (2 feet) and a nominal height of 3 meters (10 feet) (Picha 1997, Attachment 1). They will
contain approximately 200 cubic meters (7,060 cubic feet) of glass. The estimated total mass of this high-
level radioactive waste will be between 540 and 630 metric tons (595 and 694 tons) (Picha 1998c, page
3).
Summary. Table A-24 summarizes the information in the previous paragraphs to provide the total mass
and volume projected to be disposed of at the repository.
Table A-24. High-level radioactive waste mass and volume summary.
Parameter Total^
Mass 58,000 metric tons
Volume 21,000 cubic meters
Number of canisters 22,147-22,280'
a. Sources: Picha (1997, Attachment 1); Picha (1998a, Attachment 1).
b. To convert metric tons to tons, multiply by 1 .1023; to convert cubic meters to cubic
yards, multiply by 1 .3079.
c. The number of canisters depends on the amount of surplus weapons-usable
plutonium immobilized (see Section A.2.4.5.2.I).
A.2.3.5.2 Amount and Nature of Radioactivity
The following paragraphs present radionuclide inventory information for the individual sites. They
present the best available data at varying dates; however, in most cases, the data are conservative because
the inventories are for dates earlier than the date of disposal, and additional radioactive decay would
occur before disposal. Any differences due to varying amounts of radioactive decay are small.
Hanford Site. Table A-25 lists the estimated radionuclide inventory for Hanford high-level radioactive
glass waste, including strontium-90 and cesium-137 currently stored in capsules (Picha 1997, Table
RL-1). With the exception of hydrogen-3 and carbon-14, this table makes the conservative assumption
that 100 percent of a radionuclide in Hanford's 177 tanks and existing capsules is vitrified. Consistent
with Hanford modeling for the Integrated Data Base (DOE 1997b, page 2-24), pretreatment and
vitrification would separate hydrogen-3 and carbon-14 from the high-level radioactive waste stream such
A-37
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-25. Radionuclide distribution for Hanford Site high-level radioactive waste."''
Curies per
Curies per
Radionuclide
Total curies
canister
Radionuclide
Total curies
canister
Hydrogen-3
c
—
Thorium-229
1.8
1.3x10""
Carbon- 14
9.6x10"^
6.6x10"*
Thorium-230
—
—
Chlorine-36
—
—
Thorium-232
2.1
1.5x10""
Nickel-59
9.3x10'
6.4x10"'
Protactinium-23 1
1.6x10'
1.1x10"'
Nickel-63
9.2x10^
6.3
Uranium-232
1.2x10'
8.5x10"'
Cobalt-60
1.2x10"
8.5x10"'
Uranium-233
4.8x10'
3.3x10"'
Selenium-79
7.7x10'
5.3x10"'
Uranium-234
3.5x10'
2.4x10"'
Krypton-85
—
—
Uranium-235
1.5x10'
1.0x10"'
Strontium-90
9.7x10^
6.7x10^
Uranium-236
9.6
6.6x10"
Niobium-93m
2.7x10^
1.9x10"'
Uranium-238
3.2x10'
2.2x10"'
Niobium-94
—
—
Neptunium-237
1.4x10'
9.7x10"'
Zirconium-93
3.6x10^
2.5x10'
Plutonium-238
2.8x10'
1.9x10"'
Technetium-99
3.3x10*
2.3
Plutonium-239
3.9x10"
2.7
Rhodium- 101
—
~
Plutonium-240
8.9x10'
6.2x10"'
Rhodium- 102
~
~
Plutonium-241
2.3x10^
1.6x10'
Ruthenium- 106
l.OxlO'
7.2
Plutonium-242
1.2
8.0x10"'
Palladium- 107
—
—
Americium-241
7.0x10"
4.8
Tin- 126
1.2x10^
8.2x10"'
Americium-242m
—
—
Iodine- 129
3.2x10'
2.2x10"'
Americium-243
9.3
6.4x10""
Cesium- 134
8.9x10"
6.1
Curium-242
7.7x10'
5.3x10"'
Cesium- 135
—
—
Curium-243
l.OxlO'
6.9x10""
Cesium- 137
1.1x10*
7.7x10'
Curium-244
2.4x10'
1.7x10"'
Samarium- 151
2.8x10*
1.9x10'
Curium-245
~
—
Lead-210
—
—
Curium-246
—
—
Radium-226
6.3x10"'
4.4x10"*
Curium-247
~
—
Radium-228
7.7x10'
5.3x10"'
Curium-248
~
—
Actinium-227
8.8x10'
6.0x10"'
Califomium-252
-
-
a. Sources: Picha (1997, Table RL-l);Picha (1998a, Attachment 1).
b. Decayed to January 1, 1994.
c. — = not found in appreciable quantities.
that essentially 0.0 percent and 0.002 percent of each, respectively, would be present in the glass. A large
portion of iodine- 129 could also be separated, but the analysis assumed a conservative 50-percent
retention (Picha 1998a, Attachment 1). Table A-25 uses the estimated number of canisters (14,500) to
develop the curies-per-canister value.
Idaho National Engineering and Environmental Laboratory. Table A-26 contains a baseline
radionuclide distribution for the three Idaho National Engineering and Environmental Laboratory high-
level radioactive waste streams. For each waste stream, the total radionuclide inventory is provided, as is
the worst-case value for curies per canister. For Idaho Nuclear Technology and Engineering Center glass,
the calculated inventories are decayed to 2035. For Argonne National Laboratory -West waste matrices,
the calculated inventories are decayed to 2000.
Savannah River Site. The Waste Qualification Report details the projected radionuclide distribution in
the high-level radioactive waste from the Savannah River Site (Plodinec and Marra 1994, page 10). Table
A-27 lists the quantities of individual radionuclides in 2015, the expected time of shipment (Pearson
1998, all). The curie-per-canister values were obtained by dividing the total radionuclide projection by
the expected number of canisters (5,978).
West Valley Demonstration Project. DOE used the 0RIGEN2 computer code to estimate the
radionuclide inventory for the West Valley Demonstration Project, simulating each Nuclear Fuel Services
A-38
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-26. Radionuclide distribution for Idaho National Engineering and Environmental Laboratory
high-level radioactive waste/''
INTEC
glass
ANL-W
ceramic'^
ANL-W metal'^
Total curies
Curies per
Total curies for
Curies per
Total curies
Radionuclides
for 2035
canister''
2000
canister""
for 2000
Curies per canister''
Hydrogen-3
3.6x10'
4.3
.."
—
—
—
Carbon- 14
2.8x10-^
8.3x10'
—
—
4.3
4.3
Chlorine-36
-
—
—
—
-
—
Cobalt-60
3.2x10'
3.6x10"'
—
_
3.2x10'
3.2x10'
Nickel- 59
—
—
—
—
1.1x10'
l.lxlO'
Nickel-63
~
—
—
—
4.1x10'
3.9x10'
Selenium-79
—
-
—
—
—
—
Krypton-85
-
-
—
-
—
-
Strontium-90
7.0x10'
1.2x10*
7.1x10'
4.7x10*
-
~
Niobium-93
4.7x10^
1.4
—
-
2.9x10'
2.9x10'
Niobiuni-94
5.4x10'
1.6x10'
—
_
2.7
2.7
Zircomum-93
—
—
—
_
—
—
Technetium-99
3.4x10'
9.9
~
_
1.3x10'
1.3x10'
Rhodium- 101
~
-
—
—
—
-
Rhodium- 102
2.0x10'
2.2x10'
—
—
-
-
Ruthenium- 106
l.OxlO'
8.7x10"
—
_
2.1x10*
2.1x10*
Palladium- 107
—
—
—
—
-
-
Tin- 126
8.9x10'
2.6x10'
—
—
2.8
2.1
Iodine-129
5.6
1.7x10"'
3.4x10'
1.8x10"'
—
_
Cesium- 134
3.3x10-^
3.6x10'
7.9x10'
5.1x10'
—
—
Cesium- 135
1.6x10^
2.5x10"'
1.6x10'
8.8x10"'
_
—
Cesium- 137
6.0x10'
1.2x10*
8.5x10'
5.3x104
_
—
Samarium- 151
—
-
—
-
—
—
I^ad-210
—
-
-
-
—
--
Radium-226
9.7x10'
7.2x10"'
3.0x10"'
2.1x10"*
—
—
Radium-228
—
—
—
—
_
—
Actinium-227
—
—
—
—
—
—
Thorium-229
—
—
—
—
_
—
Thorium-230
4.0x10'
2.8x10"'
4.7x10"'
8.9x10^
—
—
Thorium-232
9.9x10'
S.OxlO"'"
2.3x10"'
1.3x10""
_
—
Protactinium-231
-
-
-
—
—
—
Uranium-232
4.6x10'
5.2x10"*
2.6x10"'
1.8x10"*
1.2x10^
1.2x10"*
Uranium-233
1.3x10'
6.1x10"*
2.0x10^
1.4x10"'
5.8x10"'
5.8x10"'
Uranium-234
1.0x10^
1.1x10"'
2.8
1.9x10'
7.7x10"'
7.7x10"'
Uranium-235
5.9x10'
6.6x10^
8.8x10'
5.9x10"'
2.5x10"'
2.5x10"'
Uranium-236
1.5
1.7x10"'
6.3x10"'
4.2x10"'
1.8x10"'
1.8x10"'
Uranium-238
2.9x10-^
3.3x10"'
2.8x10'
4.9x10"'
9.7x10"'
8.8x10"'
Neptunium-237
6.3
2.8x10"'
1.3
5.8x10"'
2.4x10"'
2.3x10"'
Plutonium-238
9.0x10*
l.Oxltf
3.6x10'
2.9x10'
6.6x10'
6.6x10'
Plutonium-239
1.8x10'
2.0
1.7x10*
8.1x10'
3.3x10'
3.3x10"'
Plutonium-240
1.6x10'
1.8
1.5x10'
6.9x10'
2.9x10"'
2.9x10'
Plutomum-241
1.9x10*
2.2x10'
1.1x10*
1.3x10'
1.9x10-'
1.9x10'
Plutomum-242
3.4
3.8x10"'
1.2x10"'
2.3x10"'
2.0x10"*
2.0x10^
Americium-241
1.3x10*
1.4x10'
1.6x10'
3.4x10'
3.1x10"'
2.1x10'
Americium-242/242m
1,5x10-^
9.4x10"'
1.4x10'
2.1x10"'
2.7x10"*
2.1x10"*
Americium-243
1.4x10"^
l.IxlO^
2.8x10"'
1.9x10'
4.8x10^
4.8x10^
Curium-242
1.2x10-^
7.7x10"'
1.2x10'
1.8x10'
2.3x10-*
1.8x10^
Curium-243
4.7x10^
3.4x10"*
1.6x10"'
3.1x10"'
3.0x10^
2.1x10^
Curium-244
1.0x10^
7.7x10"'
1.9
1.3x10'
3.1x10"'
3.1x10'
Curium-245
3.7x10-*
2.8x10"'
6.8x10"'
4.7x10'
1.1x10"'
l.lxlO'
Curium-246
8.7x10'
6.6x10""'
4.2x10"'
2.9x10'
7.1x10"
7.1x10"
Curium-247
3.1x10'*
2.4x10"
2.4x10"
1.6x10'*
4.0x10"
4.0x10"
Curium-248
9.4x10"
7.2x10""
2.6x10"'*
1.8x10"
4.4x10"
4.4x10""
Califoniium-252
-
-
6.5x10"
1.6x10"
-
-
a. Sources: Picha ( 1 997, Table ID-2); Goff ( 1 998a, all).
b. INTEC = Idaho Nuclear Technology and Engineering Center; ANL-W = Argonne National Laboratory- West.
c. Matrices based on treating all sodium-bonded fuels. Waste input streams and associated radioactivity for 2000 averaged for total number of
canisters produced. Curie values based on calculated data from stored material.
d. Curie per canister values were provided as worst case rather than a homogenous mixture.
e. — = not found in appreciable quantities.
A-39
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-27. Radionuclide distribution for Savannah River Site high-
•level radioactive waste (2015)."
Total
Curies per
Total
Radionuclide
(curies)
canister
Radionuclide
(curies)
Curies per canister
Hydrogen-3
__b
—
Thorium-229
—
—
Carbon- 14
~
—
Thorium-230
2.4x10"^
4.0x10"*
Chlorine-36
—
—
Thorium-232
—
—
Nickel-59
1.1x10^
1.8x10"^
Protactinium-23 1
—
—
Nickel-63
1.2x10'
2.1
Uranium-232
. ~
—
Cobalt-60'
—
4.5x10'
Uranium-233
~
—
Selenium-79
1.1x10^
1.8x10"'
Uranium-234
1.6x10^
2.7x10"^
Krypton-85
—
—
Uranium-235
—
—
Strontium-90
1.7x10*
2.9x10"
Uranium-236
—
—
Niobium-93m
1.3x10"
2.2
Uranium-238
5.0x10'
8.3x10"'
Niobium-94
—
—
Neptunium-237
4.1x10^
6.8x10"^
Zirconium-93
3.0x10"
5.0
Plutonium-238
3.0x10*
5.0x10^
Technetium-99
1.5x10"
2.5
Plutonium-239
3.7x10"
6.2
Rhodium- 101
~
~
Plutonium-240-
2.5x10"
4.1
Rhodium- 102
—
—
Plutonium- 241
3.3x10*
5.4x10^
Ruthenium- 106"
—
2.4
Plutonium-242
3.5x10'
■ 5.8x10"'
Palladium- 107
7.3x10'
1.2x10"^
Americium-241
1.6x10^
2.6x10'
Tin- 126
2.6x10^
4.3x10"'
Americium-242m
—
—
Iodine- 129
—
—
Americium-243
1.1x10^
1.8x10'
Cesium- 134"
—
1.2x10'
Curium-242
—
—
Cesium- 135
4.0x10^
6.7x10"^
Curium-243
—
—
Cesium- 137
1.5x10*
2.4x10"
Curium-244
4.9x10'
8.3x10'
Samarium- 151
3.3x10^
5.5x10^
Curium-245
—
~
Lead-210
~
—
Curium-246
—
—
Radium-226
—
~
Curium-247
~
—
Radium-228
—
—
Curium-248
~
—
Actinium-227
-
-
Californium-252
-
-
a. Sources: Plodinec and Marra (1994, page 10); Pearson (1998, all).
b. - = not found in appreciable quantities. -j
c. Total curie content not provided for these nuclides; curie per canister values provided for 10 years after production.
irradiated fuel campaign. A detailed description of the development of these estimates is in the West
Valley Demonstration Project Waste Qualification Report (WVNS 1996, WQR-1.2, Appendix 1). Table
A-28 lists the estimated activity by nuclide and provides the total curies, as well as the curies per canister,
based on 260 canisters.
A.2.3.5.3 Chemical Composition
Hanford Site. The Integrated Data Base (DOE 1997b, page 2-29) provides the best available
information for the proposed representative chemical composition of future high-level radioactive waste
glass from Hanford. Table A-29 combines the percentages by weight of chemical constituents obtained
from the hitegrated Data Base with the estimated mass to present the expected chemical composition of
the glass in terms of mass per chemical compound.
Idaho National Engineering and Environmental Laboratory
Idaho Nuclear Technology and Engineering Center Glass Matrix. This waste stream is composed
of three primary sources — zirconium calcine, aluminum calcine, and sodium-bearing waste.
The distribution of these sources is 55 percent, 15 percent, and 30 percent, respectively (Heiser 1998, all).
Table A-30 lists the chemical composition of the total waste stream.
A-40
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-28. Radionuclide distribution for West Valley Demonstration Project high-level radioactive
waste (2015)."
Curies per
Curies per
Radionuclide
Total curies
canister
Radionuclide
Total curies
canister
Hydrogen-3
2.0x10'
7.8x10-^
Thorium-229
2.3x10"'
8.9x10*
Carbon- 14
1.4x10^
5.3x10'
Thorium-230
6.0x10"^
2.3x10"*
Chlorine-36
__b
—
Thorium-232
1.6
6.3x10"'
Nickel-59
1.1x10^
4.1x10'
Protactinium-23 1
1.5x10'
5.9x10"^
Nickel-63
7.1x10'
2.7x10'
Uranium-232
5.9
2.3x10"^
Cobalt-60
2.9x10'
1.1x10"'
Uranium-233
9.5
3.7x10"^
Selenium-79
6.0x10'
2.3x10'
Uranium-234
5.0
1.9x10"^
Krypton-85
—
—
Uranium-235
1.0x10"'
3.9x10"*
Strontium-90
3.7x10'
1.4x10*
Uranium-236
3.0x10"'
■ 1.1x10"'
Niobium-93m
2.5x10^
9.5x10'
Uranium-238
8.5x10'
3.3x10"'
Niobium-94
—
—
Neptunium-237
2.4x10'
9.2x10"^
Zirconium-93
2.7x10^
1.1
Plutonium-238
7.0x10'
2.7x10'
Technetiuni-99
1.7x10^
6.5
Plutonium-239
1.7x10'
6.4
Rhodium-101
—
—
Plutonium-240
1.2x10'
4.7
Rhodium- 102
—
—
Plutonium-241
2.5x10*
9.5x10'
Ruthenium- 106
5.0x10"^
1.9x10"'
Plutonium-242
1.7
6.4x10"'
Palladium- 107
l.lxlO'
4.2x10"^
Americium-241
5.3x10*
2.0x10^
Tin- 126
1.0x10^
4.0x10"'
Americium-242m
2.7x10^
1.0
Iodine- 129
2.1x10'
8.1x10"*
Americium-243
3.5x10^
1.3
Cesium- 134
1.2
4.4x10"'
Curium-242
2.2x10^
8.4x10"'
Cesium- 135
1.6x10^
6.2x10"'
Curium-243
7.3x10'
2.8x10'
Cesium- 137
4.1x10*
1.6x10*
Curium-244
2.9x10'
l.lxio'
Samarium- 151
7.0x10'
2.7x10^
Curium-245
8.8x10"'
3.4x10"'
Lead-210
—
—
Curium-246
1.0x10"'
3.9x10"*
Radium-226
~
—
Curium-247
~
—
Radium-228
1.6
6.3x10"'
Curium-248
~
—
Actinium-227
1.2x10'
4.6x10"^
Califomium-252
—
—
a. Source: WVNS (1996, WQR-1.2, Appendix 1).
b. - = not found in appreciable quantities.
Table A-29. Expected chemical composition of Hanford high-level radioactive
waste
glass (kilograms).^'
Compound
Mass
Compound
Mass
Aluminum oxide
4,100,000
Sodium oxide
5,190,000
Boron oxide
3,090,000
Sodium sulfate
44,000
Bismuth trioxide
510,000
Nickel monoxide
480,000
Calcium oxide
370,000
Phosphorous pentaoxide
690,000
Ceric oxide
500,000
Lead monoxide
62,000
Chromic oxide
160,000
Silicon oxide
20,300,000
Ferric oxide
1,980,000
Strontium oxide
79,000
Potassium oxide
75,000
Thorium dioxide
4,400
Lanthanum oxide
48,000
Uranium oxide
2,940,000
Lithium oxide
880,000
Zirconium dioxide
1,630,000
Manganese dioxide
510,000
Other
75,000
Sodium fluoride
280,000
Total
44,000,000
a. Sources: DOE (1997b, page 2-29); Picha
b. To convert kilograms to pounds, multiply
(1998a. Attachment 1).
by 2.2046.
Argonne National Laboratory-West Ceramic and Metal Matrices. Electrometallurgical processing
of DOE spent nuclear fuel containing thermal-bond sodium would result in two high-level radioactive
waste forms for repository disposal, depending on decisions to be based on an going EIS [DOE/EIS-0306
A-41
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-30. Expected glass matrix chemical composition at Idaho Nuclear Technology and Engineering
Center (kilograms)."''
Compound or element
Mass
Compound or element
Mass
Aluminum oxide
Ammoniummolybdophosphate
Boron oxide
Calcium fluoride
Calcium oxide
Ceric oxide
Ferric oxide
Sodium oxide
Phosphorous pentaoxide
130,000
Silicon oxide
1,020,000
26,000
Zirconium dioxide
18,000
200,000
Arsenic
100
140,000
Cadmium
42,000
4,100
Chromium
14,000
300
Mercury*"
200
800
Nickel
1,400
250,000
Lead
1,800
1,000
Total"
1,860,000
a. Sources: Picha (1997, Table ID-3); Heiser (1998, all).
b. Masses are rounded to the nearest 100 kilograms; to convert kilograms to pounds, multiply by 2.2046.
c. Assumes only 0. 1 percent capture of original mercury in the feed materials.
d. Trace amounts of antimony, beryllium, barium, selenium, silver, and thallium were also reported.
(Notice of Intent, 64 FR 8553, February 22, 1999)]. The first form would be a glass-bonded ceramic
composite.
It would stabilize the alkali, alkaline earth, lanthanide, halide, and transuranic materials in processed spent
nuclear fuel. These elements would be present as halides after fuel treatment. For disposal, these
compounds would be stabilized in a zeolite-based material (Goff 1998a, all).
The chemical formula for zeolite-4A, the typical starting material, is Nai2[(A102)i2(Si02)i2]- In the waste
form, zeolite would contain approximately 10 to 12 percent of the halide compounds by weight. The
zeolite mixture typically would be combined with 25-percent glass frit by weight, placed in a
stainless-steel container, and processed into a solid monolith using a hot isostatic press. The zeolite
would convert to the mineral sodalite in the process (Goff 1998a, all). Table A-31 lists the composition
of the waste form.
Table A-31. Expected ceramic waste matrix chemical composition at
Argonne National Laboratory-West (kilograms).
a,b
Component
Mass
Component
Mass
Zeolite-4A
92,000
Potassium iodide
10
Silicon oxide
24,000
Cesium chloride
160
Boron oxide
6,800
Barium chloride
70
Aluminum oxide
2,500
Lanthium chloride
90
Sodium oxide
2,700
Ceric chloride
140
Potassium oxide
140
Praseodymium chloride
70
Lithium-potassium chloride
13,000
Neodymium chloride
240
Sodium chloride
980
Samarium chloride
40
Rubidium chloride
20
Yttrium chloride
J
Strontium chloride
70
Total'
14. JO
a. Source: Goff (1998a, all).
b. To convert kilograms to pounds, multiply by 2.2046.
c. Includes trace amounts of potassium bromide and europium chloride.
The halide composition would depend on the fuel processed. The final bulk composition of the ceramic
waste form by weight percentages would be 25 percent glass, 63 to 65 percent zeolite-4A, and 10 to 12
percent halide salts.
A-42
In ventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-32 lists tne estimated composition of the second high-level radioactive waste form, which is a
metal matrix waste form. The table combines percentage weight distribution with the total expected mass
of the metal waste form to achieve a distributed mass by element (Goff 1998a, all).
Savannah River Site. Fowler et al. (1995, page 4) describes the chemical composition of the Defense
Waste Processing Facility glass in detail. Table A-33 lists the distributed mass of the chemical
constituents that comprise the current design-basis glass for the Savannah River Site. These values are
based on a total mass of the glass of 1 1,600 metric tons (12,800 tons) (Picha 1997, Attachment 1).
West Valley Demonstration Project. The West Valley Demonstration Project will produce a single
type of vitrified high-level radioactive waste. WVNS (1996, WQR-1.1, page 7) provides a target
composition for ail chemical constituents in the high-level radioactive waste. Table A-34 lists the
expected chemical composition based on this target composition and the upper range of the projected total
glass mass, 630 metric tons (694 tons).
Table A-32. Expected metal waste matrix
chemical composition at Argonne National
Laboratory-West (kilograms).^
Component Mass
Iron
4,200
Chromium
1,500
Nickel
1,100
Manganese
180
Molybdenum
220
Silicon
90
Zirconium
1,400
NMFPs"
360
Others'
20
Total
9,000
Source: Goff (1998a, all); to convert
kilograms to pounds, multiply by 2.2046.
NMFPs = Noble metal fission products;
includes silver, niobium, palladium, rhodiiun,
ruthenium, antimony, tin, tantalum,
technetium, and cobalt in small amounts.
Others include trace amounts of carbon,
phosphoms, and sulftir.
A.2.3.5.4 Thermal Output
Hanford Site. The estimated total thermal power from radioactive decay in the 14,500 reference
canisters would be 1,190 kilowatts (as of January 1, 1994). This total heat load equates to an average
power of 82 watts per canister. These values represent the hypothetical situation in which washed sludges
from 177 tanks, cesium concentrates from the decontamination of low-level supemates, and strontium and
cesium materials from capsules would be uniformly blended before vitrification. Realistically, uniform
blending would not be likely. Current planning calls for merging all capsule materials with tank wastes
from 2013 through 2016, which would create much hotter canisters during these years. In the extreme,
the nonuniform blending of cesium concentrates and capsule materials into a relatively small volume of
sludge waste could produce a few canisters with specific powers as high as 2,540 watts, which is the limit
for the nominally 4.5-meter (15-foot) Hanford canisters in the Civilian Radioactive Waste Management
System Baseline (Picha 1997, Attachment 1, page 2; Taylor 1997, all).
A-43
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-33. Expected Savannah River Site high-level radioactive waste
chemical composition (kilograms).'*'^
Glass component
Mass
Glass component
Mass
Aluminum oxide
460,000
Sodium chloride
22,000
Barium sulfate
31,000
Neodymium
13,000
Calcium oxide
110,000
Nickel monoxide
100,000
Calcium sulfate
9,300
Neptunium
100
Cadmium
140
Promethium
210
Cerium
6,800
Praseodymium
3,300
Chromic oxide
14,000
Rubidium
120
Cesium oxide
14,000
Selenium
270
Copper oxide
51,000
Silicon oxide
5,800,000
Europium
200
Samarium
2,200
Ferric oxide
1,200,000
Tin
120
Potassium oxide
450,000
Tellurium
2,200
Lanthanum
3,500
Thorium dioxide
22,000
Lithium oxide
510,000
Titanium dioxide
100,000
Magnesium oxide
160,000
Uranium oxide
250,000
Manganese oxide
230,000
Zirconium
13,000
Molybdenum
14,000
Other"^
58,000
Sodium oxide
1,000,000
Sodium sulfate
12,000
Total
11,600,000
a. Sources: Fowler etal.
(1995, page 4); Picha (1997, Attachment 1).
b. To convert kilograms
to pounds, multiply by 2.2046.
c. Includes trace amounts of silver, americium,
, cobalt, and antimony.
Table A-34. Expected West Valley Demonstration Project chemical
composition (kilograms)."''
Compound
Mass
Compound
Mass
Aluminum oxide
38,000
Nickel monoxide
1,600
Boron oxide
82,000
Phosphorous pentaoxide
7,600
Barium oxide
1,000
Rubidium oxide
500
Calcium oxide
3,000
SiHcon oxide
260,000
Ceric oxide
2,000
Strontium oxide
100
Chromic oxide
900
Thorium dioxide
23,000
Ferric oxide
76,000
Titanium dioxide
4,300
Potassium oxide
32,000
Uranium oxide
3,000
Lithium oxide
24,000
Zinc oxide
100
Magnesium oxide
5,600
Zirconium dioxide
7,100
Manganese oxide
5,200
Others
3,900
Sodium oxide
51,000
Neodymium oxide
900
Total
630,000
a. Sources: WVNS (1996, WQR- 1.1, page
7); Picha (1998c, page 3).
b. To convert kilograms to pounds, multiply by 2.2046.
Idaho National Engineering and Environmental Laboratory. The Laboratory has three proposed
high-level radioactive waste streams. Table A-35 lists the thermal output of these waste streams per
waste canister.
Savannah River Site. The radionuclide inventories reported for the Savannah River Site high-level
radioactive waste in Section A.2.3.5.2 were used to calculate projected heat generation rates for single
canisters.
For the design-basis waste form, the heat generation rates 10 and 20 years after production are 465 and
302 watts per canister, respectively (Plodinec, Moore, and Marra 1993, pages 8 and 9).
A-44
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-35. Idaho National Engineering and Environmental Laboratory waste stream thermal output
(watts).""
Output per waste canister INTEC glass matrix ANL-W ceramic matrix ANL-W metal matrix
Average' 7.1 160 170
Worst case ** 180 620 410
a. Source: Picha (1997, Attachment 1, page 2).
b. INTEC = Idaho Nuclear Technology and Engineering Center; ANL-W = Argonne National Laboratory-West.
c. Based on average case; 2035 used as base year for Idaho Nuclear Technology and Engineering Center glass and 2000 for
ANL-W matrices.
d. Based on worst case; 2020 used as base year for Idaho Nuclear Technology and Engineering Center glass and 2000 for
ANL-W matrices.
West Valley Demonstration Project. West Valley has calculated heat generation rates for a nominal
West Valley canister after several different decay times (WVNS 1996, WQR-3.8, page 2). In the nominal
case, the ORIGEN2-computed heat generation rate was 324 watts at the calculational base time in 1988.
The heat generation rate would decrease continuously from 324 watts to about 100 watts after 50 years of
additional decay.
A.2.3.5.5 Quantity of Waste Per Canister
Table A-36 lists the estimated mass of glass per waste canister for each high-level radioactive waste
stream.
Table A-36. Mass of high-level radioactive waste glass per canister
(kilograms).'
Waste stream"
Mass per canister
Source
Hanford
3,040
Picha (1997, Attachment 1,
page 2)
INEEL
I^r^EC
1,560
Picha (1997, Attachment 1,
page 2)
ANL-W ceramic"
960-1,500
Goff( 1998a, all)
ANL-W metal'
1,500-4,850
Goff( 1998a, all)
Savannah River Site
2,000
Pearson (1998, all)
WVDP
2,000
Picha (1997, Attachment 1,
page 2)
a. To convert kilograms to pounds, multiply by 2.2046.
b. INEEL = Idaho National Engineering and Environmental Laboratory; INTEC = Idaho
Nuclear Technology and Engineering Center; ANL-W = Argonne National
Laboratory-West; WVDP = West Valley DemonsUation Project.
c. These values are estimates. ANL-W is evaluating waste package configiu-ations
compatible with existing storage and remote hot cell facilities. The geometries would
be compatible with the Defense Waste Processing Facility high-level radioactive waste
canister.
A.2.3.5.6 High-Level Radioactive Waste Canister Parameters
Hanford Site. Table A-37 lists preliminary physical parameters for a Hanford Tank Waste Remediation
System standard canister (Picha 1997, Table RL-3).
Idaho National Engineering and Environmental Laboratory. The Idaho Nuclear Technology and
Engineering Center would use stainless-steel canisters identical in design to those used at the Savannah
River Site in the Defense Waste Processing Facility. A similar canister would also be used to contain the
ceramic and metal waste matrices resulting from the proposed high-level radioactive waste processing at
Argonne National Laboratory-West (Picha 1997, Table ID-1).
A-45
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-37. Parameters of proposed Tank Waste Remediation System standard canister for Hanford
high-level radioactive waste disposal."
Parameter
Value"
Comments'^
Length
Outer diameter
Material
Wall thickness
Canister weight
Flange opening
Dished bottom
Available volume
Nominal percent fill
Glass volume
4.50 meters
0.61 meter
304 stainless steel
0.95 centimeter
720 kilograms
0.41 meters
Yes
1 .2 cubic meters
90 percent
1 . 1 cubic meters
1.5 meters longer than DWPF and WVDP canisters - nominal
4.5-meter length
Same as DWPF and WVDP canisters
Same as DWPF and WVDP canisters
Same as DWPF
Same as WVDP canister; large opening
Same as DWPF and WVDP
Provides approximately same void volume as WVDP canister
a. Source: Picha (1997, Table RL-3).
b. To convert meters to feet, multiply by 3.2808; to convert centimeters to inches, multiply by 0.3937; to convert kilograms to
tons, multiply by 0.001 1023; to convert cubic meters to cubic feet, multiply by 35.314.
c. DWPF = Etefense Waste Processing Facility; WVDP = West Valley Demonstration Project.
Savannah River Site. The fabrication specifications of the Defense Waste Processing Facility high-
level radioactive waste canisters are described in detail in Marra, Harbour, and Plodinec (1995, all). The
canisters are fabricated from four basic pieces of A240 304L austenitic stainless steel — the main cylinder,
the bottom head, the top head, and a nozzle. The nominal wall thickness of the canister is 0.95 centimeter
(0.37 inch).
West Valley Demonstration Project. The West Valley canister is designed, fabricated, and handled in
accordance with the specifications in the West Valley Demonstration Project Waste Qualification Report
(WVNS 1996, WQR-2.2, all). The West Valley canisters are fabricated from four principal 304L
austenitic stainless-steel components. The nominal wall thickness of the canister is 0.34 centimeter (0.13
inch).
A.2.3.5.7 Nonstandard Packages
Each site that would ship high-level radioactive waste to the repository has provided additional data on an
estimate of nonstandard packages for possible inclusion in the candidate waste material. The mass,
volume, and radioactivity of potential nonstandard packages would be dominated by failed melters from
the vitrification facilities. Final disposition plans for these melters are in development and vary from site
to site. The EIS used the following assumptions to estimate the potential inventory.
Hanford Site. DOE could need to ship such nonstandard high-level radioactive waste packages as failed
melters and failed contaminated high-level radioactive waste processing equipment to the repository. For
this EIS, the estimated volume of nonstandard packages available for shipment to the repository from the
Hanford Site would be equivalent to that described below for the Savannah River Site.
Idaho National Engineering and Environmental Laboratory. DOE proposes to treat and dispose of
nonstandard packages under existing regulations. However, to bound the number of failed melters the
Idaho National Engineering and Environmental Laboratory could ship to the repository, this EIS uses the
same ratio of failed melters to the number of canisters produced as the Savannah River Site (Palmer 1997,
page 2). The Idaho National Engineering and Environmental Laboratory would produce approximately
20 percent of the number of canisters produced at the Savannah River Site, which assumes 10 failed
A-46
Inventory and Characteristics of Spent Nuclear Fuel, High- Level Radioactive Waste, and Other Materials
melters. Therefore, the Idaho National Engineering and Environmental Laboratory assumes two failed
melters. The volumes and other parameters would then be twice the values listed in Table A-38 for an
individual melter.
Table A-38. Parameters of nonstandard packages from Savannah River Site."
Parameter
Value
Volume
Activity
Mass
Chemical composition
Quantity per disposal package
Thermal generation
10 melters based on current planning to 2021
4.5 equivalent DWPF'' canisters for each melter
1,000 metric tons"^ for 10 melters (filled melter: 100 metric tons)
Glass (see Section A.2.3.5.3)
Melter - Refractory brick
Aluminum
Stainless steel
Inconel
1 melter per disposal package
4.5 times the heat generation of a single canister for each melter
a. Source: Pearson (1997, Attachment 1, pages 3 and 4).
b. DWPF = Defense Waste Processing Facility.
c. To convert metric tons to tons, multiply by 1.1023.
Savannah River Site. Table A-38 lists the estimated parameters of nonstandard packages for repository
shipment from the Savannah River Site.
West Valley Demonstration Project. The West Valley Demonstration Project anticipates that it would
send only one melter to the repository at the end of the waste solidification campaign. It would be treated
as a nonstandard waste package. Table A-39 lists the estimated parameters of nonstandard packages from
the West Valley Demonstration Project.
Table A-39. Parameters of nonstandard packages from West Valley Demonstration Project.'
Parameter Value*"
Volume
Activity
Mass
Chemical composition
Quantity per disposal package
Thermal generator
Source: Rowland (1997, all).
1 melter (24 cubic meters)
1.1 equivalent West Valley canisters
52 metric tons
Melter refractories (38 metric tons)
Inconel ( 1 1 metric tons)
Stainless steel ( 1 .6 metric tons)
Glass (see Table A-34)
1 melter per package
1.1 times the heat generation of a single canister (A.2.3.5.4)
a.
b. To convert cubic meters to cubic feet, multiply by 35.314; to convert meuic tons to tons, multiply by 1.1023.
A.2.4 SURPLUS WEAPONS-USABLE PLUTONIUM
A.2.4.1 Background
The President has declared approximately 50 metric tons (55 tons) of weapons-usable plutonium to be
surplus to national security needs (DOE 1998a, page 1-1). This material includes the following:
• Purified plutonium in various forms (metal, oxide, etc.)
• Nuclear weapons components (pits)
A-47
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
• High-purity materials that DOE could process in the future to produce purified plutonium
• Plutonium residues that DOE previously saved for future recovery of purified plutonium
These materials are currently stored at the Pantex Plant, the Rocky Flats Environmental Technology Site,
the Savannah River Site, the Hanford Site, the Idaho National Engineering and Environmental Laboratory
(Argonne National Laboratory-West), and the Oak Ridge, Los Alamos, and Lawrence Livermore
National Laboratories. DOE would draw the specific surplus weapons-usable plutonium it ultimately
disposed of from the larger inventory primarily stored at these sites.
DOE could process the surplus weapons-usable plutonium as two material streams. One stream would
be an immobilized plutonium ceramic form that DOE would dispose of using a can-in-canister technique
with high-level radioactive waste. The second stream would be mixed uranium and plutonium oxide
fuel assemblies that would be used for power production in light-water reactors and disposed of as
commercial spent nuclear fuel. The Surplus Plutonium Disposition Environmental Impact Statement
(DOE 1998a, page 1-1) evaluates the quantity of plutonium processed in each stream. This EIS assumes
that approximately 18 metric tons (20 tons) of surplus weapons-usable plutonium would be immobilized
and approximately 32 metric tons (35 tons) would be made into mixed-oxide commercial nuclear fuel.
The actual split could include the immobilization of between 1 8 and 50 metric tons (55 tons).
A.2.4.2 Sources
DOE would produce the immobilized plutonium and/or mixed-oxide fuel at sites determined in a Record
of Decision for the Surplus Plutonium Disposition Environmental Impact Statement (DOE 1998a, page
1-9). The Department has selected for further environmental review six alternative commercial light-
water reactors in which it proposes to irradiate the mixed-oxide fuel: both units at Catawba in York,
South Carolina; both units at McGuire in Huntersville, North Carolina; and both units at North Anna
Power Station in Mineral Springs, Virginia (DOE 1999, all).
A.2.4.3 Present Storage and Generation Status
DOE would begin production of the immobilized plutonium in 2006 with an estimated completion by
2016. The immobilization of 18 metric tons (20 tons) of plutonium would produce an estimated
77 additional canisters of high-level radioactive waste, which the production location would store until
shipment to the repository. The immobilization of 50 metric tons (55 tons) of plutonium would produce
an estimated 210 additional canisters of high-level radioactive waste. This EIS assumes that the
production location would be the Savannah River Site and, therefore, used the physical dimensions of the
Defense Waste Processing Facility canisters to calculate these values (DOE 1998a, pages 2-26 and 2-27).
Commercial light-water reactors would use mixed-oxide fuel assemblies for power production starting as
early as 2(X)7. This fuel would replace the low-enriched uranium fuel that normally would be in the
reactors. After the fuel assemblies were discharged from the reactors as spent mixed-oxide fuel, the
reactor sites would store them until shipment to the repository. Mixed-oxide fuel use would produce an
insignificant number of additional spent nuclear fuel assemblies (less than 0.1 percent ) (DOE 1998a,
page 4-378).
A.2.4.4 Final Waste Form
The final waste form would be immobilized plutonium or spent mixed-oxide fuel. Section A.2.4.5
discusses the characteristics of these materials. The spent mixed-oxide fuel discussed here has different
characteristics than the mixed-oxide fuel included in the National Spent Fuel Program (LMIT 1997, all)
and described in Section A.2.2.
A-48
Inventory and Characteristics of Spent Nuclear Fuel, Higli-Level Radioactive Waste, and Other Materials
A.2.4.5 Material Characteristics
A.2.4.5.1 Mixed-Oxide Fuei
A.2.4.5.1.1 Mass and Volume. The EIS on surplus weapons-usable plutonium disposition (DOE
1998a, page 1-9) evaluates the disposal of approximately 32 metric tons (35 tons) of plutonium as mixed-
oxide fuel. The amount of plutonium and uranium measured in metric tons of heavy metal going to a
repository would depend on the average percentage of plutonium in the fuel. The percentage of
plutonium would be influenced by the fuel design. DOE has chosen pressurized-water reactors for the
proposed irradiation of these assemblies. For pressurized-water reactors, the expected average plutonium
percentages would be approximately 4.6 percent; however, they could range between 3.5 and 6 percent
(Stevenson 1997, pages 5 and 6). Table A-40 lists estimates and ranges for the total metric tons of heavy
metal (uranium and plutonium) that would result from disposing of 32 metric tons (35 tons) of plutonium
in mixed-oxide fuel. The table also lists a corresponding estimate for the number of assemblies required,
based on using the typical assemblies described in Section A.2.1.4. The ranges of metric tons of heavy
metal account for the proposed range in potential plutonium percentage.
Table A-40. Estimated spent nuclear fuel quantities for disposition of 32 metric tons of plutonium in
mixed-oxide fuel.^''
Plutonium Best estimate Assemblies Range
Reactor and fuel type percentage (MTHM) required (MTHM)
Pressurized-water reactor 4^56 700 1,500 500-900
a. Source: Stevenson (1997, pages 5 and 6).
b. MTHM = metric tons of heavy metal; to convert metric tons to tons, multiply by 1.1023.
DOE assumed that each spent mixed-oxide assembly irradiated and disposed of would replace an energy-
equivalent, low-enriched uranium assembly originally intended for the repository. The mixed-oxide
assemblies would be part of the 63,(XX) metric tons (69,000 tons) that comprise the commercial spent
nuclear fuel disposal amount in the Proposed Action (Person 1998, all). DOE also assumes that the
average bumup levels for the pressurized-water reactor would be the same as that for the energy-
equivalent, low-enriched uranium fuel. Table A-41 lists the assumed bumup levels and the amount of
heavy metal in an assembly.
Table A-41. Assumed design parameters for typical mixed-oxide assembly.'
Parameter Pressurized-water reactor
Mixed-oxide and low-enriched uranium burnup (MWd/MTHM)'' 45,000
Mixed-oxide assembly mass (kilograms'^ of heavy metal) 450
Mixed-oxide assembly percentage of plutonium 4.56
a. Source; Stevenson (1997, page 7).
b. MWd/MTHM = megawatt days per metric ton of heavy metal; to convert metric tons to tons, multiply by 1 .1023.
c. To convert kilograms to pounds, multiply by 2.2046.
The analysis assumed that the mixed-oxide spent nuclear fuel would replace the low-enriched uranium
fuel. Because of the similarities in the two fuel types, impacts to the repository would be small. Nuclear
criticality, radionuclide release rates, and heat generation comparisons are evaluated in Stevenson (1997,
pages 35 to 37).
A.2.4.5.1. 2 Amount and Nature of Radioactivity. Tables A-42 and A-43 list isotopic composition
data for spent mixed-oxide fuel assemblies. The tables reflect SCALE data files from an Oak Ridge
National Laboratory report used with computer simulation to project the characteristics of spent mixed-
oxide fuel in pressurized-water reactors (Ryman, Hermann, and Murphy 1998, Volume 3, Appendix B).
The tables summarize data for two different potential fuel assemblies: a typical pressurized-water reactor.
A-49
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-42. Radionuclide activity for typical pressurized-water reactor spent mixed-oxide assembly-
Isotope
Curies per assembly
Isotope
Curies per assembly
Hydrogen-3
Carbon- 14
Cobalt-60
Nickel-59
Nickel-63
Krypton-85
Strontium-90
Zirconium-93
Niobium-93m
Niobium-94
Technetium-99
Ruthenium- 106
Iodine- 129
Cesium- 134
Cesium- 137
2.0x10^
3.4x10"'
1.7x10^
1.1
1.4x10^
1.9x10'
1.7x10"
6.5x10"^
2.8x10'
6.8x10"'
6.3
1.6x10*
2.1x10"^
1.4x10"
4.7x10"
Samarium- 151
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Plutonium-238
Plutonium-239
Plutonium-240
Plutonium-241
Americium-241
Americium-242/242m
Americium-243
Curium-242
Curium-243
Curium-244
5.3x10'
4.9x10"^
l.OxIO"'
6.4x10"'
1.4x10"'
1.2x10'
6.6x10^
8.6x10^
2.0x10'
2.2x10'
3.4x10'
2.4x10'
6.0x10'
3.2x10'
2.6x10'
a. Source: Ryman, Hermann, and Murphy (1998, Volume 3, Appendix B).
Table A-43. Radionuclide activity for high-bumup pressurized-water reactor spent mixed-oxide
assembly."
Isotope
Curies per assembly
Isotope
Curies per assembly
Hydrogen-3
Carbon- 14
Cobalt-60
Nickel-59
Nickel-63
Krypton-85
Strontium-90
Niobium-93m
Niobium-94
Technetium-99
Ruthenium- 106
Iodine- 129
Cesium- 134
Cesium- 1 37
Samarium- 151
2.9x10'
5.4x10"'
2.4x10'
1.7
2.3x10^
2.6x10'
2.4x10"
3.9x10'
9.8x10"'
9.0
1.8x10"
3.0x10"^
2.5x10"
7.0x10"
5.4x10^
Uranium-234
Uranium-235
Uranium-236
Uranium-238
Plutonium-238
Plutonium-239
Plutonium-240
Plutonium-241
Americium-241
Americium-242/242m
Americium-243
Curium-242
Curium-243
Curium-244
6.8x10"'
6.7x10"
7.7x10''
1.5X10-'
2.7x10'
4.6x10^
8.8x10^
2.2x10'
2.5x10'
4.9x10'
5.6x10'
1.0x10^
8.5x10'
8.9x10'
a. Sources: Ryman, Hermann, and Murphy (1998, Volume 3, Appendix B).
and a high-bumup pressurized-water reactor. A high bumup pressurized-water assembly would be
irradiated for three cycles in comparison to the two cycles for the typical assemblies. For each of these
assemblies, the tables provide radioactivity data for the common set of nuclides used in this EIS for the
assumed 5-year minimum cooling time.
A.2.4.5.1 .3 Chemical Composition. Tables A-44 and A-45 list the elemental distributions for the
typical and high-bumup pressurized-water reactor spent mixed-oxide fuel assemblies.
A.2.4.5.1. 4 Thermal Output. Table A-46 lists the decay heat from the representative mixed-oxide
spent fuel assemblies at a range of times after discharge.
A.2.4.5.1 .5 Physical Parameters. Because the mixed-oxide fuel would replace low-enriched
uranium fuel in existing reactors, Section A.2. 1.5.5 describes the physical parameters, with the exception
of uranium and plutonium content, which are listed in Table A-41.
A-50
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-44. Elemental distribution of typical bum-up pressurized-water reactor spent mixed-oxide
assembly.^
Element
Grams per
assembly''
Percent'^
Element
Grams per
assembly
Percent
Americium
770
0.12
Palladium
1,200
0.19
Barium
750
0.12
Phosphorus
140
0.02
Carbon
67
0.01
Plutonium
17,000
2.59
Cerium
1,100
0.16
Praseodymium
500
0.08
Cesium
1,500
0.23
Rhodium
360
0.05
Chromium
2,300
0.36
Rubidium
91
0.01
Europium
90
0.01
Ruthenium
1,300
0.20
Iodine
150
0.02
Samarium
440
0.07
Iron
4,600
0.71
Silicon
66
0.01
Krypton
100
0.02
Strontium
210
0.03
Lanthanum
540
0.08
Technetium
370
0.06
Manganese
110
0.02
Tellurium
260
0.04
Molybdenum
1,700
0.27
Tin
1900
0.28
Neodymium
1,700
0.26
Uranium
428,000
65.92
Neptunium
72
0.01
Xenon
2500
0.38
Nickel
4,400
0.68
YtU-ium
110
0.02
Niobium
330
0.05
Zirconium
111,000
17.10
Oxygen
62,000
9.56
Totals
648,000
99.73
a. Source: Murphy (1998, all).
b. To convert grams to ounces, multiply by 0.035274.
c. Table includes only elements that constitute at least 0.01 percent of the total; therefore, total is slightly less
than 100 percent.
Table A-45. Elemental distribution of high bum-up pressurized-water reactor spent mixed-oxide
assembly.'^
Grams per
Grams per
Element
assembly*"
Percent*"
Element
assembly
Percent
Americium
1,000
0.16
Palladium
2,000
0.30
Barium
1,200
0.18
Phosphorus
140
0.02
Carbon
70
0.01
Plutonium
14,000
2.22
Cerium
1,600
0.24
Praseodymium
750
0.11
Cesium
2,100
0.33
Rhodium
460
0.07
Chromium
2,300
0.36
Rubidium
140
0.02
Europium
140
0.02
Ruthenium
2,000
0.31
Iodine
220
0.03
Samarium
630
0.10
Iron
4,600
0.71
Silicon
66
0.01
Krypton
150
0.02
Strontium
300
0.05
Lanthanum
810
0.12
Technetium
520
0.08
Manganese
100
0.02
Tellurium
390
0.06
Molybdenum
2,500
0.39
Tin
1,900
0.29
Neodymium
2,500
0.39
Uranium
421,000
64.84
Neptunium
93
0.01
Xenon
3,700
0.57
Nickel
4,400
0.68
YtU-ium
170
0.03
Niobium
330
0.05
Zirconium
111,000
17.10
Oxygen
62,000
9.56
Totals
646,000
99.46
a. Source: Murphy (1998, all).
b. To convert grains to ounces, multiply by 0.035274.
c. Table includes only elements that constitute at least 0.01 percent of the total; therefore, total is slightly less than 100 percent.
A-51
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-46. Mixed-oxide spent nuclear fuel
thermal profile (watts per assembly)."
Years Typical PWR** High-burnup PWR
1
6,100
8,000
5
1,000
1,600
10
670
1,100
15
610
970
30
540
780
100
370
430
300
240
260
1,000
110
110
3,000
42
38
10,000
25
22
30,000
10
7.9
100,000
1.5
1.3
250,000
0.5
0.6
a. Source: Ryman, Hermann, and Murphy (1998,
Volume 3, Appendix B).
b. PWR = pressurized-water reactor.
A.2.4.5.2 Immobilized Plutonium
At present, approximately 50 metric tons (55 tons) of weapons-usable plutonium have been declared
to be surplus to national needs. DOE has not yet determined the total quantity of plutonium for
immobilization. The Department assumes that approximately 32 metric tons (35 tons) is "clean" metal
suitable for use in mixed-oxide fuel, and that it could dispose of this material by buming it in reactors
(DOE 1998a, page 1-1). The remaining surplus plutonium would require considerable additional
chemical processing to make it suitable for reactor use. This EIS evaluates two cases, one in which
DOE immobilizes only the "impure" materials (base case) and a second in which it immobilizes the entire
50-metric-ton surplus inventory. The base case is evaluated for the Proposed Action because it is DOE's
preferred alternative (DOE 1998a, page 1-1). The EIS evaluates the second case for potential cumulative
impacts (Modules 1 and 2) because it would conservatively predict the largest number of required high-
level radioactive waste canisters.
A.2.4.5.2.1 Mass and Volume. In DOE's preferred disposition alternative, immobilized plutonium
would arrive at the repository in canisters of vitrified high-level radioactive waste that would be
externally identical to standard canisters from the Defense Waste Processing Facility at the Savannah
River Site. Smaller cans containing immobilized plutonium in ceramic disks would be embedded in each
canister of high-level radioactive waste glass. This is the can-in-canister concept. Because the design of
the can-in-canister is not final, DOE has not determined final waste loadings per canister, volume
displaced by the cans, or other specifications. The current baseline concept calls for cylindrical cans that
are 53 centimeters (21 inches) high with a 7.6-centimeter (3-inch) diameter. The gross volume of each
can would be 2.4 liters (150 cubic inches). DOE estimates that each canister would contain 28 cans, but
has not yet finalized the actual number. One of the limitations on the number of cans is determined by the
ability to ensure that the high-level radioacfive waste glass would fill completely around the cans;
increasing the volume that the cans would occupy in a canister could increase the difficulty of achieving
this. Final confirmation of the design will be confirmed by actual test pours at scale (Stevenson 1997,
page 41).
Marra, Harbour, and Plodinec (1995, page 2) describes the volume of a high-level radioactive waste
canister. Each canister has a design capacity of 2,000 kilograms (4,4(X) pounds) of high-level radioactive
waste glass. A nominal glass density of 2.7 grams per cubic centimeter (0. 10 pound per cubic inch)
A-52
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
yields a design glass volume of 620 liters (22 cubic feet). The 28 cans containing plutonium would
displace 68 liters (2.4 cubic feet), or about 1 1 percent of the available volume. The rack holding the cans
would displace about an additional 1 percent of the available volume, yielding a total displacement of
about 12 percent.
Each plutonium can would contain 20 cylindrical pellets, 6.7 centimeters (2.6 inches) in diameter and
2.5 centimeters (1 inch) in height. The pellets would have an average density of 5.5 grams per cubic
centimeter (0.20 pound per cubic inch) and would contain 10.5 percent of plutonium by weight. Each
can, therefore, would contain about 1 kilogram (2.2 pounds) of plutonium, yielding a total of about
28 kilograms (62 pounds) per canister (1 kilogram of plutonium per can multiplied by 28 cans per
canister).
Table A-47 lists the number of high-level radioactive waste canisters required to dispose of immobilized
surplus plutonium using the loading and volumetric assumptions given above for both the base and
50-metric-ton (55-ton) cases. It also lists the number of additional canisters DOE would have to produce
(in addition to those the high-level radioactive waste producer would already have produced) due to the
displacement of high-level radioactive waste glass by the plutonium-containing canisters. The total
number of required canisters would be a function of both the number of cans in each canister and the
plutonium loading of the immobilization form. The number of additional canisters would depend only on
the plutonium loading of the immobilization form.
Table A-47. Number of canisters required for immobilized plutonium disposition.^''
Canisters Base case 50-metric-ton case
Containing plutonium 635 1,744
Inexcessof those required for DWPF*( 12% of total canisters) 77 210
Additional'' 13% 3.5%
a. Source: DOE (1998a, pages 2-26 and 2-27).
b. Assumes 28 kilograms (62 pounds) of plutonium per canister and displacement of 12 percent of the high-level radioactive
waste glass by plutonium cans and rack.
c. DWPF = Defense Waste Processing Facility.
d. As percentage of total planned DWPF canisters (about 6,000).
A.2.4.5.2.2 Amount and Nature of Radioactivity. Assuming the current 10.5-percent plutonium
loading in the ceramic (Stevenson 1997, page 49), the expected isotopic composition of the various
materials in the feedstream for ceramic production, and the nominal quantity of ceramic in each canister,
Stevenson (1997, page 49) calculated the activity of the immobilized material in each high-level
radioactive waste canister. The figures do not include the radioactivity of the vitrified high-level
radioactive waste that would surround the cans of immobilized plutonium. Calculation of the total
radioactivity of a canister requires the subtraction of approximately 12 percent from the radioactivity of a
full high-level radioactive waste canister to account for the displacement of the immobilized plutonium
and its rack. Those reduced numbers, added to the appropriate figures in Table A-48, produce the total
activity of a plutonium-containing high-level radioactive waste canister.
Values for the base case and the 50-metric-ton case are different because the plutonium in the base
case contains more transuranic radionuclides, other than plutonium-239, than does the remainder of the
plutonium [32 metric tons (35 tons)]. Thus, the "other" transuranic radionuclides are diluted in the
50-metric-ton case. From a thermal output and radiological impact standpoint, the base case is a more
severe condition and, therefore, DOE has used it for the Proposed Action analysis.
Section A.2.3.5.2 contains information on the radioactivity contained in a standard Defense Waste
Processing Facility high-level radioactive waste canister.
A-53
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-48. Average total radioactivity of immobilized
Plutonium ceramic in a single canister in 2010 (curies).^''
Nuclide
Base case
50-metric-ton case
Plutonium-238
120
60
Plutonium-239
1,600
1,700
Plutonium-240
550
430
Plutonium-241
4,700
2,800
Plutonium-242
0.098
0.046
Americium-241
720
430
Uranium-234
<0.000015'^
< 0.000005
Uranium-235
0.0024
< 0.001 1
Uranium-238
0.019
0.019
Thorium-232
< 0.00003
< 0.00003
Totals
7,700
5,400
a. Source: Stevenson (1997, page 49).
b. Assumes 10.5 percent of plutonium by weight in ceramic form, 1:2
molar ratio of plutonium to uranium, and 28 kilograms (62 pounds)
of plutonium per canister. These values account only for the
radioactivity in the immobilized form; they do not include that in the
surrounding high-level radioactive waste glass.
c. < = less than.
A.2.4.5.2.3 Chemical Composition. The current design for a ceramic immobilization form is a
multiphase titanate ceramic, with a target bulk composition listed in Table A-49. The neutron absorbers,
hafnium and gadolinium, are each present at a 1-to-l atomic ratio to plutonium, and the atomic ratio of
uranium to plutonium is approximately 2-to-l. For the base case, the presence of impurities in some
categories of surplus weapons-usable plutonium would result in the presence of a few weight percent of
other nonradioactive oxides in some of the actual ceramic; Table A-49 does not list these impurities
(Stevenson 1997, page 51).
Table A-49. Chemical composition of baseline ceramic
immobilization form."
Oxide Approximate percent by weight
Titanium oxide 36
Hafnium oxide 10
Calcium oxide 10
Gadolinium oxide 8
Plutonium oxide 12
Uranium oxide 24
a. Source: Stevenson (1997, page 51).
The ceramic phase assemblage is mostly Hf-pyrochlore [(CaGd)(Gd,Pu,U,Hf)Ti207], with subsidiary
Hf-zirconolite [(CaGd)(Gd,Pu,U,Hf)Ti207)], and minor amounts of brannerite [(U,Pu,Gd)Ti206] and
rutile [(Ti,Hf)02]. Pyrochlore and zirconolite differ in their crystalline structures. The presence of silicon
as an impurity in the plutonium could lead to the formation of a minor amount of a silicate glass phase in
the ceramic. This phase could contain a trace amount of the immobilized plutonium. Some residual
plutonium oxide (less than 0.5 percent of the total quantity of plutonium) could also be present. The
residual plutonium oxide contains uranium with smaller amounts of gadolinium and hafnium as a result of
partial reaction with the other constituents of the ceramic (Stevenson 1997, page 51). Section A.2.3.5.3
describes the chemical composition of the high-level radioactive waste glass surrounding the plutonium-
containing cans.
A-54
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
A.2.4.5.2.4 Thermal Output. Stevenson (1997, page 49) has presented the heat generation of the
immobilized ceramic. These figures represent only the heat from the ceramic; they do not account for the
heat from the surrounding high-level radioactive waste glass. The total heat from a Defense Waste
Processing Facility canister containing high-level radioactive waste and immobilized plutonium would be
the value listed in Table A-50 combined with 88 percent of the value listed in Section A.2.3.5.4 for the
heat from a Defense Waste Processing Facility canister.
Table A-50. Thermal generation from immobilized
plutonium ceramic in a single canister in 2010 (watts
per canister)."
Case Thermal production
Base case 8.6
50-metric-ton'' case 7^0
a. Source: Stevenson (1997, page 49).
b. To convert metric tons to tons, multiply by 1.1023.
A.2.4.5.2.5 Quantity of Material Per Canister. As discussed in Section A.2.4.5.2.1, DOE has yet
to determine the actual configuration of the can-in-canister disposal package. Although the final
configuration could use either the Savannah River Site or Hanford canisters, this EIS assumes the use of
the Savannah River Site canister. The current baseline concept (described above) would result in a per-
canister loading of 28 kilograms (62 pounds) of plutonium. Table A-48 lists the radioactivities of these
materials. Section A.2.3.5.5 discusses the quantity of high-level radioactive waste associated with each
Defense Waste Processing Facility canister. The quantity of high-level radioactive waste in each
plutonium-containing canister would be less than the nominal content of a standard Defense Waste
Processing Facility canister because the displacement of the plutonium cans and the support rack would
amount to an estimated 12 percent of the net canister volume.
The canisters would differ internally from normal Defense Waste Processing Facility canisters due to the
presence of the stainless-steel cans of immobilized plutonium and a stainless-steel rack holding the cans
in place during pouring of molten high-level radioactive waste glass into the canister.
A.2.5 COMMERCIAL GREATER-THAN-CLASS-C LOW-LEVEL WASTE
A.2.5.1 Background
Title 10 of the Code of Federal Regulations, Part 61 (10 CFR Part 61), establishes disposal requirements
for three classes of waste — A, B, and C — suitable for near-surface disposal. Class C has the highest level
of radioactivity and therefore the most rigorous disposal specifications. Wastes with concentrations
above Class C limits (listed in 10 CFR 61.55 Tables 1 and 2 for long and short half-life radionuclides,
respectively) are called Greater-Than-Class-C low-level waste, and are not generally suitable for near-
surface disposal (DOE 1994, all).
Commercial nuclear powerplants, research reactors, radioisotope manufacturers, and other manufacturing
and research institutions generate waste that exceeds the Nuclear Regulatory Commission Class C
shallow-land-burial disposal limits. Public Law 99-240 assigns the Federal Government, specifically
DOE, the responsibility for disposing of this Greater-Than-Class-C waste. DOE could use a number of
techniques for the disposal of these wastes, including engineered near-surface disposal, deep borehole
disposal, intermediate-depth burial, and disposal in a deep geologic repository (DOE 1994, all).
The activities of nuclear electric utilities and other radioactive waste generators to date have produced
relatively small quantities of Greater-Than-Class-C waste. As the utilities take their reactors out of
service and decommission them, they could generate more waste of this type (DOE 1994, all).
A-55
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Greater-Than-Class-C waste could include the following materials:
• Nuclear powerplant operating wastes
• Nuclear powerplant decommissioning wastes
• Sealed radioisotope sources that exceed Class C limits for waste classification
• DOE-held Greater-Than-Class-C waste (addressed in Section A.2.6)
• Greater-Than-Class-C waste from other generators
This section describes the quantities and characteristics of these waste types.
A.2.5.2 Sources
Sources or categories of Greater-Than-Class-C waste include:
• DOE facilities (addressed in Section A.2.6)
• Nuclear utilities
• Sealed sources
• Other generators
Nuclear utility waste includes activated metals and process wastes from commercial nuclear powerplants.
Sealed sources are radioactive materials in small metallic capsules used in measurement and calibration
devices. Other generator wastes consist of sludge, activated metals, and other wastes from radionuclide
manufacturers, commercial research, sealed-source manufacturers, and similar operations. The
decommissioning of light-water reactors probably will generate additional Greater-Than-Class-C waste.
Some internal reactor components will exceed Class C disposal limits.
A.2.5.3 Present Status
Nuclear utilities store their Greater-Than-Class-C waste at the generator site, where it will remain until a
disposal option becomes available.
Sealed sources are held by a Nuclear Regulatory Commission or Agreement State licensee. Current DOE
sealed-source management plans call for the licensees to store their sealed-source wastes until a disposal
option becomes available. If storage by a licensee became physically or financially impossible and a
threat to public health and safety, the Nuclear Regulatory Commission would determine if the source was
a candidate for DOE storage. At that time, the Commission could request that DOE accept the source for
storage, reuse, or recycling. The inventory projections do not include such a transfer of material.
In 1993, there were 13 identified "other generators" of Greater-Than-Class-C waste (DOE 1994,
Appendix D), which were categorized into seven business types:
• Carbon- 14 user
• Industrial research and development
• Irradiation laboratory
• Fuel fabricator
• University reactor
• Sealed-source manufacturer
• Nonmedical academic institution
These generators store their wastes at their sites and will continue to do so until a disposal site becomes
operational.
A-56
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
A.2.5.4 Final Waste Form
The final disposition method for Greater-Than-Class-C waste is not known. If DOE was to place such
waste in a repository, it is assumed that it would be placed in a disposal package before shipment. The
EIS assumes the use of a package similar to the naval dual-purpose canister, which is described in Section
A.2.2.5.6, for all shipments by rail and a package similar to the high-level radioactive waste canisters for
all shipments by truck.
A.2.5.5 Waste Characteristics
Table A-5 1 lists existing and projected volumes for the three Greater-Than-Class-C waste generator
sources. DOE conservatively projects the volume of nuclear utility wastes to 2055 because that date
would include the majority of this waste from the decontamination and decommissioning of commercial
nuclear reactors. The projected volumes conservatively reflect the highest potential volume and activity
based on inventories, surveys, and industry production rates. DOE projects the other two generator
sources (sealed sources and other generators) to 2035 (DOE 1994, all).
Table A-51. Greater-Than-Class-C waste volume
a,b
by generator source (cubic meters).
Source
1993
volume
Projected
volume
Nuclear electric utility
Sealed sources
Other generators
Totals
26
39
74
139
1,300
240
470
2,010
a. Source: DOE (1994, all).
b. To convert cubic meters to cubic feet, multiply by 35.314.
The data concerning the volumes and projections are from Greater-Than-Class-C Waste Characterization:
Estimated Volumes, Radionuclide Activities, and Other Characteristics (DOE 1994), Appendix A-1,
which provides detailed radioactivity reports for such waste currently stored at nuclear utilities. Table
A-52 summarizes the radioactivity data for the primary radionuclides in the waste, projected to 2055.
Table A-52. Commercial light-water reactor
Greater-Than-Class-C waste radioactivity (curies) by
nuclide (projected to 2055).'
Nuclide Radioactivity
Carbon- 14 6.8x10"
Cobalt-60 3.3x10''
Iron-55 1.8x10'
Hydrogen-3 1.2x10"
Manganese-54 3.2xlO"
Niobium-94 9.8x10^
Nickel-59 2.5xl0'
Nickel-63 3.7x10^
Transuranics 2.0x1
0^
Total 8.8x10^
a. Source: DOE (1994, Appendix A-1).
A-57
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Appendix B of DOE (1994) provides detailed radioactivity reports for the sealed sources, which could be
candidate wastes for the repository. Table A-53 summarizes the radioactivity data for the radionucHdes in
these sources, projected to 2035.
Table A-53. Sealed-source Greater-Than-Class-C
waste radioactivity (curies) by nuclide (projected to
2035).'
Nuclide
Radioactivity
Americium-241
8.0x10"
Curiuni-244
1.6x10^
Cesium- 137
4.0x10''
Plutonium-238
1.6x10"
Plutonium-239
l.lxlO'
Plutonium- 241
2.8x10'
Technetium-99
5.8x10^
Uranium-238
5.7x10'
Total
4.2x10'
a. Source: DOE (1994, Appendix B).
DOE (1994, Section 5) also identifies the 13 other generators and the current and projected volumes and
total radioactivity of Greater-Than-Class-C waste held by each. It does not provide specific radionuclide
activity by nuclide. DOE used the data to derive a distribution, by user business type, of the specific
nuclides that comprise the total radioactivity. Table A-54 lists this distributed radioactivity for other
generators.
Table A-54. Other generator Greater-Than-Class-C
waste radioactivity (in curies) by nuclide (projected
to 2035)."
Nuclide
Radioactivity
Carbon- 14
7.7x10^
Transuranic
2.2x10'
Cobalt-60
1.5x10'
Nickel-63
1.5x10'
Americium-241
2.4x10'
Cesium- 137
6.6x10'
Technetium-99
5.1x10'
Total"
1.3x10*
a. Source: Derived from DOE (1994, Appendix D).
b. Total differs from sum of values due to rounding.
A detailed chemical composition by weight percentage for current Greater-Than-Class-C waste is not
available. However, Table A-55 lists the typical composition of such wastes by generator.
Table A-55. Typical chemical composition of Greater-Than-
Class-C wastes."
Source Typical composition
Nuclear electric utility Stainless steel-304, and zirconium
alloys
Sealed sources Stainless steel-304 (source material
has very small mass contribution)
Other generators Various materials
a. Source: DOE (1994, all).
A-58
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
The heat generation rates or thermal profiles for this waste type are not included in the source
documentation. However, the contribution to the total thermal load at the repository from the
Greater-Than-Class-C radioactive waste would be very small in comparison to commercial spent nuclear
fuel or high-level radioactive waste.
A.2.6 SPECIAL-PERFORMANCE-ASSESSMENT-REQUIRED LOW-LEVEL WASTE
A.2.6.1 Background
DOE production reactors, research reactors, reprocessing facilities, and research and development
activities generate wastes that exceed the Nuclear Regulatory Commission Class C shallow-land-burial
disposal limits. The Department is responsible for the safe disposal of such waste, and could use a
number of techniques such as engineered near-surfac*-, disposal, deep borehole disposal, intermediate-
depth burial, or disposal in a deep geologic repository. These wastes have been designated as Special-
Performance-Assessment Required wastes.
DOE Special-Performance-Assessment-Required waste could include the following materials:
• Production reactor operating wastes
• Production and research reactor decommissioning wastes
• Non-fuel-bearing components of naval reactors
• Sealed radioisotope sources that exceed Class C limits for waste classification
• DOE isotope production-related wastes
• Research reactor fuel assembly hardware
A.2.6.2 Sources
DOE has identified Special-Performance-Assessment-Required waste inventories at several locations.
Table A-56 lists the generators and amounts of these wastes. These amounts include current and
projected inventory. The Department will generate additional waste as it decommissions its nuclear
facilities.
Table A-56. Estimated Special-Performance-Assessment-Required low-level
waste volume and mass by generator source.^
Source'' Volume (cubic meters)'^ Mass (kilograms)''
Hanford
INEEL'
ORNL
WVD?
ANL-E
Naval Reactors Facility
Totals
a. Source: Picha (1998b, all).
b. INEEL = Idaho National Engineering and Environmental Laboratory (including Argonne
National Laboratory-West); ORNL = Oak Ridge National Laboratory; WVDP = West Valley
Demonstration Project; ANL-E = Argonne National Laboratory-East.
c. To convert cubic meters to cubic yards, multiply by 1.3079.
d. To convert kilograms to pounds, multiply by 2.2046.
e. Includes Argonne National Laboratory- West.
20
360,000
20
280,000
2,900
4,700,000
550
5,200,000
1
230
500
2,500,000
4,000
13,040,230
A-59
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
A.2.6.3 Present Status
DOE stores its Special-Performance-Assessment-Required waste at the generator sites listed in
Table A-56. Tables A-57 through A-60 list the waste inventories at the individual sites. For
radionuclides, these tables include only the reported isotopes with inventories greater than 1 x 10"' curies.
Table A-6I lists the chemical composition of this material at each site.
Table A-57. Hanford Special-Performance-Assessment-
Required low-level waste radioactivity by nuclide
(curies)."
Nuclide Radioactivity
Cesium- 137 6.0x10'*
Strontium-90 6.0x10"
a. Source: Picha (1998b, all).
Table A-58. Idaho National Engineering and Environmental
Laboratory (including Argonne National Laboratory-West)
Special-Performance-Assessment-Required low-level waste
radioactivity by nuclide (curies)."
Nuclide Radioactivity
Hydrogen-3 5.9x10*
Carbon- 14 8.3x10^
Cobalt-60 1.1x10*
Nickel-59 9.0x10*
Nickel-63 1.3x10"
Strontium-90 7.4x10^
Niobium-94 1.4x10^
Technetium-99 3.3
Cesium-137 3.1xlO'
Radium-226 3.0x10'
Plutonium-239 2.0x10*
Americium-241 2.4x10^
a. Source: Picha (1998b, all).
Table A-59. Oak Ridge National Laboratory Special-
Performance-Assessment-Required low-level waste
radioactivity by nuclide (curies)."
Nuclide Radioactivity
Hydrogen-3 1.9x10*
Carbon- 14 l.OxlO'
Cobalt-60 1.9x10*
Nickel-59 7.6x10^
Nickel-63 7.5x10^
Strontium-90 8.3x10''
Niobium-94 l.OxlO"
Technetium-99 8.0x10"'
Iodine-129 7.5x10"'
Cesium-137 1.7x10""
a. Source: Picha (1998b, all).
A-60
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Table A-60. Radioactivity of naval Special -Performance-Assessment-Required waste (curies per
package)/
Isotope
Short canister
Long canister
Isotope
Short canister
Long canister
Americium-241
5.4x10"^
6.0x10"^
Nickel-59
2.2x10^
2.5x10^
Americium-242m
5.8x10-^
6.5x10-"
Nickel-63
2.7x10"
3.0x10"
Americium-243
5.8x10"
6.5x10""
Plutonium-239
2.1x10"^
2.4x10"'
Carbon- 14
3.2
3.6
Plutonium-240
5.4x10"'
6.0x10"'
Chlorine-36
5.3x10"^-
6.0x10"^
Plutonium-241
4.1
4.6
Curium-242
1.4x10'
1.5x10"'
Plutonium-242
4.5x10"'
5.1x10"'
Curium-243
6.6x10"
7.4x10""
Ruthenium- 106
2.1x10"'
2.3x10"'
Curium-244
7.0x10"^
7.9x10"^
Selenium-79
1.2x10"'
1.3x10"'
Curium-245
1.3x10"'
1.5x10"'
Samarium- 151
1.7x10"^
1.9x10"'
Cesium- 134
1.6
1.8
Tin-126
1.2x10"'
1.3x10"'
Cesium- 135
1.1x10-'
1.2x10"'
Strontium-90
4.2x10"'
4.7x10"'
Cesium- 137
1.1
1.3
Technetium-99
5.3x10""
6.0x10"
Hydrogen-3
1.5
1.7
Uranium-232
1.2x10""
1.4x10""
Krypton-85
4.9x10"^
5.6x10"^
Uranium-233
7.8x10"'
8.8x10"'
Niobium-93m
3.6x10'
4.1x10"'
Zirconium-93
3.8x10"'
4.3x10"'
Niobium-94
5.9x10'
6.7x10"'
a. Source: Beckett (1998, Attachment 1).
Table A-61. Typical chemical composition of Special-Performance-Assessment-
Required low-level waste.^
Source
Composition
Hanford
INEEL
ORNL
WVDP
Naval Reactors
Other generators
Vitrified fission products in glass waste form; hot cell waste
Activated metal
Activated metal; isotope production waste; hot cell waste
Activated metal; vitrified transuranic waste
Activated metal (zirconium alloy, Inconel, stainless steel)
Stainless-steel sealed sources
a. Source: Picha (1998b, all).
b. INEEL = Idaho National Engineering and Environmental Laboratory; ORNL = Oak Ridge National
Laboratory; WVDP = West Valley Demonstration Project.
A.2.6.4 Final Waste Form
The final disposal method for DOE Special-Performance-Assessment-Required waste is not known. If
the Department disposed of such waste in a repository, it is assumed that the material would be placed in
a disposable package before shipment to the repository. The EIS assumes the use of a dual-purpose
canister similar to those used for naval fuels for all rail shipments and packages similar to a high-level
radioactive waste canister for all truck shipments.
A.2.6.5 Waste Characteristics
The low-level waste from West Valley consists of material in the Head End Cells (5 cubic meters
[177 cubic feet]) and remote-handled and contact-handled transuranic waste (545 cubic meters [19,(XX)
cubic feet]). The estimated radioactivity of the material in the Head End Cells is 6,750 curies, while the
activity of the remote-handled and contact-handled transuranic waste is not available at present (Picha
1998b, all). The naval Special-Performance-Assessment-Required waste consists primarily of zirconium
alloys, Inconel, and stainless steel (Beckett 1998, all); Table A-60 lists the specific radioactivity of the
projected material 5 years after discharge.
A-61
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
The specific activity associated with the radium sources at Argonne National Laboratory-East has not
been determined. However, in comparison to the other Special-Performance-Assessment-Required waste
included in this section, its impact would be small.
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Cole 1998a
Cole 1998b
Davis and Wells 1997
Dirkmaat 1997a
Dirkmaat 1997b
Dirkmaat 1998a
Dirkmaat 1998b
Beckett, T. H., 1998, "Response to Data Request," interoffice
memorandum to P. J. Dirkmaat (Idaho Operations Office) and K. G.
Picha (Office of Waste Management), Office of Naval Reactors, U.S.
Department of Energy, Washington, D.C. [MOL. 199905 1 1 .0293]
Cole, B., 1998a, "EIS Comments," memorandum to J. Rivers (Jason
Associates, Inc.), U.S. Department of Energy, Washington, D.C.
[MOL. 199905 11.0303]
Cole, B., 1998b, "Stainless Steel Clad SNF," memorandum to J. Rivers
(Jason Associates, Inc.), U.S. Department of Energy, Washington,
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Davis, N. R., and M. N. Wells, 1997, High-Level Waste System Plan
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attachment to K. Skipper (Office of Civilian Radioactive Waste
Management), U.S. Department of Energy, Idaho Operations Office,
Idaho Falls, Idaho. [MOL. 19970725.0067, correspondence;
MOL. 19970725.0068, attachment]
Dirkmaat, P. J., 1997b, "Revision 1 Response to Repository
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attachment]
Dirkmaat, P. J., 1998a, "Response to Repository Environmental Impact
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Coitmients," interoffice memorandum to K. Skipper (Yucca Mountain
Site Characterization Office), Idaho Operations Office, U.S. Department
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Dirkmaat, P. J., 1998b, "Response to U.S. Department of Energy
(DOE) Appendix A MEPAS Input Parameters Review, and
Miscelaneous [sic] Data - Yucca Mountain Repository Environmental
Impact Statement (EIS)," interoffice memorandum to K. Skipper
(Office of Civilian Radioactive Waste Management), OPE-SFP-98-
171, U.S. Department of Energy, Idaho Operations Office, Idaho Falls,
Idaho. [MOL. 199905 11.0295]
A-62
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
DOE 1985
DOE 1992
DOE 1994
DOE 1995a
DOE 1995b
DOE 1995c
DOE 1996
DOE 1997a
DOE (U.S. Department of Energy), 1985, An Evaluation of
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Level Waste, DOE/DP/0020/1, Director of Defense Waste and
Byproducts, Deputy Assistant Secretary for Nuclear Materials,
Assistant Secretary for Defense Programs, Washington, D.C.
[235263]
DOE (U.S. Department of Energy), 1992, Characteristics of Potential
Repository Wastes, DOE/RW-0184-R1, Oak Ridge National
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HQO. 19920827.0004, Volume 4]
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Level Radioactive Waste Characterization: Estimated Volumes,
Radionuclide Activities, and Other Characteristics, DOE/LLW-1 14,
Revision 1, Idaho National Engineering Laboratory, Idaho Falls,
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Civilian Radioactive Waste Management, Washington, D.C.
[MOV. 19960910.0021]
DOE (U.S. Department of Energy), 1995b, Record of Decision -
Department of Energy Programmatic Spent Nuclear Fuel Management
and the Idaho National Engineering Laboratory Environmental
Restoration and Waste Management Programs, Office of
Environmental Management, Idaho Operations Office, Idaho Falls,
Idaho. [243787]
DOE (U.S. Department of Energy), 1995c, Department of Energy
Programmatic Spent Nuclear Fuel Management and Idaho National
Engineering Laboratory Environmental Restoration arui Waste
Management Programs: Final Environmental Impact Statement,
DOE/EIS-0203-F, Office of Environmental Management, Idaho
Operations Office, Idaho Falls, Idaho. [102617]
DOE (U.S. Department of Energy), 1996, "Amendment to the Record
of Decision for the Department of Energy (DOE) Programmatic Spent
Nuclear Fuel Management and Idaho National Engineering Laboratory
(INEL) Environmental Restoration and Waste Management Programs
Environmental Impact Statement (EIS)," Idaho Operations Office,
Idaho Falls, Idaho. [243792]
DOE (U.S. Department of Energy), 1997a, Nuclear Power Generation
and Fuel Cycle Report 1997, DOE/EIA-0436 (97), Office of Coal,
Nuclear, Electric and Alternate Fuels, Energy Information
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A-63
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
DOE 1997b
DOE 1998a
DOE 1998b
DOE 1998c
DOE 1998d
DOE 1999
Dreyfus 1995
Fillmore 1998
Fowler et al. 1995
DOE (U.S. Department of Energy), 1997b, Integrated Data Base for
1996: U.S. Spent Nuclear Fuel and Radioactive Waste Inventories,
Projections, and Characteristics, DOE/RW-0006, Revision 13, Office
of Environmental Management, Oak Ridge National Laboratories, Oak
Ridge, Tennessee. [242471]
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Office of Fissile Materials Disposition, Washington, D.C. [243236]
DOE (U.S. Department of Energy), 1998b, Viability Assessment of a
Repository at Yucca Mountain, DOE/RW-0508, Office of Civilian
Radioactive Waste Management, Washington, D.C. [U.S.
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MOL. 1998 1007.0028, Volume 1; MOL. 1998 1007.0029, Volume 2;
MOL. 1998 1007.0030, Volume 3; MOL.19981007.0031, Volume 4;
MOL. 1998 1007.0032, Volume 5]
DOE (U.S. Department of Energy), 1998c, Preliminary Design
Specification for Department of Energy Standardized Spent Nuclear
Fuel Canisters, Volume 1 -Design Specification, DOE/SNF/REP-011,
Revision 0, Office of Spent Fuel Management and Special Projects,
Assistant Secretary for Environmental Management, Idaho Operations
Office, Idaho Falls, Idaho. [240539]
DOE (U.S. Department of Energy), 1998d, Savannah River Site Spent
Nuclear Fuel Management Draft Environmental Impact Statement,
DOE/EIS-0279-D, Savannah River Operations Office, Aiken, South
Carolina. [243456]
DOE (U.S. Department Of Energy), 1999, Supplement to the Surplus
Plutonium Disposition Draft Environmental Impact Statement,
DOE/EIS-0283-DS, Office of Fissile Materials Disposition,
Washington, D.C. [244066]
Dreyfus, D. A., 1995, "Proposed Mix Of DOE-Owned High Level
Waste And Spent Nuclear Fuel," interoffice memorandum to J. E.
Lytle (Office of Environmental Management), November 9, Office of
Civilian Radioactive Waste Management, U.S. Department of Energy,
Washington, D.C. [MOL. 1 99903 1 9.034 1 ]
Fillmore, D. L., 1998, Parameter Selection For Department of Energy
Spent Nuclear Fuel To Be Used in the Yucca Mountain Viability
Assessment, INEEL/EXT-98-(X)666, Idaho National Engineering and
Environmental Laboratory, Lockheed Martin Idaho Technologies
Corporation, Idaho Falls, Idaho. [MOL. 199905 1 1 .0296]
Fowler, J. R., R. E. Edwards, S. L. Marra, and M. J. Plodinec, 1995,
Chemical Composition Projections for the DWPF Product (U),
WSRC-IM-91-1 16-1, Revision 1, Westinghouse Savannah River
Company, Aiken, South Carolina. [232731]
A-64
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Goff 1998a
Goff 1998b
Heath 1998
Heiser 1998
Knecht et al. 1999
LMIT 1997
LMIT 1998
Lytle 1995
Marra, Harbour, and Plodinec
1995
Murphy 1998
Goff, K. M., 1998a, "Revision to Original INEEL Response to Yucca
Mountain Site Characterization Office Data Call for High-Level
Waste," memorandum to M. B. Heiser (Lockheed Martin Idaho
Technologies Corporation), Argonne National Laboratory-West, Idaho
Falls, Idaho. [MOL. 19990608.0032]
Goff, K. M., 1998b, "ANL-West Comments from Review of Appendix
A - Yucca Mountain Repository Environmental Impact Statement,"
memorandum to M. B. Heiser (Lockheed Martin Idaho Technologies
Corporation), Argonne National Laboratory-West, Idaho Falls, Idaho.
[MOL. 199905 11.0377]
Heath, C. A., 1998, "DE-AC08-91RW00134; OCRWM Fiscal Year
1998 Annual Work Plan . . .," letter to D. Shelor (Office of Civilian
Radioactive Waste Management, U.S. Department of Energy),
September 24, TRW Environmental Safety Systems Inc., Vienna,
Virginia. [MOV.19981005.0009]
Heiser, M. B., 1998, "INEL HLW vit Breakdown," facsimile to J.
Rivers (Jason Associates, Inc.), March 5, Lockheed Martin Idaho
Technologies Corporation, Idaho Falls, Idaho. [MOL. 199905 1 1 .0370]
Knecht, D. A., J. H. Valentine, A. J. Luptak, M.D. Staiger, H. H. Loo,
and T. L. Wichmann, 1999, Options for Determining Equivalent
MTHM for DOE High-Level Waste, INEEL/EXT-99-00317, Revision
1, Lockheed Martin Idaho Technologies Company, Idaho Falls, Idaho.
[244063]
LMIT (Lockheed Martin Idaho Technologies Corporation), 1997,
DOE National Spent Nuclear Fuel Database, Version 3.2, Idaho Falls,
Idaho. [DTN: M09906DOESFVER32.000]
LMIT (Lockheed Martin Idaho Technologies Corporation), 1998,
Accelerating Cleanup: Paths to Closure, Idaho Operations Office,
PNL-177, Idaho National Engineering and Environmental Laboratory,
Idaho Falls, Idaho. [243437]
Lytle, J. E., 1995, "Disposal of DOE-owned High Level Waste and
Spent Nuclear Fuel," interoffice memorandum to D. A. Dreyfus (Office
of Civilian Radioactive Waste Management), Office of Environmental
Management, U.S. Department of Energy, Washington, D.C.
[HQO. 1995 11 16.0015]
Marra, S. L., J. R. Harbour, and M. J. Plodinec, 1995, DWPF Canister
Procurement, Control, Drop-Test, and Closure (U), WSRC-IM-91-
1 16-8, Revision 1, Westinghouse Savannah River Company, Aiken,
South Carolina. [240797]
Murphy, B. D., 1998, "EIS, Jason requests," internal memorandum to
K. A. Williams, June 4, Computational Physics and Engineering
Division, Oak Ridge National Laboratory, Oak Ridge, Tennessee.
[MOL. 199905 11.0288]
A-65
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Palmer 1997
Parsons 1999
Pearson 1997
Pearson 1998
Person 1998
Picha 1997
Picha 1998a
Picha 1998b
Picha 1998c
Palmer, W. B., 1997, "Clarification to Yucca Mountain Site
Characterization Office Data Call for High Level Waste-WBP- 13-97,"
memorandum to T. L. Wichman (Idaho Operations Office, U.S.
Department of Energy), November 13, Lockheed Martin Idaho
Technologies Company, Idaho Falls, Idaho. [MOL. 19990526.0031]
Parsons Infrastructure and Technology Group, Inc., 1999, Multi-
Canister Overpack Fabrication Specification, HNF-S-0453, Revision 3,
Richland, Washington. [243785]
Pearson, W. D., 1997, "Repository Environmental Impact Statement
(EIS) Data Call for High-Level Waste (HLW)," memorandum to K. G.
Picha (Office of Planning and Analysis), October 22, Savannah River
Operations Office, U.S. Department of Energy, Aiken, South Carolina.
[MOL. 19990303.0336]
Pearson, W. D., 1998, "SRS Data Request Followup," electronic
communication to J. Rivers (Jason Associates Corporation), February
18, U.S. Department of Energy, Savannah River Site, Aiken, South
Carolina. [MOL. 199905 11. 0281]
Person, R., 1998, "Status of MOx in RFP," memorandum to J. Rivers
(Jason Associates Corporation), May 4, U.S. Department of Energy,
Office of Fissile Material Disposition, Washington, D.C.
[MOL. 199905 11.0286]
Picha, K. G., Jr., 1997, "Response to Repository Environmental Impact
Statement Data Call for High-Level Waste," interoffice memorandum
to W. Dixon (Yucca Mountain Site Characterization Office),
September 5, Office of Waste Management, U.S. Department of
Energy, Washington, D.C. [MOL. 19970917.0273]
Picha, K. G., Jr., 1998a, "Clarification of High-Level Waste and
Special Performance Assessment Required Data for Repository
Environmental Impact Statement," interoffice memorandum with
attachments to K. Skipper (Yucca Mountain Site Characterization
Office), May 8, Office of Waste Management, U.S. Department of
Energy, Washington, D.C. [MOL. 19990610.0297]
Picha, K. G., Jr., 1998b, "Special Performance Assessment Required
Waste Supplement for the Yucca Mountain Repository Environmental
Impact Statement," interoffice memorandum with attachments to W.
Dixon (Yucca Mountain Site Characterization Office), May 8, Office
of Waste Management, U.S. Department of Energy, Washington, D.C.
[MOL.19990319.0331, correspondence; MOL.19990319.0332,
attachment]
Picha, K. G., Jr., 1998c, "Follow Up Response to Repository EIS Data
Call for High-Level Waste," interoffice memorandum to W. Dixon,
U.S. Department of Energy, Office of Waste Management,
Washington, D.C. [MOL. 19981006.0206]
A-66
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
Plodinec and Marra 1994
Plodinec, Moore, and Marra
1993
Raddatz and Waters 1996
Rowland 1997
Ryman, Hermann, and
Murphy 1998
Stevenson 1997
Taylor 1997
TRW 1997
TRW 1998
USN 1996
Plodinec, M. J., and S. L. Marra, 1994, Projected Radionuclide
Inventories and Radiogenic Properties of the DWPF Product (U),
WSRC-lM-91-1 16-3, Revision 0, Westinghouse Savannah River
Company, Aiken, South Carolina. [242337]
Plodinec, M. J., F. S. Moore, and S. L. Marra, 1993, Reporting Dose
and Heat Generation Rates of the DWPF Product (U), WSRC-IM-91-
116-12, Revision 0, Westinghouse Savannah River Company, Aiken,
South Carolina. [232736]
Raddatz, M. G., and M. D. Waters, 1996, Information Handbook on
Independent Spent Fuel Storage Installations, NUREG-1571, Spent
Fuel Project Office, Office of Nuclear Material Safety and Safeguards,
U.S. Nuclear Regulatory Commission, Washington, D.C. [231666]
Rowland, T. J., 1997, "Repository Environmental Impact Statement
Data Call for High-Level Waste," interoffice memorandum with
Attachment A to K. G. Picha (Office of Waste Management),
November 26, West Valley Demonstration Project, U.S. Department of
Energy, West Valley, New York. [MOL. 19990608.0048]
Ryman, J. C, O. W. Hermann, and B. D. Murphy, 1998,
Characteristics of Spent Fuel from Plutonium Disposition Reactors,
Volumes 2 and 3, ORN17TM-13170A^2 and V3, Computational
Physics and Engineering Division, Oak Ridge National Laboratory,
Oak Ridge, Tennessee. [239236, Volume 2; 237138, Volume 3]
Stevenson, B., 1997, "Delivery of Data Reports," interoffice
memorandum to W. Dixon (Yucca Mountain Site Characterization
Office), U.S. Department of Energy, Office of Fissile Materials
Disposition, Washington, D.C. [MOL.19971119.0155]
Taylor, W. J., 1997, "Response to Clarification Data for the Repository
Environmental Impact Statement (EIS) Data Call Memorandum Dated
October 3, 1997," interoffice memorandum to K. J. Picha (Office of
Waste Management), November 17, U.S. Department of Energy,
Richland Operations Office, Richland, Washington.
[MOL. 199906 10.0295]
TRW (TRW Environmental Safety Systems Inc.), 1997, Waste
Quantity, Mix and Throughput Study Report, BOOOOOOOO-0 17 17-5705-
00059, Revision 01, TRW, Las Vegas, Nevada.
[MOL.19971210.0628]
TRW (TRW Environmental Safety Systems Inc.), 1998, Controlled
Design Assumptions Document, BOOOOOOOO-017 17-4600-00032,
Revision 05, Las Vegas, Nevada. [MOL. 19980804.0481]
USN (U.S. Navy), 1996, Department of the Navy Final Environmental
Impact Statement for a Container System for the Management of Naval
Spent Nuclear Fuel, DOE/EIS-0251, in cooperation with the U.S.
Department of Energy, Naval Nuclear Propulsion Program, U.S.
Department of the Navy, U.S. Department of Defense, Arlington,
Virginia. [227671]
A-67
Inventory and Characteristics of Spent Nuclear Fuel, High-Level Radioactive Waste, and Other Materials
WVNS 1996 WVNS (West Valley Nuclear Services, Inc.), 1996, WVDP Waste
Form Qualification Report, WVDP- 186, Revision 1, West Valley,
New York. [242094]
A-68
Appendix B
Federal Register Notices
40164
Federal Register / Vol. 60. No. 151 / Monday. August 7. 1995 / Notices
DEPARTMEffT OF ENERGY
Preparation of an Environmental
Impact Statement for a Geologic
Repository for the Disposal of Spent
Nuclear Fuel and High-Level
Radioactive Waste at Yucca Mountain,
|l Nye County, Nevada
\ agency: Department of Energy.
I ACTION: Notice of intent.
: summary: The U.S. Department of
j Energy (DOE) announces its intent to
prepare an environmental impact
statement (EIS) for a geologic repository
at Yucca Mountain. Nye County.
Nevada, for the disposal of spent
nuclear fuel and high-level radioactive
waste, in accordance with the Nuclear
1 Waste Policy Act of 1982. as amended
(NWP.A) (42 U.S.C. §10101 etseq.). the
National Environmental Policy Act
(NEPA) of 1969 (42 U.S.C. §4321 er
seq.). the Council on Environmental
Quality regulations that implement the
procedural provisions of NEPA (40 CFR
Parts 1500-1508). and the DOE
procedures for implementing NEPA (10
CFR Pan 1021). DOE invites Federal.
State, and local agencies. Native
American tribal organizations, and other
interested parties to participate in
determining the scope and content of
the EIS.
The NWPA directs DOE to evaluate
the suitability of the Yucca Mountain
site in southern Nevada as a potential
site for a geologic repository for the
disposal of spent nuclear fuel and high-
level radioactive waste. If the Secretary
of Energy determines that the Yucca
Mountain site is suitable, the Secretary
may then recommend that the President
approve the site for development of a
repository. Under the NWPA. any such
recommendation shall be considered a
major Federal action and must be
accompanied by a final environmental
impact statement. Accordingly. DOE is
preparing this EIS in conjunction with
any potential DOE recommendation
regarding the development of a
repository at Yucca Mountain.
The NWPA provides that the
environmental impact statement need
not consider the need for a repository,
the alternatives to geologic disposal, or
alternative sites to the Yucca Mountain
site. Therefore, this environmental
impact statement will evaluate a
proposal to construct, operate, and
eventually close a repository at Yucca
Mountain. The EIS will evaluate
reasonable alternatives for
implementing such a proposal in
accordance with the NWPA.
The NWPA also provides that the
Nuclear Regulatory Commission shall,
to the extent practicable, adopt DOE's
EIS in connection with any subsequent
construction authorization and license
that the Commission issues to DOE for
a repository. The EIS process is
scheduled to be completed in
September 2000 and is separate from the
licensing process that would be initiated
by any submission of a license
application by DOE to the Commission
in June 200 1 .
The EIS will be prepared over a five-
year period in conjunction with DOE's
separate but parallel site suitability
evaluation and potential license
application. DOE is beginning the EIS
process early to ensure that the
appropriate data gathering and tests are
performed to adequately assess potential
environmental impacts, and to allow cha
public sufficient time to consider this
complex program and to provide input.
B-1
DATES: DOE invites and encourages
comments and suggestions on the scope
of the EIS to ensure that all relevant
environmental issues and reasonable
alternatives are addressed. Public
scoping meetings are discussed below in
the SUPPLEMENTARY INFORMATION section.
DOE will carefully consider all
comments and suggestions received
during the 120-day public scoping
period that ends on December 5, 1995.
Comments and suggestions received
after the close of the public scoping
period will be considered to the e.xtent
practicable.
ADDRESSES: Written comments on the
scope of this EIS. requests to pre-regisier
to speak at any of the public scoping
meetings, questions concerning the
proposed action and EIS, or requests for
additional information on the EIS.
should be directed to: Wendy R. Dixon.
EIS Project Manager. Yucca .Vlountain
Site Characterization Office. Office of
Civilian Radioactive Waste
Management. U.S. Department of
Energy. 101 Convention Center Drive
Suite P-1 10. MS 010. Las Vegas. NV
89109. Telephone: 1-800-967-3477.
Facsimile: 1-800-967-0739.
FOR FURTHER INFORMATION CONTACT: For
more information about this EIS. please
contact Wendy R. Dixon at the address,
above. For information on DOE's NEP.\
process, please contact: Carol M.
Borgstrom. Director. Office of NEP.A.
Policy and .Assistance (EH-42). U.S.
Department of Ene!-g\'. 1000
Independence Avenue. S.W..
Washington. D.C. 20585. Telephone:
1-202-586-4600 or leave a message at
1-800-472-2756.
SUPPLEMENTARY INFORMATION:
Public Participation
All interested persons, including
Federal agencies. Native .American tribal
organizations. State and local
government agencies, public interest
groups, transportation interests,
industry and utility organizations,
regulators, and the general public are
encouraged to take part in the EIS
scoping process. Because of the
anticipated public interest and national
scope of the program. DOE will provide
several methods for people to express
their views and provide con-. rents,
request additional information and
copies of the EIS. or pre-register to
speak at :he scoping meetings.
Comments submitted by any of these
means will oecome par: of the official
record for scoping.
Federal Register / Vol. 60, No. 151 / Monday, August 7. 1995 / Notices
40163
Written Comments and Toll-Free
Facsimile Number
Written comments and requests may
be mailed or sent by facsimile to Wendy
R. Dixon at the address or toll-free
facsimile number listed above
Toil-Free Telephone Line
All interested parties are invited to
record their comments or request
information on the scope of the EIS by
calling a toll-free telephone number, 1-
800-967-3477. Throughout die public
scoping period, this number will be
staffed between the hours of 9 a.m. to
9 p.m. Eastern Standard Time. Monday
through Friday. During other hours,
calls will be forwarded to an answering
machine.
Electronic Mail
Comments and information requests
may be submitted by electronic mail to
the following Internet electronic mail
address: ymp — eisr@notes.ymp.gov.
Internet
The public may access the Notice of
Intent, request iriformation. and provide
comments via the World Wide Web at
the following Uniform Resource Locator
address: http://www.ymp.gov, under
the listing Environmental Impact
Statement (EIS) on the Yucca Mountain
Project Home Page. When available, the
EIS and other selected technical
documents may also be accessed at this
Uniform Resource Locator address.
Scoping Meetings
DOE will hold 15 public scoping
meetings in cities throughout the United
States to provide and discuss
information and to receive comments on
the scope of this EIS. Table 1 at the end
of this Notice lists the specific locations,
dates, and times for each scoping
meeting. Persons wishing to speak at
any of these meetings can pre-register
up to two days before the meeting by:
(1) Calling the toll-free telephone
number 1-800-967-3477, (2) writing to
Wendy R. Dixon at the address listed
above, or (3) sending their request to
pre-register by facsimile or electronic
mail, as identified above.
Persons wishing to speak who have
not registered in advance can register at
each meeting. These "walk-in
registrants" will be accommodated to
the extent practicable, following those
persons who have pre-registered. Only
one spokesperson per organization,
group, or agency may present comments
on its behalf. Oral statements will be
limited to ten minutes: however, written
comments can be of any length and
submitted any time during the scoping
period.
Each of the 1 5 public scoping
meetings will have either a morning or
afternoon session, and an evening
session. Morning sessions will begin at
8:30 a.m. and end at 12:30 p.m., and
afternoon sessions will begin at 12:00
p.m. and end at 4:00 p.m. Evening
sessions will begin at 6:00 p.m. and end
about 10:00 p.m. If additional time is
required in order to accommodate all
speakers wishing to present oral
comments, the meeting facilitator will
consult with the audience and DOE staff
and determine whether to continue the
meeting past the scheduled ending time.
A court reporter will record all portions
of the scoping meetings, and transcripts
will be prepared and made a part of the
official record of the scoping process.
Each session will have an
introductory presentation, a question
and answer period, and a public
comment segment. A facilitator will
begin the introductory presentation of
each session by explaining the scoping
meeting format. DOE staff will provide
a brief description (lasting
approximately 30-45 minutes) of the
repository program, the EIS. and the
scoping process. The question and
answer period (lasting approximately 45
minutes) will provide members of the
public an opportunity to ask questions
and discuss various aspects of the
repository and to obtain additional
information that may be useful in
formulating opinions and comments.
Each member of the public will be
allowed five minutes to ask questions.
The meeting facilitator may allow extra
time for additional questions depending
on the number of people present who
have indicated their desire to participate
during the question and answer period.
The meeting facilitator will begin the
public comment portion of the scoping
meeting after the question and answer
period. At this time, members of the
public will provide their comments on
the scope of the EIS.
Each public scoping meeting also will
have a separate information room
containing exhibits and informational
handouts about the repository program
and the EIS. DOE and contractor staff
will be available throughout the day to
answer questions in an informal setting.
A table with blank comment cards will
also be available for people to privately
prepare and submit written comments
on the scope of the EIS. These comment
cards will be included in the formal
record of each scoping meeting.
Subsequent Document Preparation
Results of scoping, including the
transcripts from the question and
answer periods and public comment
segments, and all other oral and written
1
comments received by DOE. will be
summarized in the EIS Implementation
Plan. This Plan will guide the
preparation of the EIS, and will describl
the planned scope and content of the J
EIS, record the results of the scoping i
process, and contain EIS activity
schedules. As a "living document." the
Implementation Plan may be amended
as needed to incorporate changes in
schedules, alternatives, or EIS content.
The Implementation Plan will be
available to the public for information
purposes as soon as possible after the
close of the public scoping process, and
before issuing the Draft EIS. The
Implementation Plan and the transcripts
from the public scoping meetings will
be available for inspection at major DOE
facilities and public reading rooms in
Nevada and across the country, as
identified at the end of this Notice.
Copies of the Implementation Plan, as
well as the Draft and Final EIS and
related comments, will be provided to
anyone requesting copies of these
documents.
Availability of the Draft EIS for public
review, and the locations and times of
public hearings on the Draft EIS, will be
announced in the Federal Register and
through local media (approximately in
the Fall of 1998). After considering all
public comments received on the Draft
EIS. DOE will prepare and issue a Final
EIS. followed thereafter by a Record of
Decision (approximately in the Fall of
2000).
Background
Spent nuclear fuel ' has been and is
being generated and stored in the
United States as part of commercial
power generation. The accumulation of
spent nuclear fuel from commercial
power reactor operations in the United
States probably will continue for several
decades. There are 109 operating
commercial facilities at 75 sites in 34
States where spent nuclear fuel is
stored. By the year 2035. total spent
nuclear fuel from power reactors will
amount to about 85.000 metric tons of
heavy metal (i.e.. metric tons of heavy
metal, typically uranium, without
materials such as cladding, alloy and
structural materials) (MTHM).
Spent nuclear fuel and high-level
radioactive waste -. generated from
' Spent nuclear fuel is fuel that has been
withdrawn from a nuclear reactor following
irradiation, the constituent elements of which have
not been separated by reprocessing.
= High-level radioactive waste is the highly
radioactive material resulting from reprocessing of
spent nuclear fuel. It includes liquid waste
produced directly in reprocessing and any solid
material derived from such liquid waste that
contains fission products in sufficient
ContinLed
40166
Federal Register / Vol. 60, No. 151 / Monday. August 7, 1995 / Notices
DOE's national atomic energy defense
and research activities, are primarily
located at DOE's Hanford Reservation,
the Savannah River Site, and the Idaho
National Engineering Laboratory. Other
spent nuclear fuel, either currendy in
DOE possession or which may come
under DOE possession, includes
material from foreign research reactors,
approximately 29 domestic university
reactors. 5 non-DOE research reactors,
and 4 "special case" reactors at non-
DOE locations.
In 1982, in response to the continued
accumulation of spent nuclear fuel and
high-level radioactive waste. Congress
passed the NWPA. The purpose of the
NWPA was to establish geologic
repositories that would provide
reasonable assurance that the public and
the environment would be adequately
protected from the hazards posed by
these materials. In 1987, Congress
amended the NWPA and directed DOE
to evaluate the suitability of only the
Yucca Mountain site in southern
Nevada as a potential site for the first
repository. If, based on this evaluation,
the Secretary of Energy determines that
the Yucca Mountain site is suitable, the
Secretary may then recommend that the
President approve the site for
development of a repository.
Under the NWPA. DOE is prohibited
from emplacing more than 70.000
MTHM of spent nuclear fuel and high-
level radioactive waste in the first
repository until such time as a second
repository is in operation. The current
planning basis calls for 63.000 MTHM
of commercial spent nuclear fuel to be
disposed of in the first repository,
proposed to be located at the Yucca
Mountain site. The planning basis also
calls for the disposal of 7.000 MTHM
equivalent of DOE-owned spent nuclear
fuel and high-level radioactive waste in
this first repository.
Proposed Action
If the site were found to be suitable,
the proposed action would be to
construct, operate, and eventually close
a repository at Yucca Mountain for the
geologic disposal of up to 70,000 MTHM
of commercial and DOE-owned spent
nuclear fuel and high-level radioactive
waste. Spent nuclear fuel and high-level
radioactive waste would be disposed of
in the repository in a subsurface
configuration that would ensure its
long-term isolation from the human
environment. Repository construction,
operation, and closure would be
governed by the Nuclear Regulatory
Commission's licensing process.
Construction would begin if the
Nuclear Regulatory Commission
authorizes construction of the
repository. Surface facilities would be
designed and constructed to receive,
and prepare for disposal, spent nuclear
fuel and high-level radioactive waste
that would arrive in transportation casks
by highway and by rail. Capability to
treat or package the secondary wastes
generated during disposal operations
would also be provided. Subsurface
facilities would be designed and
constructed for emplacement of spent
nuclear fuel and high-level radioactive
waste in disposal drifts. Subsurface
facilities would primarily include
access ramps, ventilationsystems.
disposal drifts, and equipment alcoves.
Disposal operations would begin once
the Nuclear Regulatory Commission
issues a license allowing receipt of
spent nuclear fuel and high-level
radioactive waste. Disposal operations
would be expected to last up to 40
years, depending on shipment
schedules. Disposal drifts would
continue to be constructed during this
time period as necessary. Spent nuclear
fuel assemblies,^ and canisters
containing assemblies •* or vitrified (i.e.,
solidified) high-level radioactive waste ^
would be shipped to the repository in
transportation casks that meet the
Nuclear Regulatory Commission and
U.S. Department of Transportation
requirements for shipping by truck or
rail *. The assemblies would be removed
from the transportation casks, which
would be placed back into service after
decontamination and maintenance or
after necessary repairs were completed.
Canisters and assemblies would be
transferred to a "hot" cell — a room
where remotely-controlled equipment
would be used to place the material in
disposal containers. These "waste
packages" (i.e.. assemblies and canisters
concentrations and other highly radioactive
material that the .Nuclear Regulatory Commission,
consistent with e.xistinglaw. determines by rule
requires permanent isolation.
'A fuel assembly is made up of fuel elements
held together by plates and separated by spacers
attached to the fuel cladding.
* Under one scenario, spent nuclear fuel
assemblies would be sealed in a multi-purpose
canister that would then be inserted into separate
caslcs/containers for storage, transportation, and
disposal. Other canisters are available and include
single-purpose systems, which require transferring
of individual assemblies from one cask/container to
another for storage, transport, and disposal. Another
alternative would be dual-purpose systems which
require storing and transporting individual
assemblies in one casl< and disposing of them in
another container.
'Vitrified high-level radioactive waste would be
sealed in canisters suitable for transport in a truck
or train cask.
* Barges may also be used for intermodal
shipments of spent nuclear fuel and high-level
radioactive waste from generator sites to nearby
locations for transfer to truck and rail.
B-3
in disposal containers) would be
transported underground in a
transportation vehicle having radiation
shielding for worker protection.
Monitoring equipment, which would
either be placed in selected drifts or
would be mobile remote-sensing
devices, would monitor performance of
waste packages and aspects of the local
repository geology.
The closure/post-closure period
would begin after the Nuclear
Regulatory Commission amends the
licen.-e to authorize permanent closure.
Underground equipment would be
removed, repository openings would be
backfilled and sealed, and the surface
facilities would be decontaminated,
decommissioned, and dismanded or
converted to other uses. Institutional
controls, such as permanent markers
and monuments, would be designed and
constructed to last thousands of years
and discourage human activities that
could compromise the waste isolation
capabilities of the repository.
The disposal and closure/post-closure
activities would be designed and
implemented so that the combination of
engineered (i.e.. waste package and any
backfill) and natural (geologic system)
barriers would isolate the spent nuclear
fuel and high-level radioactive waste.
The combination of barriers would meet
a standard to be specified by the
Environmental Protection Agency,
which has been entrusted to develop a
radiation release standard pursuant to
Section 801 of the Energy Policy Act of
1992 (42 U.S.C. §10141 note):
individual barriers would perform
according to Nuclear Regulatory
Commission requirements, including its
performance objectives at 10 CFR
60. 1 13. The engineered barrier must
provide substantially complete
containment of spent nuclear fuel and
high-level radioactive waste for between
300 and 1.000 years by using corrosion
resistant materials in the waste package.
Beyond 1.000 years, continued
isolation would be assisted by features
that would limit the rate at which
radioactive components of the waste
would be released. The rate of release
would be substantially affected by
natural conditions, the heat generation
rate of spent nuclear fuel and high-level
radioactive waste (i.e.. thermal load),
and its rate of heat dissipation. First,
different thermal loads would affect
directly the internal and external waste
package temperatures, thereby affecting
the corrosion rate and integrity of the
waste package. Second, the heat would
affect the geochemistry, hydrology, and
mechanical stability of the disposal
drifts, which in turn would influence
the flow of groundwater and the
Federal Register / Vol. 60, No. 151 / Monday. August 7, 1995 / Notices
40167
transport of radionuclides from the
engineered and natural barrier systems
to tlie environment. Therefore, the long-
term performance of the repository
would be managed by appropriately
spacing the waste packages within
disposal drifts and the distances
between disposal drifts, and by
selectively placing spent nuclear fuel
and high-level radioactive waste
packages to account for their individual
heat generation rates.
Alternatives
DOE has preliminarily identified for
analysis in the EIS a full range of
reasonable implementation alternatives
for the construction, operation, and
closure/post-closure of a repository at
Yucca Mountain. These implementation
alternatives are based on thermal load
objectives and include High Thermal
Load. Intermediate Thermal Load, and
Low Thermal Load alternatives.
Under each implementation
alternative, DOE will evaluate different
spent nuclear fuel and high-level
radioactive waste packaging and
transportation options. DOE anticipates
that these options would produce the
broadest range of potential
configurations for both surface facilities
and possible operational and disposal
conditions at the repository. Evaluation
of these options will identify the full
range of reasonably foreseeable impacts
to human health and the environment
associated with each implementation
alternative.
High Thermal Load Alternative
Under the High Thermal Load
implementation alternative, spent
nuclear fuel and high-level radioactive
waste would be disposed in an
underground configuration that would
generate the upper range of repository
temperatures while meeting
performance objectives to isolate the
material in compliance with
Environmental Protection Agency
standards and Nuclear Regulatory
Commission requirements. Under this
alternative, the emplacement density
would likely be greater than 80 MTHM
per acre. This alternative would
represent the highest repository thermal
loading based on available information
and expected test results.
Intermediate Thennal Load Alternative
Under the Intermediate Thermal Load
implementation alternative, spent
nuclear fuel and high-level radioactive
waste would be disposed in an
underground configuration that would
generate an intermediate range of
repository temperatures (compared to
the High and Low Thermal Load
alternatives) while meeting performance
objectives to isolate the material in
compliance with Environmental
Protection Agency standards and
Nuclear Regulatory Commission
requirements. Under this alternative, the
disposal density would likely range
between 40 to 80 MTHM per acre.
Low Thermal Load Alternative
Under the Low Thermal Load
implementation alternative, spent
nuclear fuel and high-level radioactive
waste would be disposed in an
underground configuration that would
provide the lowest potential repository
thermal loading (based on available
information and expected test results)
while meeting performance objectives to
isolate the material in compliance with
Environmental Protection Agency
standards and Nuclear Regulatory
Commission requirements. Under this
alternative, the disposal density would
likely be less than 40 MTHM per acre.
Packaging Options
As part of each implementation
alternative, two packaging options
would be evaluated. Under Option 1 ,
spent nuclear fuel assemblies would be
packaged and sealed in multi-purpose
canisters at the generator sites prior to
being transported to the repository in
Nuclear Regulatory Commission-
certified casks. High-level radioactive
waste also would be packaged and
sealed in canisters prior to shipment in
similar casks. Under Option 2, spent
nuclear fuel assemblies (without
canisters) and sealed canisters of high-
level radioactive waste would be
transported to the repository in Nuclear
Regulatory Commission-certified casks.
Under both options, assemblies and
canisters with intact seals would be
removed from the casks and placed in
disposal containers at the repository.
DOE recognizes that it is likely that a
mix of spent nuclear fuel assemblies
and canisters (and canister systems) of
spent nuclear fuel and vitrified high-
level radioactive waste would arrive at
the repository during disposal
operations. However, since the specific
mix is speculative, the above packaging
options were chosen to produce the
broadest range of potential
configurations for both surface facilities
and possible operational and disposal
conditions at the repository. These
options were also selected to reflect the
potential range of exposures to workers
and the public at the generator sites,
along transportation routes, and at the
repository from the packaging,
transport, and disposal of spent nuclear
fuel and high-level radioactive waste.
B-4
Transportation
As part of each implementation
alterrjative. two national transportation
options and three regional (i.e., within
the State of Nevada) transportation
options would be evaluated. These
options would be expected to result in
the broadest range of operating
conditions relevant to potential impacts
to human health and the environment.
In a national context, the first option
would consist of shipping all spent
nuclear fuel and high-level radioactive
waste by truck, from the generator site
to the repository.
The second national option would
consist of shipment by rail, except from
those generator sites (as many as 19)
that may not have existing capabilities
to load and ship rail casks. For such
sites, the spent nuclear fuel would be
transported by truck to the repository, or
to a facility near the nuclear power
plant where it would be transferred to
rail cars for shipment to the repository.
In a regional context, there are three
transportation options: two of these
options apply to shipments that would
arrive in Nevada by rail, and the third
applies to shipments that would arrive
in Nevada by legal weight truck.''
The first regional transportation
option would consist of several rail
corridors to the repository. The rail
corridor option would involve
identifying and applying siting criteria,
based on engineering considerations
(e.g.. topography and soils), potential
land use restrictions (e.g.. wilderness
areas and existing conflicting uses), and
any other factors identified from the
scoping process.
The second regional transportation
option would involve the use of heavy
haul truck » routes to the repository. The
heavy haul option would include the
construction and use of an intermodal
transfer facility to receive shipments
that would arrive in Nevada by rail: the
intermodal transfer facility would be
located at the beginning of the heavy
haul route. The heavy haul option
would include any need to improve the
local transportation infrastructure.
The third regional transportation
option would involve legal weight truck
shipments directly to the repository.
Under this option, a transfer facility
would not be required.
No Action
The No Action alternative would
evaluate termination of site
'A legal weight truck consists of a tractor, semi-
trailer, and loaded cask, with a maximum gross
weight of 80.000 pounds.
» A heavy haul truck consists of a tractor, semi-
trailer, and loaded cask, with a gross sveight in
excess of 129.000 pounds.
40168
Federal Register / Vol. 60. No. 151 / Monday, August 7. 1995 / Notices
characterization activities at Yucca
Mountain and the continued
accumulation of spent nuclear fuel and
high-level radioactive waste at
commercial storage sites and DOE
facilities. Spent nuclear fuel and high-
level radioactive waste would continue
to be managed for the foreseeable future
at existing commercial storage sites and
DOE facilities located in 34 States. The
No Action alternative, although contrary
to the Congressional desire to provide a
permanent solution for isolation of the
Nations spent nuclear fuel and high-
level radioactive waste, provides a
baseline against which the
implementation alternatives can be
compared.
At the Yucca Mountain site, the
surface facilities, excavation equipment,
and other support facilities would be
dismantled and removed for reuse or
recycling, or would be disposed of in
solid waste landfills. Disturbed surface
areas would be reclaimed and excavated
openings to the subsurface would be
sealed and backfilled.
At commercial reactors, spent nuclear
fuel would continue to be generated and
stored in either water pools or in
canisters, until storage space at
individual reactors becomes inadequate,
at which time reactor operations would
cease. DOE-owned spent nuclear fuel
and high-level radioactive waste would
continue to be managed at three primary
sites — the Hanford Reservation.
Savannah River Site, and the Idaho
National Engineering Laboratory.
Environmental Issues To Be Examined
in the EIS
This EIS will examine the site-specific
environmental impacts from
construction, operation, and eventual
closure of a repository for spent nuclear
fuel and high-level radioactive waste
disposal at Yucca Mountain. Nevada.
Transportation-related impacts of the
alternatives will also be analyzed.
Through internal discussion and
outreach programs with the public. DOE
is aware of many environmental issues
related to the construction, operation,
and closure/post-closure phases of such
a repository. The issues identified here
are intended to facilitate public scoping.
The list is not intended to be all-
inclusive or to predetermine the scope
of the EIS. but should be used as a
starting point from which the public can
help DOE define the scope of the EIS.
• Radiological and non-radiological
releases. The potential effects to the
public and on-site workers from
radiological and nonradiological
releases:
• Public and Worker Safety and
Health. Potential health and safety
impacts (e.g.. injuries) to on-site workers
during the unloading, temporary surface
storage, and underground emplacement
of waste packages at Yucca Mountain;
• Transportation. The potential
impacts associated with national and
regional shipments of spent nuclear fuel
and high-level radioactive waste from
reactor sites and DOE facilities to the
Yucca Mountain site will be assessed.
Regional transportation issues include:
(a) technical feasibility, (b)
socioeconomic impacts, (c) land use and
access impacts, and (d) impacts of
constructing and operating a rail spur, a
heavy haul route, and/or a transfer
facility:
• Accidents. The potential impacts
from reasonably foreseeable accidents,
including any accidents with low
probability but high potential
consequences;
• Criticality. The likelihood that a
self-sustaining nuclear chain reaction
could occur and its potential
consequences:
• Waste Isolation. Potential impacts
associated with the long-term
performance of the repository;
• Socioeconomic Conditions.
Potential regional (i.e.. in Nevada)
socioeconomic impacts to the
surrounding communities, including
impacts on employment, tax base, and
public services;
• Environmental Justice. Potential for
disproportionately high and adverse
impacts on minority or low-income
populations;
• Pollution Prevention. Appropriate
and innovative pollution prevention,
waste minimization, and energy and
water use reduction technologies to
eliminate or significantly reduce use of
energy, water, hazardous substances,
and to minimize environmental
impacts:
• Soil. Water, and Air Resources.
Potential impacts to soil, water quality,
and air quality;
• Biological Resources. Potential
impacts to plants, animals, and habitat,
including impacts to wetlands, and
threatened and endangered species:
• Cultural Resources. Potential
impacts to archaeological/historical
sites. Native American resources, and
other cultural resources:
• Cumulative impacts from the
proposed action and implementing
alternatives and other past, present, and
reasonably foreseeable future actions:
• Potential irreversible and
irretrievable commitment of resources.
Under the No Action alternative,
potential environmental effects
associated with the shutdown of site
characterization activities at Yucca
Mountain will be estimated. Potential
B-5
environmental effects from the
continued accumulation of spent
nuclear fuel and high-level radioactive
waste at commercial reactors and DOE
sites will be addressed by summarizing
previous ;-elevani environmental
analyses and by performing new
analyses of representative sites, as
appropriate. At the Yucca Mountain
site, the potential environmental
consequences from the reclamation of
disturbed surface areas, and the sealing
of excavated openings following the
dismandement and removal of facilities
and equipment, will be quantified.
These analyses would be similar in level
of detail to the analyses of the
implementing alternatives. At the
commercial reactor and DOE sites, the
potential environmental consequences
will be addressed in terms of risk to the
environment and the public from long-
term management of spent nuclear fuel
and high-level radioactive waste. In
addition, the loss of storage capacity,
the need for additional capacity, and
their potential consequences to
continued reactor operations, will be
described.
Consultations With Other Agencies
The NWP.^ requires DOE to solicit
comments on the EIS from the
Department of the Interior, the Council
on Environmental Quality, the
Environmental Protection Agency, and
the Nuclear Regulatory Commission (42
U.S.C. § 10134(a)(1)(D)). DOE also
intends to consult with the Departments
of the Navy and Air Force and will
solicit comments from other agencies,
the State of Nevada, affected units of
local government, and Native American
tribal organizations, regarding the
environmental issues to be addressed by
the EIS.
Relationship to Other DOE NEPA
Reviews
DOE is preparing or has completed
other NEPA documents that may be
relevant to the Office of Civilian
Radioactive Waste Management
Program and this EIS. If appropriate.
this EIS will incorporate by reference
and update information taken from
these other NEPA documents. These
documents (described below) are
available for inspection by the public at
the DOE Freedom of Information
Reading Room (IE- 190). Forrestal
Building. 1000 Independence .A.ve.,
S.W.. Washington. DC. and will be
made available in Nevada at locations to
be announced at the public scoping
meetings. These documents include the
following:
• Environmental Assessment. Yucca
Mountain Sice. Nevada Research and
Federal Register / Vol. 60, No. 151 / Monday. August 7, 1995 / Notices
40169
Development Area, Nevada. DOE/RW-
0073. 1986.
• Environmental Assessment for a
Monitored Retrievable Storage Facility.
DOE/RW-0035. 1986.
• Environmental Impact Statement
for a Multi-Purpose Canister System for
the Management of Civilian and Naval
Spent Nuclear Fuel. The Notice of Intent
was published on October 24, 1994 (59
FR 53442). The scoping process for this
EIS has been completed and an
Implementation Plan is being prepared.
The Draft EIS is scheduled to be issued
for public review in late 1995.
• Programmatic Spent Nuclear Fuel
Management and Idaho National
Engineering Laboratory Environmental
Restoration and Waste Management
Programs Environmental Impact
Statement [Final EIS issued April 1995
POE/EIS-0203-F); Record of Decision
(60 FR 28680-96. June 1. 1995)]. This
EIS analyzes the potential
environmental consequences of
managing DOE's inventory of spent
nuclear fuel over the next 40 years. The
Nevada Test Site was considered but
was not selected as a DOE spent nuclear
fuel management site.
• Waste Management Programmatic
Environmental Impact Statement
(formerly Environmental Management
Programmatic EIS). A revised Notice of
Intent was published January 24. 1995
(60 FR 4607). This Programmatic EIS
will address impacts of potential DOE
waste management actions for the
treatment, storage, and disposal of
waste. The Draft EIS is scheduled to be
issued for public review in September
1995.
• Environmental Impact Statement
for a Proposed Nuclear Weapons
Nonproliferation Policy Concerning
Foreign Research Reactor Spent Nuclear
Fuel [Notice of Intent published October
21, 1993 (58 FR 54336)]. The draft EIS
was issued for public review in March
1995 (DOE/EIS-0218D). This EIS
addresses the potential environmental
impacts of the proposed policy's
implementation. Under the proposed
policy, the United States could accept
up to 22.700 foreign research reactor
spent nuclear fuel elements over a 10-
15 year period.
• Environmental Impact Statement
on the Transfer and Disposition of
Surplus Highly Enriched Uranium
(formerly part of the Programmatic
Environmental Impact Statement for
Long-Term Storage and Disposition of
Weapons-Usable Fissile Materials). The
Notice of Intent was issued April 5.
1995 (60 FR 17344). This EIS will
address disposition of DOE's surplus
highly enriched uranium to support the
President's Nonproliferation Policy. The
Draft EIS is scheduled to be issued in
September 1995.
• Programmatic Environmental
Impact Statement for Storage and
Disposition of Weapons-Usable Fissile
Materials [Notice of Intent published
June 21. 1994 (59 FR 31985)]. This
Programmatic EIS will evaluate
alternatives for long-term storage of all
weapons-usable fissile materials
(primarily plutonium and highly
enriched uranium retained for strategic
purposes — not surplus) and disposition
of surplus weapons-usable fissile
materials (excluding highly enriched
uranium), so that risk of proliferation is
minimized. The Nevada Test Site is a
candidate storage site.
• Tritium Supply and Recycling
Programmatic Environmental Impact
Statement. A revised Notice of Intent
was published October 28. 1994 (59 FR
54175), and the Draft Programmatic EIS
was issued in March 1995 (60 FR 14433,
March 17. 1995). Public hearings on the
Draft Programmatic EIS were held in
April 1995. and a Final Programmatic
EIS is scheduled for October 1995. This
EIS addresses how to best assure an
adequate tritium supply and recycling
capability. The Nevada Test Site is an
alternative site for new tritium supply
and recycling facilities.
• Stockpile Stewardship and
Management Programmatic
Environmental Impact Statement. A
Nodce of Intent was published June 14.
1995 (60 FR 31291). A prescoping
workshop was held on May 19. 1995.
and scoping meetings are scheduled to
be held during July and August 1995.
This Programmatic EIS will evaluate
proposed future missions of the
Stockpile Stewardship and Management
Program and potential configuration
(facility locations) of the nuclear
weapons complex to accomplish the
Stockpile Stewardship and Management
Program missions. The Nevada Test Site
is an alternative site for potential
location of new or upgraded Stockpile
Stewardship and Management Program
facilities.
• Site-Wide Environmental Impact
Statement for the Nevada Test Site
[Notice of Intent published August 10.
1994 (59 FR 40897)]. This EIS will
address resource management
alternatives for the Nevada Test Site to
support current and potential future
missions involving defense programs,
research and development, waste
management, environmental restoration,
infrastructure maintenance,
transportation of wastes, and facility
upgrades and alternative uses. The
public scoping process has been
completed, and the Implementation
Plan was issued in July 1995. The Draft
B-6
EIS is scheduled to Lo issued for public
review in September 1995.
• Environmental Impact Statement
for the Continued Operation of the
Pantex Plant and Associated Storage of
Nuclear Weapon Components [Notice of 1
Intent published May 23. 1994 (59 FR '
26635); an amended Notice of Intent
published June 23, 1995 (60 FR 32661)].
This EIS will address the potential
environmental impacts of the continued
operation of the Pantex Plant, which
includes near- to mid-term foreseeable
activities and the nuclear component
storage activities at other DOE sites
associated with nuclear weapon
disassembly operations at the Pantex
Plant. The Nevada Test Site is being
considered as an alternative site for
relocation of interim plutonium pit
storage.
Public Reading Rooms
Copies of the Implementation Plan,
and the Draft and Final EISs, will be
available for inspection during normal
business hours at the following public
reading rooms. DOE may establish
additional information locations and
will provide an updated list at the
public scoping meetings.
Albuquerque Operations Office,
National Atomic Museum. Bldg.
20358. Wyoming Blvd.. S.E.. Kirtland
Air Force Base. Albuquerque. NM
871 17. Attn: Diane Leute (505) 845-
4378
Atlanta Support Office, U.S. Dept. of
Energy, Public Reading Room, 730
Peachtree Street, Suite 876, Atlanta,
GA 30308-1212. Attn: Nancy Mays/
Laura Nicholas (404) 347-2420
Bartlesville Project Office/National
Institute for Petroleum and Energy
Research, Library, U.S. Dept. of
Energy, 220 Virginia Avenue,
Bartlesville. OK 74003. Attn: Josh
Stroman (918) 337-4371
Bonneville Power Administration. U.S.
Dept. of Energy. BPA-C-KPS-1. 905
N.E. 1 1th Street. Portland. OR 97208.
Attn: Sue Ludeman (503) 230-7334
Chicago Operations Office. Document
Dept.. University of Illinois at
Chicago, 801 South Morgan Street,
Chicago. IL 60607. Attn: Seth Nasatir
(312) 996-2738
Dallas Support Office, U.S. Dept. of
Energy. Public Reading Room, 1420
Mockingbird Lane, Suite 400, Dallas,
TX 75247. Attn: Gailene Reinhold
(214) 767-7040
Fernald Area Office, U.S. Dept. of
Energy. Public Information Room.
FERMCO. 7400 Willey Road,
Cincinnati. OH 45239. Attn: Gary
Stegner (513) 648-3153
Headouarters Office, U.S. Dept. of
E.T-rgy. Rol .£-190. Forrestal Bldg.,
40170
Federal Register / Vol. 60. No. 151 / Monday, August 7, 1995 / Notices
1000 Independence Avenue. S.W.,
Washington. D.C. 20585. Attn: Gayla
Sessonis (202) 586-5955
Idaho Operations OfRce, Idaho Public
Reading Room, 1776 Science Center
Dr.. Idaho Falls. ID 83402. Attn: Brent
Jacobson (208) 526-1144
Kansas City Support Office. U.S. Dept.
of Energy, Public Reading Room. 911
Walnut Street, 14th Floor, Kansas
City. MO 64106. Attn: Anne Scheer
(816) 426-4777
Office of Civilian Radioactive Waste
Management National Information
Center. 600 Maryland Avenue. S.W.,
Suite 760, Washington, D.C. 20024.
Attn: Paul D'Anjou (202) 488-6720
Oak Ridge Operations Office, U.S. Dept.
of Energy, Public Reading Room, 55
South Jefferson Circle, Room 112, Oak
Ridge, TN 37831-8510. Attn: Amy
Rothrock (615) 576-1216
Oakland Operations Office, U.S. Dept. of
Energy. Public Reading Room, EIC,
8th Floor, 1301 Clay Street, Room
700N, Oakland, CA 94612-5208. Attn:
Laura Noble (510) 637-1762
Pittsburgh Energy Technology Center,
U.S. Dept. of Energy, Bldg. 922/M210,
Receiving Department, Building 166.
Cochrans Mill Road. Pittsburgh. PA
15236-0940. Attn: Ann C. Dunlap
(412) 892-6167
Richland Operations Office, U.S. Dept.
of Energy, Public Reading Room, 100
Sprout Rd.. Room 130 West, Mailstop
H2-53, Richland. WA 99352. Attn:
Terri Traub (509) 376-8583
Rocky Flats Field OfRce, Front Range
Community College Library, 3645
West 1 12th Avenue, Westminster, CO
80030. Attn: Nancy Ben (303) 469-
4435
Savannah River Operations Office.
Gregg-Graniteville Library. University
of S. Carolina-Aiken. 171 University
Parkway. Aiken. SC 29801. Attn:
James M. Gaver (803) 725-2889
Southeastern Power Administration.
U.S. Dept. of Energy, Legal Library,
Samuel Elbert Bldg., 2 South Public
Square, Elberton, GA 30635-2496.
Table 1 .—Scoping Meetings
Attn: Joel W. Stymour/Carol M.
Franklin (706) 213-3800
Southwestern Power Administration,
U.S. Dept. of Energy, Public Reading
Room, 1 West 3rd. Suite 1600, Tulsa.
OK 74103. Attn: Marti Ayers (918)
581-7426
Strategic Petroleum Reserve Project
Management Office, U.S. Dept. of
Energy. SPRPMO/SEB Reading Room.
900 Commerce Road East. New
Orleans. LA 70123. Attn: Ulysess
Washington (504) 734-4243
Yucca Mountain Science Centers
Yucca Mountain Science Center. U.S.
95— Star Route 374. Beatty, NV
89003. Attn: Marina Anderson (702)
553-2130
Yucca Mountain Science Center,
4101-B Meadows Lane, Las Vegas.
NV 89107. Attn: Melinda D'ouville
(702) 295-1312
Yucca Mountain Science Center, 1141
South Hwy. 160. Pahrump. NV
89041. Attn: Lee Krumm (702) 727-
0896
Location of scoping meeting
Dates/times '
Pahrump Community Center, 400 N. Hwy. 160, Pahrump, NV 89048 ....
Boise Centre on the Grove, 850 W. Front St., Boise, ID 83702
Lawlor Events Center, University of Nevada-Reno Campus, Reno, NV
89667.
University of Chicago, Downtown MBA Center, 450 N. Cityfront Plaza
Drive, Chicago, IL 60611.
Cashman Reld, 850 Las Vegas Blvd. North, Las Vegas, NV 89101
Denver Convention Complex, 700 14th Street, Denver, CO 80202
Sacramento Public Library, 828 I Street, Sacramento, CA 95814
Arlington Community Center, 2800 South Center Street, Dallas, TX
76004.
Caliente Youth Center, Highway 93, Caliente, NV 89008
Hilton Inn, 150 West 500 South, Salt Lake City, UT 84111
Maritime Institute of Technology and Graduate Studies. 5700 Ham-
monds Ferry Rd.. Linthicum (near Baltimore), MD 21090.
Russell Sage Conference Center, 45 Ferry St., Troy (Albany), NY
12180.
Georgia Intemational Convention Center, 1902 Sullivan Road, College
Pari* (Atlanta). GA 30337.
Penn Valley Community College, 3201 S.W. Trafficway, Kansas City,
MO 64111.
Tonopah Convention Center, 301 Brougher, Tonopah, NV 89049
Tuesday, August 29, 1 995, morning/evening sessions.
Wednesday, September 6, 1995, morning/evening sessions.
Friday, September 8, 1995, morning/evening sessions,
Tuesday, September 12, 1995, morning/evening sessions.
Friday, September 15, 1995, moming/evening sessions .
Tuesday, September 19, 1995, afternoon/evening sessions.
Thursday, September 21, 1995, afternoon/evening sessions,
Tuesday, September 26, 1995, aftemoon/evening sessions,
Thursday, September 28, 1995, moming/evening sessions.
Thursday, October 5, 1995, aftemoon/evening sessions.
Wednesday, October 11, 1995, moming/evening sessions.
Friday, October 13, 1995, aftemoon/evening sessions.
Tuesday, October 17, 1995, moming/evening sessions.
Friday, October 20, 1995, aftemoon/evening sessions.
Tuesday, October 24, 1995, moming/evening sessions.
' Session times are as follows: Moming (8:30 a.m.-12:30 p.m.), Afternoon (12:00 a.m.-4:00 p.m.), Evening (6:00 p.m.-10;00 p.m.).
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Federal Register / Vol. 64. No. 112 /Friday, June 11. 1999 /Notices
DEPARTMENT OF ENERGY
Roodplain and Wetlands Involvement;
Geologic Repositot7 for the Disposal
of Spent Nuclear Fuel and High-Level
Radioactive Waste at Yucca Mountain,
Nye County, Nevada
AGENCY: Department of Energy.
ACTION: Notice of floodplain and
wetlands involvement.
summary: The U.S. Department of
Energy (DOE) is proposing to construct,
operate and monitor, and eventually
close a geologic repository for the
disposal of spent nuclear fuel and high-
level radioactive waste at Yucca
Mountain, Nye County, Nevada. As part
of its proposal, DOE is considering
shipping spent nuclear fuel and high-
level radioactive waste in the State of
Nevada over a rail line that would be
constructed or over an existing highway
route that may need upgrading to
accommodate heavy-haul trucks.
Portions of the rail cortidor or highway
route would cross perennial and
ephemeral streams and their associated
floodplains, as well as possible
wetlands. Furthermore, portions of the
transportation system in the immediate
vicinity of the proposed repository
would be located within the 100-year
floodplains of Midway Valley Wash,
Drillhole Wash. Busted Butte Wash and/
or Fortymile Wash. No other aspect of
repository-related operations or nuclear
or nonnuclear repository facilities
would be located within the 500-year or
100-year floodplains of these washes. In
accordance with DOE regulations for
Compliance with Floodplain/Wetlands
Environmental Review Requirements
(10 CFR Part 1022), DOE will prepare a
floodplain and wetlands assessment
commensurate with proposed decisions
and available information. The
assessment will be included in the
Environmental Impact Statement (EIS)
for a Geologic Repository for the
Disposal of Spent Nuclear Fuel and
High-Level Radioactive Waste at Yucca
Mountain. Nye County. Nevada. A draft
of this EIS is scheduled to be published
during the summer of 1999.
DATES: The public is invited to comment
on ihis notice on or before July 1, 1999.
Comments received after this date will
be considered to the extent practicable.
ADDRESSES: Comments on this notice
should be addressed to Ms. Wendy
Di.xon. EIS Project Manager. Yucca
Mountain Site Characterization Office,
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Federal Register / Vol. 64. No. 112/Friday. June 11. 1999/Noiices
31555
U.S. Department of Energy, P.O. Box
30307. M/S 010, Las Vegas. Nevada
89036-0307. Comments also can be
submitted via electronic mail to:
eisr@notes.ymp.gov.
FOR FURTHER INFORMATION CONTACT :
Proposed Action: Ms. Wendy Dixon.
EIS Project Manager, at the above
address, or by calling (800) -88 1-7292.
Floodplain and Wetlands
Environmental Review Requirements:
Ms. Carol Borgstrom. Office of NEPA
Policy and Assistance (EH-42). U.S.
Department of Energy. 1000
Independence Avenue. S.W..
Washington. D.C. 20585. (202)-586-
4600 or leave a message at (800) 472-
2756.
SUPPLEMENTARY INFORMATION : In
accordance with the Nuclear Waste
Policy Act. as amended. DOE is
studying Yucca Mountain in Nye
County. Nevada, to determine its
suitability for the deep geologic disposal
of commercial and DOE spent nuclear
fuel and high-level radioactive waste. In
1989. DOE published a Notice of
Floodplain/Wetlands Involvement (54
FR 6318, February 9. 1989) for site
characterization at Yucca Mountain, and
in 1992 published a Floodplain
Statement of Findings (57 FR 48363.
October 23. 1992).
DOE is now preparing an EIS (DOE-
EIS-0250) to assess the potential
environmental impacts from the
construction, operation and monitoring,
and eventual closure of the proposed
geologic repository. DOE issued a Notice
of Intent to prepare the EIS on August
7. 1995 (60 FR 40164). As part of its
proposal. DOE is considering shipping
spent nuclear fuel and high-level
radioactive waste in the State of Nevada
over a rail line that would be
constructed or over an existing highway
route that may need upgrading to
accommodate heavy-haul trucks. For the
rail mode. DOE is evaluating five
potential corridors (Figure 1). For the
heavy-haul truck mode. DOE is
evaluating three potential locations for
an intermodal transfer station associated
with five potential highway routes
(Figure 2: an intermodal transfer station
is a facility at which shipping casks
containing spent nuclear fuel and high-
level radioactive waste would be
transferred from trains to trucks, and
empty shipping casks would be
transferred from trucks to trains). The
rail corridors would be about 400 meters
(0.25 mile) wide. The Carlin Corridor
would be the longest at 520 kilometers
(323 miles) followed by the Caliente
(513 kilometers. 319 miles). Caliente-
Chalk Mountain (345 kilometers. 214
miles). Jean (181 kilometers. 112 miles).
and Valley Modified (159 kilometers. 98
miles) corridors. The heavy-haul routes
would utilize existing roads and rights-
of-ways which typically would be less
than 400 meters (0.25 miles) in width.
The Caliente Route would be the longest
at 533 kilometers (331 miles) followed
by the Caliente-Las Vegas (377
kilometers. 234 miles), Caliente-Chalk
Mountain (282 kilometers. 175 miles),
Sloan/Jean (190 kilometers. 1 18 miles)
and Apex/Dry Lake (183 kilometers. 114
miles) routes.
Portions of the transportation system
in the immediate vicinity of the
proposed repository are likely to be
located within the 100-year floodplains
of Midway Valley Wash. Drillhole
Wash. Busted Butte Wash and/or
Fortymile Wash (Figure 3). Fortymile
Wash, a major wash that flows to the
Amargosa River, drains the eastern side
of Yucca Mountain. Midway Valley
Wash. Drillhole Wash and Busted Butte
Wash are tributaries to Fortymile Wash.
Although water flow in Fortymile Wash
and its tributaries is rare, the area is
subject to flash flooding from
thunderstorms and occasional sustained
precipitation. There are no naturally
occurring wetlands near the proposed
repository facilities, although there are
two man-made well ponds in Fortymile
Wash that support riparian vegetation.
If the Proposed Action were
implemented. DOE would use an
existing road during construction of the
repository that crosses the 100-year
floodplain of Fortymile Wash (Figure 3).
This road and other features of site
characterization that involve floodplains
have previously been examined by DOE
and a Statement of Findings was issued
in 1992 (57 FR 48363. October 23.
1992). It is uncertain at this time
whether this existing road would
require upgrading to accommodate the
volume and type of construction
vehicles.
In addition, transportation
infrastructure would be constructed
either in Midway Valley Wash, Drillhole
Wash and Busted Butte Wash, or in
Midway Valley Wash. Drillhole Wash
and Fortymile Wash. The decision on
which washes would be involved is
dependent on future decisions regarding
the mode of transport (rail or truck)
which, in turn, would require the
selection of one rail corridor or the
selection of one site for an intermodal
transfer station and its associated heavy-
haul route. Structures that might be
constructed in a floodplain could
include one or more bridges to span the
washes, one or more roads that could
pass through the washes, or a
combination of roads and culverts in the
washes. No other aspect of repository-
related operation of nuclear or
nonnuclear facilities would be located
within 500-year or 100-year floodplains.
Outside of the immeciiate vicinity of
the proposed repository, the five rail
corridors, and the three sites for an
intermodal transfer station and
associated five heavy-haul routes,
would cross perennial and ephemeral
streams, and possibly wetlands. It is
likely that a combination of bridges,
roads and culverts, or other engineered
features, would be needed to span or
otherwise cross the washes and possible
wetlands, although the location of such
structures is uncertain at this time.
DOE will prepare an initial floodplain
and wetlands assessment commensurate
with the proposed decisions and
available information. This assessment
will be included in the Draft EIS that is
scheduled to be issued for public
comment later this summer. If. after a
possible recommendation by the
Secretary of Energy, the President
considers the site qualified for an
application to the U.S. Nuclear
Regulatory Commission for a
construction authorization, the
President will submit a
recommendation of the site to Congress.
If the site designation becomes effective,
the Secretary of Energy will submit to
the Nuclear Regulatory Commission a
License Application for a construction
authorization. DOE would then
probably select a rail corridor or a site
for an intermodal transfer station among
those considered in the EIS. Following
such a decision, additional field
surveys, environmental and engineering
analyses, and National Environmental
Policy Act reviews would likely be
needed regarding a specific rail
alignment for the selected corridor or
the site for the intermodal transfer
station and its associated heavy-haul
truck route. When more specific
information becomes available about
activities proposed to take place within
floodplains and wetlands. DOE will
conduct further environmental review
in accordance with 10 CFR Part 1022.
Information that would be considered in
a subsequent assessment includes, for
e.xample, the identification of 500-year
and 100-year floodplains among feasible
alignments of the selected rail corridor
or the site of the intermodal transfer
station and its associated heavy-haul
route, identification of individual
wetlands, and whether the floodplains
and wetlands could be avoided. If the
Hoodplains and wetlands could not be
avoided, information on specific
engineering designs and associated
construction activities in the floodplains
and wetlands also would be needed to
permit a more detailed assessment and
B-9
31556 Federal Register / Vol. 64. No. 112/Friday. June 11. 1999/No(:v;es
to ensure that DOE minimizes potential Issued in Las Vegas. Nevada, on the 4th
harm to or within any affected '^y of June 1999.
floodplains or wetlands. Wendy Dixon,
EIS Project Manager.
BILUNG CODE C4S0-01-P
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Federal Register /Vol. 64, No. 112 /Friday. June 11. 1999 /Notices
31557
Figure 1. Potential Nevada rail corridors to Yucca Mountatin.
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31558
Federal Register / Vol. 64. No. 112 /Friday, June 11, 1999 /Notices
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Federal Register/Vol. 64. No. 112/Friday. June 11. 1999/Notices
31559
Figure 3. Yucca Mountain site topography, plains, and potential rail corridors.
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Appendix C
Interagency and
Intergovernmental
Interactions
Interagency and Intergovernmental Interactions
TABLE OF CONTENTS
Section Page
C.l Summary of Activity C-1
C.2 Interests of Selected Agencies and Organizations in the Yucca Mountain
Repository Proposal C-1
C.2.1 Bureau of Land Management C-1
C.2.2 U.S. Air Force C-4
C.2.3 Naval Nuclear Propulsion Program C-4
C.2.4 Fish and Wildlife Service C-5
C.2.5 National Marine Fisheries Service C-5
C.2.6 U.S. Department of Transportation , C-6
C.2.7 U.S. Environmental Protection Agency C-6
C.2.8 U.S. Nuclear Regulatory Commission C-6
C.2.9 U.S. Army Corps of Engineers C-7
C.2. 10 U.S. Department of Agriculture C-7
C.2. 11 Native American Tribes C-7
C.2. 12 Affected Units of Local Government C-9
C.2.13 National Park Service C-9
C.2.14 State of Nevada C-9
C.2. 15 Advisory Council on Historic Preservation and Nevada State Historic
Preservation Officer C-10
C.3 Requests for Cooperating Agency Status C-1 1
References C-16
LIST OF TABLES
Table Page
C-1 Organizations with which DOE has initiated interactions C-2
C-2 History of requests for cooperating status and similar proposals C-12
C-iii
Interagency and Intergovernmental Interactions
APPENDIX C. INTERAGENCY AND
INTERGOVERNMENTAL INTERACTIONS
In the course of producing this environmental impact statement (EIS), the U. S. Department of Energy
(DOE) has interacted with a number of governmental agencies and other organizations. These interaction
efforts have several purposes, as follows:
• Discuss issues of concern with organizations having an interest in or authority over land that the
Proposed Action (to construct, operate and monitor, and eventually close a geologic repository at
Yucca Mountain) would affect directly, or organizations having other interests that some aspect of the
Proposed Action could affect.
• Obtain information pertinent to the environmental impact analysis of the Proposed Action.
• Initiate consultations or permit processes, including providing data to agencies with oversight, review,
or approval authority over some aspect of the Proposed Action.
Section C.l summarizes the interactions. DOE has completed several efforts and will complete all
required consultations before publishing the Final EIS. Section C.2 describes interests held by agencies
and organizations involved in consultations and other interactions.
C.1 Summary of Activity
Table C-I lists organizations with which DOE has initiated interaction processes concerning the proposed
Yucca Mountain Repository and the status of those interactions.
C.2 Interests of Selected Agencies and Organizations in the Yucca
Mountain Repository Proposal
Regulations that establish a framework for interactions include 40 CFR 1502.25, which provides for
consultations with agencies having authority to issue applicable licenses, permits, or approvals, or to
protect significant resources, and 10 CFR 1021.341(b), which provides for interagency consultations as
necessary or appropriate.
C.2.1 BUREAU OF LAND MANAGEMENT
The Bureau of Land Management has a range of interests potentially affected by the Proposed Action.
The Bureau, as a part of the U.S. Department of the Interior:
• Controls a portion of the land that would need to be withdrawn by Congress to accommodate the
proposed repository
• Controls portions of land in Nevada in the five corridors for a potential branch rail line and along the
five potential routes for heavy-haul trucks
• Has responsibility for wild horse and wild burro management areas (Public Law 92-195, as amended.
Section 3; 43 CFR Part 2800) and wildlife management areas (43 CFR 24.4) in Nevada that
alternative rail corridors and routes for heavy-haul trucks cross
• Has power to grant rights-of-way and easements for transportation routes across lands it controls
C-1
Interagency and Intergovernmental Interactions
Table C-1. Organizations with which DOE has initiated interactions (page 1 of 2).
Organization
Authority/interest
Interactions
Bureau of Land
Management
U.S. Air Force
Controls part of land required for repository.
Controls portions of lands in Nevada that
transportation corridors cross. Has responsibility
for management and use of lands it controls,
including management of habitat and species. Has
data on topography, habitat, species, and other
topics on land it controls.
Controls part of land being considered for
withdrawal for repository (on the Nellis Air Force
Range) and for one Nevada rail implementing
alternative and one heavy-haul truck implementing
alternative. Has identified security concerns over
potential development of the Nevada rail and heavy-
haul truck implementing alternatives that would
pass through land it controls.
DOE provided a briefing on the EIS
during a meeting on September 15, 1998.
Naval Nuclear
Propulsion
Program
Fish and Wildlife
Service
National Marine
Fisheries Service
U.S. Department
of Transportation
DOE has provided a briefing for USAF
personnel on the process DOE is
following for this EIS and on the range
of issues being analyzed. DOE and Air
Force personnel have held informal
meetings to discuss specific issues and
update EIS status. The Air Force has
provided a statement of its concerns
regarding certain transportation
alternatives DOE is considering.
Ongoing dialogue and information
exchange.
Discussions have been held and species
list information has been obtained.
Interaction activities under the
Endangered Species Act are ongoing.
Oversees compliance with Marine Protection Discussions have been held and
Research and Sanctuaries Act and, for some species, information has been obtained,
with the Endangered Species Act. Interaction activities under the
Endangered Species Act are ongoing.
The Naval Nuclear Propulsion Program is a joint
U.S. Navy and DOE organization responsible for
management of naval spent nuclear fuel.
Oversees compliance with the Endangered Species
Act for some species and compliance with the Fish
and Wildlife Coordination Act.
Has regulatory authority over transportation of
nuclear and hazardous waste materials, including
packaging design, manufacture and use, pickup,
carriage, and receipt, and highway route selection.
U.S. Has regulatory authority over radiological standards
Environmental and groundwater protection standards. Mandatory
Protection Agency role in review of EIS adequacy.
U.S. Nuclear Required by NWPA to adopt Yucca Mountain
Regulatory Repository EIS to the extent practicable with the
Commission issuance by the Commission of any construction
authorization and license for a repository. Has
licensing authority over spent nuclear fuel and high-
level radioactive waste geologic repositories. Has
licensing authority over spent nuclear fuel and high-
level radioactive waste geologic repositories. Has
regulatory authority over commercial nuclear power
plants, storage of spent nuclear fuel at commercial
sites, and packaging for transportation of spent
nuclear fuel and high-level radioactive waste. Has
general authority over possession and transfer of
radioactive material.
EIS status briefing has been provided.
DOE and DOT have held informal
discussions concerning modeling
techniques and analytical methods DOE
is using in its evaluation of
transportation issues.
DOE and EPA have held a meeting at
which DOE provided a briefing on its
approach to the EIS and on scope and
content. At this meeting, EPA described
its EIS rating process and personnel from
the two agencies discussed methods for
addressing any EIS comments that EPA
may submit.
Discussions have been held on the
purpose and need for the action and on
the status of the EIS. Numerous
interactions related to the potential
repository program in general.
C-2
Interagency and Intergovernmental Interactions
Table C-1. Organizations with which DOE has initiated interactions (page 2 of 2).
Organization
Authority/interest
Interactions
U.S. Army Corps
of Engineers
U.S. Department
of Agriculture
Native American
Tribes
Affected units of
local government
National Park
Service
Advisory Council
on Historic
Preservation and
Nevada State
Historic
Preservation
Officer
State of Nevada
Department of
Transportation
Has authority over activities that discharge dredge
or fill material into waters of the United States.
Responsible for protection of prime farm lands for
agriculture in areas potentially affected by the
Proposed Action.
Have concern for potential consequences of
repository development and transportation
activities on cultural resources, traditions, and
spiritual integrity of the land. Have governmental
status. All interactions required for the American
Indian Religious Freedom Act, the Native
American Graves Protection and Repatriation Act,
and the National Historic Preservation Act are
being accomplished.
Local governments with general jurisdiction over
regions or communities that could be affected by
implementation of the Proposed Action.
Potential for proposal to affect water supply in
Death Valley region. Effect of any water
appropriation required for repository, EIS status,
and approach to EIS development.
Protection and preservation of historic properties
and cultural resources of importance to Native
Americans and others. Administration of the
National Historic Preservation Act and of
regulatory requirements supporting that act.
Has authority over transportation and highways
in Nevada.
Discussed strategies for minimizing
impacts and obtaining permits for waters
of the United States.
Letter exchange has resolved issues
regarding repository's potential effect on
farmlands. Need for additional
interaction is uncertain.
Ongoing discussions on a range of topics
at least twice per year. Tribal
representatives have prepared and
submitted the American Indian
Perspectives on the Yucca Mountain Site
Characterization Project and the
Repository Environmental Impact
Statement (AIWS 1998, all).
Meetings that include discussions,
information exchange, and status
briefings.
Discussion completed. National Park
Service concerns in regard to use of
water for repository construction and
operation were addressed.
Following discussions among DOE, the
Advisory Council on Historic
Preservation, and the Nevada State
Historic Preservation Officer, DOE and
the Advisory Council on Historic
Preservation have entered into a
programmatic agreement (DOE 1988,
all) establishing procedures DOE is to
follow during site characterization and
during the Secretary of Energy's
development of a repository site
recommendation. The Advisory Council
on Historic Preservation indicated that it
would be available to assist DOE in
complying with environmental review
requirements for historic properties.
DOE and Nevada Department of
Transportation personnel have had
informal discussions on Nevada
transportation issues. The State of
Nevada has requested a formal briefing
on this draft EIS after DOE publishes the
document. DOE has agreed to provide a
briefing to the state.
The Bureau of Land Management would have a continuing interest in the development of a repository at
Yucca Mountain and associated transportation routes in the State of Nevada. Any comments from the
Secretary of the Interior on the EIS must be included in the Secretary of Energy's recommendations to the
President on the Yucca Mountain site.
Interaction
DOE held a meeting with the Bureau of Land Management on September 15, 1998.
C-3
Interagency and Intergovernmental Interactions
C.2.2 U.S. AIR FORCE
The U.S. Air Force operates Nellis Air Force Base northeast of Las Vegas, and the NelHs Air Force
Range, which occupies much of south-central Nevada. The Nellis Range is an important facility for
training American and Allied combat pilots and crews (USAF 1999, pages 1-1 and 1-3).
A portion of the land being considered for withdrawal for the proposed repository is on the Nellis Range.
If the land were withdrawn and development of the proposed repository proceeded, the Air Force would
hold a continuing interest in the potential for construction, operation and monitoring, and closure
activities at the repository to have consequences for Air Force operations on the adjoining land.
One Nevada rail implementing alternative and one Nevada heavy-haul truck implementing alternative that
DOE is evaluating for the transportation of spent nuclear fuel and high-level radioactive waste would pass
through a portion of the Nellis Range, for which the Air Force has national security concerns.
Interaction
DOE provided a briefing for USAF personnel on the process DOE is following for this EIS and on the
range of issues being analyzed. DOE and Air Force persoimel have held informal meetings to discuss
specific issues. The Air Force has provided a statement of concerns about certain transportation
alternatives DOE considered in the EIS.
C.2.3 NAVAL NUCLEAR PROPULSION PROGRAM
The Naval Nuclear Propulsion Program is a joint U.S. Navy and DOE program responsible for all matters
pertaining to naval nuclear propulsion (USN 1996, page 2-2). This program is responsible for the nuclear
propulsion plants aboard more than 93 nuclear-powered warships with more than 108 reactors and for
nuclear propulsion work performed at four naval shipyards and two private shipyards. It is also
responsible for two government-owned, contractor-operated laboratories, two moored training ships, two
land-based prototype reactors, and the Expended Core Facility at the Naval Reactors Facility at the Idaho
National Engineering and Environmental Laboratory.
The Naval Nuclear Propulsion Program manages naval spent fuel after its withdrawal from nuclear-
powered warships and prototype reactors at the Expended Core Facility. The program has conducted
studies and performed environmental impact analyses on the management and containerization of naval
spent nuclear fuel to prepare it for shipment to the proposed repository or other spent fuel management
system (USN 1996, all). Information from these studies is relevant to the containerization of other spent
nuclear fuel that could be shipped to the proposed repository.
Interaction
Since the beginning of preparations for this EIS, the Naval Nuclear Propulsion Program has participated
in quarterly meetings with DOE to discuss information relevant to the emplacement of naval spent
nuclear fuel in a monitored geologic repository. Detailed information about naval spent nuclear fuel is
classified; therefore, the Naval Nuclear Propulsion Program performed a parallel set of thermal, nuclear,
and dose calculations and provided unclassified results to DOE for inclusion in this EIS. In some cases
DOE used those results as input parameters for additional analyses. Representatives of the program
participated throughout the review process to ensure the accurate presentation of information on naval
spent nuclear fuel.
C-4
Interagency and Intergovernmental Interactions
C.2.4 FISH AND WILDLIFE SERVICE
The Fish and Wildlife Service, a bureau of the U.S. Department of the Interior, has a role in the overall
evaluation of the impacts from the Proposed Action under consideration in the repository EIS. Under
the Endangered Species Act of 1973, as amended, the Fish and Wildlife Service has responsibility to
determine if projects such as the proposed Yucca Mountain Repository would have an adverse impact on
endangered or threatened species or on species proposed for listing. Any comments from the Secretary of
the Interior on the EIS must accompany the Secretary of Energy's recommendation to the President on the
Yucca Mountain site.
No endangered or proposed species occur on lands that would be needed for the repository. The desert
tortoise is the only threatened species known to exist on this land, which lies at the northern edge of the
range for desert tortoises (Buchanan 1997, pages 1 to 4). The repository would not need or impact any
critical habitat.
To evaluate the potential for the proposed repository to affect the desert tortoise, DOE and the Fish and
Wildlife Service are following a process that, in surrmiary, includes three steps:
1 . DOE submits a study (biological assessment) containing information on desert tortoise activities and
habitat in the vicinity of the proposed project, a description of project activities that could affect the
desert tortoise, and the potential for adverse impacts to desert tortoises or habitat. Based on this
information, DOE will determine if the project would result in adverse impacts to the species.
2. DOE and the Fish and Wildlife Service will meet as necessary to discuss details of the potential for
interaction between desert tortoises and project activities, and to consider appropriate protective
measures DOE could take to reduce the potential for project impact to desert tortoises.
3. The Fish and Wildlife Service will issue a biological opinion that states its opinion on whether the
proposed project may proceed without causing adverse impacts to the desert tortoise, jeopardizing the
continued existence of the species, or resulting in harassment, harm, or death of individual animals.
The biological opinion may contain protective measures and conditions that DOE would have to
implement during construction, operation and monitoring, and closure of the proposed repository to
minimize adverse impacts and the potential for tortoise deaths.
DOE, which has conducted site characterizations at Yucca Mountain since 1986, and the Fish and
Wildlife Service have conducted previous consultation processes that addressed the potential for site
characterization activities to affect the desert tortoise. These processes resulted in biological opinions,
published in 1990 and 1997, that determined that site characterization activities could proceed without
unacceptable harm to the desert tortoise and that the protective measures and conditions stated in the
biological opinions should apply to DOE activities. None of the proposed repository land is critical
habitat for tortoises. The current consultation process on the desert tortoise will build on the information
gathered and the practices developed in the previous consultations, and on the positive results obtained.
Interaction
DOE is currently preparing a Biological Assessment to be submitted to the Fish and Wildlife Service.
C.2.5 NATIONAL MARINE FISHERIES SERVICE
The National Marine Fisheries Service exercises protective jurisdiction over aspects of the marine
environment, including research activities, marine sanctuaries, and certain species protected by the
Endangered Species Act. Potential DOE actions associated with transportation to the repository (for
C-5
Interagency and Intergovernmental Interactions
example, barging and construction or modification of bridges and docking facilities) could require
interaction with the National Marine Fisheries Service.
Interaction
DOE participated in an informal discussion that identified National Marine Fisheries Service jurisdiction
relevant to the Yucca Mountain Project and potential project activities of jurisdictional interest to the
National Marine Fisheries Service in fulfilling its responsibilities.
C.2.6 U.S. DEPARTMENT OF TRANSPORTATION
The U.S. Department of Transportation has the authority to regulate several aspects of the transportation
of spent nuclear fuel and high-level radioactive waste to the proposed Yucca Mountain Repository. The
general authority of the Department of Transportation to regulate carriers and shippers of hazardous
materials includes packaging procedures and practices, shipping of hazardous materials, routing, carrier
operations, equipment, shipping container construction, and receipt of hazardous materials (49 USC 1801;
49 CFR Parts 171 through 180).
Interaction
DOE and the Department of Transportation have exchanged letters and informal communications
on topics pertaining to the proposed Yucca Mountain Project that are within the Department of
Transportation's regulatory interest. DOE and the Department of Transportation have held informal
discussions on the modeling techniques and analytical methods DOE used in its evaluation of
transportation issues.
C.2.7 U.S. ENVIRONMENTAL PROTECTION AGENCY
The U.S. Environmental Protection Agency has two primary responsibilities in relation to the proposed
Yucca Mountain Repository. It is responsible for promulgating regulations that set radiological
protection standards for media that would be affected if radionuclides were to escape the confinement
of the repository. In addition, the Agency oversees the National Environmental Policy Act process for
Federal EISs. Council on Environmental Quality regulations implementing the National Environmental
Policy Act specify procedures that agencies must follow and actions that agencies must take in preparing
EISs. Depending on the level of concern that the Agency might have with environmental aspects of
the Yucca Mountain Project Draft EIS, it can initiate a consultation between DOE and the Council on
Environmental Quality. The Secretary of Energy's recommendation to the President must include both
the Final EIS and the Environmental Protection Agency's comments on the EIS.
Interaction
DOE and the Environmental Protection Agency held a meeting at which DOE provided a briefing on its
approach to the EIS and its scope and content. At that meeting, the Environmental Protection Agency
described its EIS rating process, and personnel from the two agencies discussed methods for addressing
EIS comments that the Agency might submit.
C.2.8 U.S. NUCLEAR REGULATORY COMMISSION
The Nuclear Waste Policy Act (42 USC 10101 et seq.) establishes a multistep procedure for reviews
and decisions on the proposal to construct, operate and monitor, and close a geologic repository at
Yucca Mountain. The final steps in this procedure require DOE to make an application to the
U.S. Nuclear Regulatory Commission for authorization to construct a repository at Yucca Mountain and
the Commission to consider this information and make a final decision within 3 years on whether to
approve the application. The Nuclear Waste Policy Act directs the Commission to adopt this EIS to the
C-6
Interagency and Intergovernmental Interactions
extent practicable in support of its decisionmaking process. Any Nuclear Regulatory Commission
comment on this EIS must accompany the Secretary of Energy's recommendation to the President.
The Nuclear Regulatory Commission also has authority under the Atomic Energy Act of 1954, as
amended, to regulate persons authorized to own, possess, or transfer radiological materials. In
addition, the Commission regulates transportation packaging, transportation operations, and the design,
manufacture, and use of shipping containers for radiological materials with levels of radioactivity greater
than Department of Transportation Type A materials. Determination as to whether radiological materials
are Type A or greater are made in accordance with a procedure set forth in 49 CFR 173.431.
Interaction
Discussions have been held on the purpose and need for the Proposed Action and on the status of the EIS.
Interactions with the Nuclear Regulatory Commission will include those necessary to process any
application to construct a repository at Yucca Mountain.
C.2.9 U.S. ARMY CORPS OF ENGINEERS
The Clean Water Act of 1977 (42 USC 1251 et seq.) gives the U.S. Army Corps of Engineers permitting
authority over activities that discharge dredge or fill material into waters of the United States. If DOE
activities associated with a repository at Yucca Mountain discharged dredge or fill into any such waters,
DOE could need to obtain a permit from the Corps. The construction or modification of rail lines or
highways to the repository would also require Section 404 permits if those actions included dredge and
fill activities or other activities that would discharge dredge or fill into waters of the United States. DOE
has obtained a Section 404 permit for site characterization-related construction activities it might conduct
in Coyote Wash or its tributaries or in Fortymile Wash.
interaction
Strategies for minimizing any impacts and obtaining permits have been discussed.
0.2.10 U.S. DEPARTIVIENT OF AGRICULTURE
The U.S. Department of Agriculture has the responsibility to ensure that the potential for Federal
programs to contribute to unnecessary and irreversible conversion of farmlands to nonagricultural uses is
kept to a minimum. Proposed Federal projects must obtain concurrence from the Natural Resource
Conservation Service of the Department of Agriculture that potential activities would not have
unacceptable effects on farmlands (7 USC 4201 et seq.).
Interaction
DOE has had written communication with the Department of Agriculture. The process has resulted in a
concurrence that a repository at Yucca Mountain would not affect farmlands.
C.2.11 NATIVE AI\/IERICAN TRIBES
Many tribes have historically used the area being considered for the proposed Yucca Mountain
Repository, as well as nearby lands (AIWS 1998, page 2-1). The region around the site holds a range of
cultural resources and animal and plant resources. Native American tribes have concerns about the
protection of cultural resources and traditions and the spiritual integrity of the land. Tribal concerns
extend to the propriety of the Proposed Action, the scope of the EIS, and opportunities to participate in
the EIS process, as well as issues of environmental justice and the potential for transportation impacts
(AIWS 1998, pages 2-2 to 2-26, and 4-1 to 4-12). Potential rail and legal-weight truck routes would
follow existing rail lines and highways, respectively. The legal-weight truck route would pass through
C-7
Interagency and Intergovernmental Interactions
the Moapa Indian Reservation and the potential rail line would pass near the Reservation. Potential routes
for legal-weight and heavy-haul trucks would follow existing highways, and would pass through the Las
Vegas Paiute hidian Reservation.
DOE Order 1230.2 recognizes that Native American tribal governments have a special and unique legal
and political relationship with the Government of the United States, as defined by history, treaties,
statutes, court decisions, and the U.S. Constitution. DOE recognizes and commits to a govemment-to-
govemment relationship with Native American tribal governments. DOE recognizes tribal governments
as sovereign entities with, in most cases, primary authority and responsibility for Native American
territory. DOE recognizes that a trust relationship derives from the historic relationship between the
Federal Government and Native American tribes as expressed in certain treaties and Federal law. DOE
has and will consult with tribal governments to ensure that tribal rights and concerns are considered
before taking actions, making decisions, or implementing programs that could affect tribes. These
interactions ensure compliance with provisions of the American Indian Religious Freedom Act (42 USC
1996 et seq.), the Native American Graves Protection and Repatriation Act (25 USC 3001 et seq.), DOE
Order 1230.2 {American Indian Tribal Government Policy), Executive Order 13007 {Sacred Sites),
Executive Order 13084 {Consultation and Coordination with Indian Tribal Governments), and the
National Historic Preservation Act (16 USC 470f).
Interaction
The Native American Interaction Program was formally begun in 1987. Representatives from the
Consolidated Group of Tribes and Organizations have met in large group meetings twice yearly with
DOE on a range of cultural and other technical concerns. Additionally, specialized Native American
subgroups have been periodically convened to interact with DOE on specific tasks including ethnobotany,
review of artifact collections, field archaeological site monitoring, and the EIS process.
The Consolidated Group of Tribes and Organizations consists of the following:
• Southern Paiute
Kaibab Paiute Tribe, Arizona
Paiute Indian Tribes of Utah
Moapa Band of Paiutes, Nevada
Las Vegas Paiute Tribe, Nevada
Pahrump Paiute Tribe, Nevada
Chemehuevi Paiute Tribe, California
Colorado River Indian Tribes, Arizona
• Western Shoshone
Duckwater Shoshone Tribe, Nevada
Ely Shoshone Tribe, Nevada
Yomba Shoshone Tribe, Nevada
Timbisha Shoshone Tribe, California
• Owens Valley Paiute and Shoshone
Benton Paiute Tribe, California
Bishop Paiute Tribe, California
Big Pine Paiute Tribe, California
Lone Pine Paiute Tribe, California
Fort Independence Paiute Tribe, California
• Other Official Native American Organizations
Las Vegas Indian Center, Nevada
C-8
Interagency and Intergovernmental Interactions
Tribal representatives have prepared and submitted the American Indian Perspectives on the Yucca
Mountain Site Characterization Project and the Repository Environmental Impact Statement (AIWS
1998, all). This document discusses site characterization at Yucca Mountain and the Proposed Action in
the context of Native American culture, concerns, and views and beliefs concerning the surrounding
region. It has been used as a resource in the preparation of the EIS; excerpts are presented in Chapter 4,
Section 4.1.13.4, to reflect a Native American point of view. The issues discussed ranged from traditional
resources to concerns related to the potential repository.
C.2.12 AFFECTED UNITS OF LOCAL GOVERNMENT
As defined by the NWPA, the affected units of local government are local governments (counties) with
jurisdiction over the site of a repository. Concerns of the affected units of local government range from
socioeconomic impacts to potential consequences of transportation activities. Nye County, Nevada, in
which DOE would build the repository, is one of the affected units of local government. Others include
Clark, Lincoln, Esmeralda, Mineral, Churchill, Lander, Eureka, White Pine, and Elko Counties in Nevada
and Inyo County in California.
DOE has offered local governments the opportunity to submit documents providing perspectives of
issues associated with the EIS. At Draft EIS publication, Nye County had prepared such a document.
In addition, other documents related to the Yucca Mountain region have been prepared in the past by
several local government units including Clark, Lincoln, and White Pine Counties.
Interaction
DOE has held formal meetings twice a year with the affected units of local government. These meetings
have included discussions and status briefings on a range of issues of interest to local governments. DOE
has also held numerous informal meetings with representatives. Documents have been received from
units of local government.
C.2.13 NATIONAL PARK SERVICE
The National Park Service, which is a bureau of the U.S. Department of the Interior, is responsible for
the management and maintenance of the Nation's national parks and monuments. The implementation
of the Proposed Action could potentially affect the water supply in Death Valley National Park, which is
downgradient from Yucca Mountain. The National Park Service, therefore, would have an interest in any
water appropriation granted to DOE for the repository. In addition, the Park Service has expressed its
interest in this EIS, its status, and the approach DOE has followed in developing the EIS.
Interaction
DOE and National Park Service representatives held a discussion during which they addressed Park
Service concerns about water use for repository construction and operation.
C.2.14 STATE OF NEVADA
If DOE receives authorization to construct, operate and monitor, and eventually close a geologic
repository at Yucca Mountain, DOE would need to obtain a range of permits and approvals from the State
of Nevada. DOE would need to coordinate application processing activities with the State to complete
the permitting processes. DOE could require permits or approvals such as the following:
• An operating permit for control of gaseous, liquid, and particulate emissions associated with
construction and operation
• A public water system permit and a water system operating permit for provision of potable water
C-9
Interagency and Intergovernmental Interactions
•
•
A general permit for storm-water discharge
A National Pollutant Discharge Elimination System permit for point source discharges to waters of
the State
A hazardous materials storage permit to store, dispense, use, or handle hazardous materials
• A permit for a sanitary and sewage collection system
• A solid waste disposal permit
• Other miscellaneous permits and approvals
DOE required similar permits and approvals from the State of Nevada to conduct site characterization
activities at Yucca Mountain. DOE and the State coordinated on a range of activities, including an
operating permit for surface disturbances and point source emissions, an Underground hijection Control
Permit and a Public Water System Permit, a general discharge permit for effluent discharges to the
ground surface, a permit for the use of groundwater, a permit from the State Fire Marshal for the storage
of flammable materials, and a permit for operation of a septic system. DOE could apply for additional or
expanded authority under the existing permits, where needed, if provisions for expansion became
applicable. DOE or its contractors could also need to coordinate transportation activities, highway uses,
and transportation facility construction and maintenance activities with the Nevada Department of
Transportation.
Interaction
The State of Nevada has requested a formal briefing on this Draft EIS after its publication, and DOE has
agreed to provide the briefing. DOE and the Nevada Department of Transportation personnel have had
information discussions on Nevada transportation issues.
C.2.15 ADVISORY COUNCIL ON HISTORIC PRESERVATION AND NEVADA STATE
HISTORIC PRESERVATION OFFICER
In the mid- to late- 1980s, DOE, the Nevada State Historic Preservation Officer and the Advisory Council
on Historic Preservation discussed the development of a Programmatic Agreement to address DOE
responsibilities under Sections 106 and 1 10 of the National Historic Preservation Act and the Council's
implementing regulations. These discussions led to a Programmatic Agreement between DOE and the
Advisory Council on Historic Preservation (DOE 1988, all) that records stipulations and terms to resolve
potential adverse effects of DOE activities on historic properties at Yucca Mountain. The activities
covered by the Agreement include site characterization of the Yucca Mountain site under the NWPA and
the DOE recommendation to the President on whether or not to develop a repository, informed by a final
EIS prepared pursuant to the National Environmental Policy Act and the NWPA.
Although not a formal signatory, the Nevada State Historic Preservation Officer has the right at any time,
on request, to participate in monitoring DOE compliance with the Programmatic Agreement. In addition,
DOE must provide opportunities for consultations with the Advisory Council on Historic Preservation,
the Nevada State Historic Preservation Officer, and Native American tribes as appropriate throughout the
process of implementing the Agreement. DOE submits an annual report to the Advisory Council and the
Nevada State Historic Preservation Officer describing the activities it conducts each year to implement
the stipulations of the Programmatic Agreement. This report includes a description of DOE coordinations
and consultations with Federal and State agencies and Native American Tribes on historic and culturally
significant properties at Yucca Mountain.
C-10
Interagency and Intergovernmental Interactions
DOE will continue to seek input from the Nevada State Historic Preservation Officer and the Advisory
Council on Historic Preservation, and will interact appropriately to meet the reporting and other
stipulations of the Programmatic Agreement.
interaction
DOE has submitted annual reports to the Nevada State Historic Preservation Officer and the Advisory
Council on Historic Preservation and has provided opportunities for consultations with agencies and
Native American Tribes as appropriate in accordance with the terms of the Programmatic Agreement.
C.3 Requests for Cooperating Agency Status
This EIS addresses a range of potential activities that are of potential concern to other agencies and to
Native Americans. Governmental agencies and Native American tribes participated in the EIS process by
submitting scoping comments and may submit comments on this Draft EIS. Representatives of Native
American tribes have submitted a document that provides their perspective on the Proposed Action.
Moreover, DOE has invited local governments in Nevada to submit reference documents providing
information on issues of concern.
DOE is the lead agency for this EIS. Regulations of the Council on Environmental Quality allow the lead
agency to request any other Federal agency that has jurisdiction by law or special expertise regarding any
environmental impact involved in a proposal (or a reasonable alternative) to be a cooperating agency for
an EIS (40 CFR 1501.6 and 1508.5). The regulations also allow another Federal agency to request that
the lead agency designate it as a cooperating agency. Finally, the regulations allow state or local agencies
of similar qualifications or, when the effects are on a reservation, a Native American Tribe, by agreement
with the lead agency to become a cooperating agency (40 CFR 1508.5). Table C-2 lists requests for
cooperating agency status and other proposals.
If the lead agency designates a cooperating agency, the lead agency's duties toward the cooperating
agency include the following:
• Requesting early participation in the National Environmental Policy Act (that is, EIS) process
• Using any environmental analysis or proposal provided by a cooperating agency with legal
jurisdiction or special expertise to the greatest extent possible consistent with its responsibilities as a
lead agency
• Meeting with a cooperating agency when the cooperating agency requests
A cooperating agency's duties include the following:
• Participating early in the National Environmental Policy Act process
• Participating in the scoping process
• If requested by the lead agency, assuming responsibility for developing information and preparing
environmental analyses including portions of the EIS for which the cooperating agency has special
expertise
• If the lead agency requests, making staff support available
• Using its own funds, except the lead agency is to fund major activities or analyses it requests to the
extent available
C-li
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C-15
Interagency and Intergovernmental Interactions
Several agencies, tribes, or tribal organizations have either requested cooperating agency status for this
EIS, made comparable proposals for participation, or stated positions in regard to the extent of their
participation. Table C-2 summarizes agency requests, proposals, and position statements together with
the DOE responses, if appropriate.
REFERENCES
AIWS 1998
Barnes 1995a
Barnes 1995b
Barnes 1995c
Barnes 1996
Barrett 1998
Benson 1996
Boland 1996
AIWS (American Indian Writers Subgroup), 1998, American Indian
Perspectives on the Yucca Mountain Site Characterization Project and
the Repository Environmental Impact Statement, American Indian
Resource Document, Consolidated Group of Tribes and Organizations,
Las Vegas, Nevada. [MOL. 19980420.0041]
Barnes, W. E., 1995a, "Nye County's Request for Cooperating Agency
Designation," letter to The Honorable Cameron McRae (Nye County
Commissioners, Tonopah, Nevada), November 21, Office of Civilian
and Radioactive Waste Management, U.S. Department of Energy, Las
Vegas, Nevada. [MOL. 19960424.0182]
Barnes, W. E., 1995b, letter to L. Bradshaw (Nuclear Waste Repository
Project Office, Tonopah, Nevada), December 1, Office of Civilian and
Radioactive Waste Management, U.S. Department of Energy, Las
Vegas, Nevada. [MOL. 19960425.03 10]
Barnes, W. E., 1995c, "Proposed Memorandum Of Understanding
(MOU) Regarding The U.S. Department Of Energy's (DOE)
Preparation Of An Environmental Impact Statement (EIS) For A
Potential Repository At Yucca Mountain, Nevada," letter to J. Regan
(Office of the Churchill County Commissioners, Fallon Nevada), July
21, Office of Civilian and Radioactive Waste Management, U.S.
Department of Energy, Las Vegas, Nevada. [MOL. 1995 1220.0136]
Barnes, W. E., 1996, letter to R. F. Boland (The Timbisha Shoshone -
Death Valley Land Restoration Project, Death Valley, California),
November 12, Office of Civilian Radioactive Waste Management, U.S.
Department of Energy, Las Vegas, Nevada. [MOL.19970210.0099]
Barrett, L. H., 1998, letter to L. W. Bradshaw (Department of Natural
Resources and Federal Facilities, Nuclear Waste Repository Project
Office, Pahrump, Nevada), September 24, Office of Civilian and
Radioactive Waste Management, U.S. Department of Energy,
Washington, D.C. [MOL. 199906 10.0300]
Benson, A. B., 1996, letter to The Honorable Edward E. Wright
(Lincoln County Commissioner), August 2, Office of Public Affairs,
Office of Civilian Radioactive Waste Management, U.S. Department of
Energy, Las Vegas, Nevada. [MOL. 19961 1 15.0045]
Boland, R. F., 1996, "Yucca Mountain High Level Nuclear Waste
Depository Siting In Nevada Threatens Native American Cultural
Resources And Adversely Affects Public Health and Safety," letter to
W. J. Clinton (President of the United States), August 14, The Timbisha
Shoshone - Death Valley Land Restoration Project, Death Valley,
California. [HQO. 19961 112.0018]
C-16
Interagency and Intergovernmental Interactions
Bradshaw 1995
Bradshaw 1998
Buchanan 1997
Buraell 1996
Dixon 1995a
Dixon 1995b
Dixon 1996
DOE 1988
Bradshaw, L. W., 1995, letter to Dr. D. Dreyfus (Office of Civilian and
Radioactive Waste Management, U.S. Department of Energy), October
4, Nuclear Waste Repository Project Office, Tonopah, Nevada.
[MOL. 199903 19.0217]
Bradshaw, L. W., 1998, "Request for Cooperating Agency Status in the
Preparation of the Yucca Mountain (YM) Environmental Impact
Statement (EIS)," letter to L. Barrett (Office of Civilian and
Radioactive Waste Management, U.S. Department of Energy,
Washington D.C.), July 30, Department of Natural Resources and
Federal Facilities, Nuclear Waste Repository Project Office, Pahrump,
Nevada. [MOL. 19980903.0847]
Buchanan, C. C, 1997, "Final Biological Opinion for Reinitiation of
Formal Consultation for Yucca Mountain Site Characterization
Studies," letter to W. Dixon (U.S. Department of Energy, Yucca
Mountain Site Characterization Office), File No. 1-5-96-F-307R, Fish
and Wildlife Service, U.S. Department of the Interior, Nevada State
Office, Reno, Nevada. [MOL. 19980302.0368]
Bumell, J. R., 1996, letter to J. Chirieleison (Office of Civilian
Radioactive Waste Management, U.S. Department of Energy), June 19,
Council of Energy Resource Tribes, Denver, Colorado.
[MOL. 19961002.0379, letter; MOL. 19961002.0380, concept paper]
Dixon, W. R., 1995a, "Proposal To Participate as A Cooperating
Agency In The Yucca Mountain Site Characterization Office's
(YMSCO) Preparation Of An Environmental Impact Statement (EIS)
For A Potential Repository At Yucca Mountain, Nevada," interoffice
letter to R. Guida, (Office of Naval Reactors), July 10, Office of
Civilian and Radioactive Waste Management, U.S. Department of
Energy, Las Vegas, Nevada. [MOL. 19990610.0298]
Dixon, W. R., 1995b, "Letter Requesting Cooperating Agency
Involvement In The Repository Environmental Impact Statement
(EIS)," letter to R. Martin, (Death Valley National Park, National Park
Service, U.S. Department of the Interior), November 14, Office of
Civilian and Radioactive Waste Management, U.S. Department of
Energy, Las Vegas, Nevada. [MOL. 19960419.0246]
Dixon, W. R., 1996, letter to J. Bumell (Council of Energy Resource
Tribes), July 26, Office of Civilian Radioactive Waste Management,
U.S. Department of Energy, Las Vegas, Nevada.
[MOL. 19961015.0306]
DOE (U.S. Department of Energy), 1988, Programmatic Agreement
Between the United States Department of Energy and the Advisory
Council on Historic Preservation for the Nuclear Waste Deep Geologic
Repository Program, Yucca Mountain, Nevada, Yucca Mountain Site
Characterization Office, Nevada Operations Office, North Las Vegas,
Nevada. [HQX. 19890426.0057]
C-17
Interagency and Intergovernmental Interactions
DOE 1997
Esmond 1997
Gaiashkibos 1995
Guida 1995
Holonich 1995
Martin 1995
McRae 1995
Nissley 1995
Regan 1995
DOE (U.S. Department of Energy), 1997, Summary of Public Scoping
Comments Related to the Environmental Impact Statement for a
Geologic Repository for the Disposal of Spent Nuclear Fuel and High-
Level Radioactive Waste at Yucca Mountain, Nye County, Nevada,
Office of Civilian Radioactive Waste Management, Yucca Mountain
Site Characterization Office, North Las Vegas, Nevada.
[MOL. 1997073 1.05 15]
Esmond, M. R., Major General, USAP, 1997, letter to R. Loux (Agency
for Nuclear Projects, Nevada Nuclear Waste Project Office), September
4, Department of the Air Force, Nellis Airforce Base, Nevada.
[MOL. 19971 124.0417]
Gaiashkibos, 1995, letter to H. O'Leary (U.S. Department of Energy),
March 1, National Congress of American Indians, Washington, D.C.
[MOL. 19990610.0304]
Guida, R. A., 1995, "Comments On Notice Of Intent For Repository
EIS," interoffice letter to L. Barrett (Office of Civilian and Radioactive
Waste Management), May 23, Office of Naval Reactors, U.S.
Department of Energy, Washington, D.C. [HQO. 199507 12.0020]
Holonich, J. J., 1995, "Identification Of Lead Contact In Nuclear
Regulatory Commission's Review And Comment Of U.S. Department
Of Energy's Draft Environmental Impact Statement," letter to R. Milner
(Office of Civilian Radioactive Waste Management, U.S. Department
of Energy), March 1, High-Level Waste and Uranium Recovery
Projects Branch, Division of Waste Management, Office of Nuclear
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission,
Washington, D.C. [MOL. 1 99906 1 0.030 1 ]
Martin, R. H., 1995, letter to W. Dixon (Office of Civilian and
Radioactive Waste Management, U.S. Department of Energy),
September 21, Death Valley National Park, National Park Service, U.S.
Department of the Interior, Death Valley, California.
[MOL. 199603 12.0266]
McRae, C, 1995, "Cooperating Agency Designation for Nye County in
the Preparation of the Yucca Mountain Environmental Impact
Statement (EIS)," letter to D. Dreyfus (Office of Civilian and
Radioactive Waste Management, U.S. Department of Energy), August
15, Nye County Commission, Tonopah, Nevada.
[MOL. 19960321.03 19]
Nissley, C, 1995, letter to W. Dixon (Office of Civilian Radioactive
Waste Management, U.S. Department of Energy), October 12,
Advisory Council on Historic Preservation, Washington, D.C.
[MOL. 19990319.0206]
Regan, J., 1995, letter to M. Powell (U.S. Department of Energy), May
30, Office of the Churchill County Commissioners, Fallon, Nevada.
[MOL.19990610.0299]
C-18
Interagency and Intergovernmental Interactions
Stablein 1997
USAF 1999
USN 1996
Wright 1996
Stablein, N. K., 1997, "Information On Naval Spent Fuel Request,"
letter to R. Guida, (Naval Nuclear Propulsion Program, U.S.
Department of the Navy), August 22, Office of Nuclear Material Safety
and Safeguards, U.S. Nuclear Regulatory Commission, Washington,
D.C. [MOL.19990610.0302]
USAF (U.S. Air Force), 1999, Renewal of the Nellis Air Force Range
Land Withdrawal: Legislative Environmental Impact Statement, Air
Combat Command, U.S. Department of the Air Force, U. S.
Department of Defense, NeUis Air Force Base, Nevada. [243264]
USN (U.S. Navy), 1996, Department of the Navy Final Environmental
Impact Statement for a Container System for the Management of Naval
Spent Nuclear Fuel, DOE/EIS-0251, in cooperation with the U.S.
Department of Energy, Naval Nuclear Propulsion Program, U.S.
Department of the Navy, U.S. Department of Defense, Arlington,
Virginia. [227671]
Wright, E. E., 1996, "Proposal for Lincoln County to Provide input into
DOE's Preliminary Transportation Strategies," letter to W. Barnes
(Office of Civilian Radioactive Waste Management, U.S. Department
of Energy), April 22, Lincoln County Board of County Commissioners,
Pioche, Nevada. [MOL. 1 9960905 .0 149]
C-19
r
rorrMn^
Appendix D
Distribution List
Distribution List
APPENDIX D. DISTRIBUTION LIST
DOE is providing copies of the Draft EIS to Federal, state, and local elected and appointed officials and
agencies of government; Native American groups; national, state, and local environmental and public
interest groups; and other organizations and individuals listed below. Copies will be provided to other
interested parties upon request.
A. United States Congress
A.1 SENATORS FROM NEVADA
The Honorable Harry Reid
United States Senate
The Honorable Richard H. Bryan
United States Senate
A.2 UNITED STATES SENATE COMMITTEES
The Honorable Pete V. Dominici
Chairman
Subcommittee on Energy and Water
Development
Committee on Appropriations
The Honorable John Warner
Chairman
Committee on Armed Services
The Honorable Harry Reid
Ranking Minority Member
Subcommittee on Energy and Water
Development
Committee on Appropriations
The Honorable Carl Levin
Ranking Minority Member
Committee on Armed Services
The Honorable Frank H. Murkowski
Chairman
Committee on Energy and Natural
Resources
The Honorable Jeff Bingaman
Ranking Minority Member
Committee on Energy and Natural
Resources
A.3 UNITED STATES REPRESENTATIVES FROM NEVADA
The Honorable Jim Gibbons
United States House of Representatives
The Honorable Shelley Berkley
United States House of Representatives
A.4 UNITED STATES HOUSE OF REPRESENTATIVES COMMITTEES
The Honorable Ron Packard
Chairman
Subcommittee on Energy and Water
Development
Committee on Appropriations
The Honorable Floyd D. Spence
Chairman
Committee on Armed Services
The Honorable Tom Bliley
Chairman
Committee on Commerce
The Honorable Joe Barton
Chairman
Subcommittee on Energy and Power
Committee on Commerce
D-1
Distribution List
The Honorable Don Young
Chairman
Committee on Resources
The Honorable John D. Dingell
Ranking Minority Member
Committee on Commerce
The Honorable Bud Shuster
Chairman
Committee on Transportation and Infrastructure
The Honorable Peter J. Visclosky
Ranking Minority Member
Subcommittee on Energy and Water
Development
Committee on Appropriations
The Honorable Dee Skelton
Ranking Minority Member
Committee on Armed Services
The Honorable Ralph M. Hall
Ranking Minority Member
Subcommittee on Energy and Power
Committee on Commerce
The Honorable George Miller
Ranking Minority Member
Committee on Resources
The Honorable James L. Oberstar
Ranking Minority Member
Committee on Transportation and Infrastructure
D-2
Distribution List
B. Federal Agencies
Mr. Andrew Thibadeau
Information Officer
Defense Nuclear Facilities Safety Board
Ms. Andree DuVamey
National Environmental Coordinator
Ecological Sciences Division
Natural Resources Conservation Service
U.S. Department of Agriculture
Dr. Frank Monteferrante
Director of Compliance
Economic Development Administration
U.S. Department of Commerce
Ms. Jean Reynolds
Deputy for Environmental Planning
Office of Environment, Safety and Occupational
Health
Department of the Air Force
U.S. Department of Defense
Mr. Timothy P. Julius
Office of the Director of Environmental
Programs
Office of the Assistant Chief of Staff for
Installation Management
Department of the Army
U.S. Department of Defense
Ms. Kimberley DePaul
Head, Environmental Planning and NEPA
Compliance Program
Office of Chief of Naval Operations/N456
Department of the Navy
U.S. Department of Defense
Mr. A. Forester Einarsen
NEPA Coordinator
Office of Environmental Policy, CECW-AR-E
U.S. Army Corps of Engineers
U.S. Department of Defense
Dr. David Bodde
Chair
Environmental Management Advisory Board
Henry W. Bloch School of Business and Public
Administration
University of Missouri-Kansas City
Mr. Jim Melillo
Executive Director
Environmental Management Advisory Board
U.S. Department of Energy
Mr. Willie R. Taylor
Director
Office of Environmental Policy and Compliance
U.S. Department of the Interior
Mr. Michael Soukup
Associate Director
Natural Resource Stewardship and Science
National Park Service
U.S. Department of Interior
Mr. William Cohen
Chief
General Litigation Section
Environment and Natural Resources Division
U.S. Department of Justice
Ms. Camille Mittleholtz
Environmental Team Leader
Office of Transportation Policy
U.S. Department of Transportation
Mr. Steve Grimm
Senior Program Analyst, RRP-24
Office of Policy and Program Development
Federal Railroad Administration
U.S. Department of Transportation
Dr. Robert McGuire, DHM2
Deputy Associate Administrator Hazardous
Materials Safety
Research and Special Programs Administration
U.S. Department of Transportation
Ms. Susan Absher
U.S. Environmental Protection Agency
Office of Federal Activities
Mr. Kenneth Czyscinski
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
D-3
Distribution List
Mr. David Huber
U.S. Environmental Protection Agency
Office of Ground Water and Drinking Water
Mr. Robert Barles
U.S. Environmental Protection Agency
Office of Ground Water and Drinking Water
Mr. Dennis O'Connor
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
Carol Browner
Administrator
U.S. Environmental Protection Agency
Mr. Karl Kanbergs (CMD-2)
Region 9
U.S. Environmental Protection Agency
Betsey Higgins
Environmental Review Coordinator
Region 1
U.S. Environmental Protection Agency
Robert Hargrove
Environmental Review Coordinator
Region 2
U.S. Environmental Protection Agency
John Forren
Environmental Review Coordinator
Region 3
U.S. Environmental Protection Agency
Heinz Mueller
Environmental Review Coordinator
Region 4
U.S. Environmental Protection Agency
Sherry Kamke, Acting
Environmental Review Coordinator
Region 5
U.S. Environmental Protection Agency
Mike Jansky
Environmental Review Coordinator
Region 6
U.S. Environmental Protection Agency
Joe Cothem
Environmental Review Coordinator
Region 7
U.S. Environmental Protection Agency
Cindy Cody
Environmental Review Coordinator
Region 8
U.S. Environmental Protection Agency
Dave Tomsovic
Environmental Review Coordinator
Region 9
U.S. Environmental Protection Agency
Wayne Elson
Environmental Review Coordinator
Region 10
U.S. Environmental Protection Agency
Mr. Mark Robinson
Director, Division of Licensing and Compliance
Federal Energy Regulatory Commission
Mr. Vic Rezendes
Director, Energy, Resources, and Sciences
Issues
U.S. General Accounting Office
Mr. Lawrence Rudolph
General Counsel
National Science Foundation
The Honorable Nils J. Diaz
Commissioner
U.S. Nuclear Regulatory Commission
The Honorable Greta Joy Dicus
Acting Chairman
U.S. Nuclear Regulatory Commission
The Honorable Edward J. McGaffigan
Commissioner
U.S. Nuclear Regulatory Commission
The Honorable Jeffrey S. Merrifield
Commissioner
U.S. Nuclear Regulatory Commission
D-4
Distribution List
Mr. William C. Reamer
Division of Waste Management
U.S. Nuclear Regulatory Commission
Mr. Keith McConnell
Division of Waste Management
U.S. Nuclear Regulatory Commission
Mr. Carl Paperiello
Office of Nuclear Material Safety & Safeguards
U.S. Nuclear Regulatory Commission
David Brooks
Division of Waste Management
U.S. Nuclear Regulatory Commission
Mr. Thomas H. Essig
Acting Chief, Generic Issues and Environmental
Projects Branch
Office of Nuclear Reactor Regulation
U.S. Nuclear Regulatory Commission
Richard Major
Advisory Committee on Nuclear Waste
U.S. Nuclear Regulatory Commission
Dr. Janet Kotra
U.S. Nuclear Regulatory Commission
Mr. Neil Jensen
U.S. Nuclear Regulatory Commission
Ms. Charlotte Abrams
U.S. Nuclear Regulatory Commission
Mr. James Firth
U.S. Nuclear Regulatory Commission
Mr. Martin Virgilio
U.S. Nuclear Regulatory Commission
Mr. King Stablein
U.S. Nuclear Regulatory Commission
Mr. Dave Matthews
U.S. Nuclear Regulatory Commission
Mr. Don Cleary
U.S. Nuclear Regulatory Commission
Ms. Cindy Carpenter
U.S. Nuclear Regulatory Commission
Ms. Susan Shankman
U.S. Nuclear Regulatory Commission
Mr. William Brach
U.S. Nuclear Regulatory Commission
Mr. Robert Fairweather
Chief, Environmental Branch
Office of Management and Budget
John Ffeiffer
Budget Examiner
Office of Management and Budget
Dr. Rosina Bierbaum
Associate Director for Environment
Office of Science and Technology Policy
Executive Office of the President
Mr. Greg Askew
Senior Specialist, NEPA
Environmental Management
Tennessee Valley Authority
Mr. Jared L. Cohon, Ph.D., P.E.
Chairman
U.S. Nuclear Waste Technical Review Board
Mr. John W. Arendt P.E.
U.S. Nuclear Waste Technical Review Board
Dr. Daniel B. Bullen, Ph.D.
U.S. Nuclear Waste Technical Review Board
Mr. Norman L. Christensen, Jr.
U.S. Nuclear Waste Technical Review Board
Dr. Paul P. Craig
U.S. Nuclear Waste Technical Review Board
Dr. Deborah S. Knopman
U.S. Nuclear Waste Technical Review Board
Dr. Priscilla P. Nelson
U.S. Nuclear Waste Technical Review Board
D-5
Distribution List
Dr. Richard Parizek
U.S. Nuclear Waste Technical Review Board
Dr. Donald Runnells
U.S. Nuclear Waste Technical Review Board
Dr. Alberto A. Sagues P.E.
U.S. Nuclear Waste Technical Review Board
Dr. Jeffrey Wong
U.S. Nuclear Waste Technical Review Board
Dr. William D. Barnard, Ph.D.
Executive Director
U.S. Nuclear Waste Technical Review Board
Ms. Paula Alford
U.S. Nuclear Waste Technical Review Board
Mr. John N. Fischer
U.S. Geological Survey
Dr. Mitchell W. Reynolds
U.S. Geological Survey
Mr. James F. Devine
U.S. Geological Survey
Mr. Daniel J. Goode
U.S. Geological Survey
Ms. Helen M. Hankins
Field Office Manager
Bureau of Land Management
Elko Field Office
Mr. Ronald G. Huntsinger
Field Office Manager
Bureau of Land Management
Tonopah Field Office
Gene A. Kolkman
Field Office Manager
Bureau of Land Management
Ely Field Office
Mr. George Meckfessel
Planning and Environmental Coordinator
Bureau of Land Management
Needles Field Office
Mr. Richard Martin
Superintendent
National Park Service
Death Valley National Park
Bureau of Land Management
Mr. John "Jack" S. Mills
Environmental Coordinator
Bureau of Land Management
California State Office
Mr. Barney Lewis
U.S. Geological Survey
Mr. Steve Addington
Field Office Manager
Bureau of Land Management
Bishop Field Office
Mr. Brian Amme
NEPA
Bureau of Land Management
Nevada State Office
Mr. Mike F. Dwyer
Field Office Manager
Bureau of Land Management
Las Vegas Field Office
Mr. Timothy M. Read
Field Manager
Bureau of Land Management
Barstow Field Office
Mr. Ahmed Mohsen
Environmental Coordinator
Bureau of Land Management
Ridgecrest Field Office
Terry Reed
Field Office Manager
Bureau of Land Management
Winnemucca Field Office
Mr. John O. Singlaub
Field Office Manager
Bureau of Land Management
Carson City Field Office
D-6
Distribution List
Mr. Gerald M. Smith
Field Office Manager
Bureau of Land Management
Battle Mountain Field Office
Field Station Manager
Bureau of Land Management
Caliente Field Station
Ms. Cathy Carlson
U.S. Department of Energy
Nevada Operations Office
Ms. Beverly Clark
Manager
U.S. Department of Energy
Idaho Operations Office
Mr. Richard Glass
Manager
U.S. Department of Energy
Albuquerque Operations Office
Mr. James C. Hall
U.S. Department of Energy
Oak Ridge Operations Office
Mr. Keith Klein
U.S. Department of Energy
Richland Operations Office
Mr. Gregory P. Rudy
Manager
U.S. Department of Energy
Savannah River Operations Office
Mr. Robert L. San Martin
Acting Manager
U.S. Department of Energy
Chicago Operations Office
Mr. James M. Tumer
Manager
U.S. Department of Energy
Oakland Operations Office
Department of Energy Advisory Boards
Mr. Jim Bierer
Chair
Femald Citizens Advisory Board
Ms. Merilyn B. Reeves
Chair
Hanford Site Advisory Board
Mr. Chuck Rice
Chair
Idaho National Engineering Laboratory Site-
Specific Advisory Board
Mr. Antontio Delgado, Ph.D., Chair
Los Alamos National Laboratory Advisory
Board
Dr. Ray Johnson
Chair
Nevada Test Site Advisory Board
U.S. Department of Energy
Ms. Sheree Black
Administrative Assistant
Oak Ridge Reservation Environmental
Management Site-Specific Advisory Board
Mr. Tom Marshall
Vice Chair
Rocky Flats Citizens' Advisory Board
Ms. Tonya Covington
Administrator
Sandia Citizens Advisory Board
Ms. Ann Loadholt
Chairperson
Savannah River Citizens Advisory Board
D-7
Distribution List
C. state of Nevada
C.I STATEWIDE OFFICES AND LEGISLATURE
The Honorable Kenny Guinn
Governor of Nevada
The Honorable Lorraine Hunt
Lieutenant Governor of Nevada
The Honorable Frankie Sue Del Papa
Attorney General of Nevada
The Honorable Peter Emaut
Chief of Staff
Office of the Governor
The Honorable William Raggio
Majority Leader
Nevada State Senate
The Honorable Dina Titus
Minority Leader
Nevada State Senate
The Honorable Joseph E. Dini, Jr.
Speaker of the House
Nevada State Assembly
The Honorable Richard Perkins
Majority Floor Leader
Nevada State Assembly
The Honorable Lynn Hettrick
Minority Floor Leader
Nevada State Assembly
The Honorable Bob Price
Chairman
Committee on High Level Radioactive Waste
Nevada State Legislature
The Honorable Mike McGinness
Vice Chairman
Committee on High Level Radioactive Waste
Nevada State Legislature
John Meder
Research Division
Legislative Council Bureau
Nevada State Legislature
C.2 STATE AND LOCAL AGENCIES AND OFFICIALS
Alan Kalt
Comptroller
Churchill County
Dennis Bechtel
Planning Manager
Clark County
Peter Chamberlin
Yucca Mountain Repository Assessment Office
Inyo County
Tammy Manzini
Program Coordinator
Lander County
Tony Cain
Program Director
Nuclear Waste Repository Oversight Program
Esmeralda County
Leonard Fiorenzi
Public Works Director
Eureka County
Eve Culverwell
Administrative Coordinator
Lincoln County
Judy Shankle
Administrator
Office of Nuclear Projects
Mineral County
D-8
Distribution List
Les Bradshaw
Manager
Department of Natural Resources and Federal
Facilities
Nye County
Nick Stellavato
On-Site Representative
Department of Natural Resources and Federal
Facilities
Nye County
Debra Kolkman
Director
Nuclear Waste Project Office
White Pine County
Robert Ferraro
Mayor of Boulder City
Kevin Phillips
Mayor of Caliente
James Gibson
Mayor of Henderson
Oscar Goodman
Mayor of Las Vegas
Larry Gray
Chair
Beatty Town Advisory Board
Gary Hollis
Pahrump Town Board
Thomas Stephens
Director
Nevada Department of Transportation
Michael Tumipseed
State Engineer
Nevada Division of Water Resources
Brian McKay
Chairman
Nevada Commission on Nuclear Projects
Robert Loux
Executive Director
Agency for Nuclear Projects
State of Nevada
Robert Halstead
Transportation Advisor
Agency for Nuclear Projects
State of Nevada
Chuck Home
Mayor of Mesquite
Michael Montandon
Mayor of North Las Vegas
Joe Strolin
Administrator of Planning
Agency for Nuclear Projects
State of Nevada
James Quirk
Amargosa Valley Town Board
D-9
Distribution List
D. Other States and Territories
The Honorable Don Siegelman
Governor of Alabama
The Honorable Tom Vilsack
Governor of Iowa
The Honorable Tony Knowles
Governor of Alaska
The Honorable Bill Graves
Governor of Kansas
The Honorable Tauese P.F. Sunia
Governor of American Samoa
The Honorable Jane Dee Hull
Governor of Arizona
The Honorable Paul E. Patton
Governor of Kentucky
The Honorable Mike Foster
Governor of Louisiana
The Honorable Mike Huckabee
Governor of Arkansas
The Honorable Angus S. King, Jr.
Governor of Maine
The Honorable Gray Davis
Governor of California
The Honorable Bill Owens
Governor of Colorado
The Honorable Parris N. Glendening
Governor of Maryland
The Honorable Argeo Paul Cellucci
Governor of Massachusetts
The Honorable John G. Rowland
Governor of Connecticut
The Honorable Thomas R. Carper
Governor of Delaware
The Honorable John Engler
Governor of Michigan
The Honorable Jesse Ventura
Governor of Minnesota
The Honorable Jeb Bush
Governor of Florida
The Honorable Roy Barnes
Governor of Georgia
The Honorable Carl T.C. Gutierrez
Governor of Guam
The Honorable Kirk Fordice
Governor of Mississippi
The Honorable Mel Camahan
Governor of Missouri
The Honorable Marc Racicot
Governor of Montana
The Honorable Benjamin J. Cayetano
Governor of Hawaii
The Honorable Mike Johanns
Governor of Nebraska
The Honorable Dirk Kempthome
Governor of Idaho
The Honorable George Ryan
Governor of Illinois
The Honorable Frank O'Bannon
Governor of Indiana
The Honorable Jeanne C. Shaheen
Governor of New Hampshire
The Honorable Christine Todd Whitman
Governor of New Jersey
The Honorable Gary E. Johnson
Governor of New Mexico
D-10
Distribution List
The Honorable George E. Pataki
Governor of New York
The Honorable William J. Janklow
Governor of South Dakota
The Honorable James B. Hunt, Jr.
Governor of North Carolina
The Honorable Don Sundquist
Governor of Tennessee
The Honorable Edward T. Schafer
Governor of North Dakota
The Honorable George W. Bush
Governor of Texas
The Honorable Pedro Tenoroio
Governor of Northern Mariana Islands
The Honorable Michael O. Leavitt
Governor of Utah
The Honorable Robert Taft
Governor of Ohio
The Honorable Howard Dean, M.D.
Governor of Vermont
I
The Honorable Frank Keating
Governor of Oklahoma
The Honorable John A. Kitzhaber
Governor of Oregon
The Honorable Tom J. Ridge
Governor of Pennsylvania
The Honorable Pedro J. Rossello Gonzalez
Governor of Puerto Rico
The Honorable Lincoln Almond
Governor of Rhode Island
The Honorable James S. Gilmore, III.
Governor of Virginia
The Honorable Charles W. TumbuU
Governor of Virgin Islands
The Honorable Gary Locke
Governor of Washington
The Honorable Cecil Underwood
Governor of West Virginia
The Honorable Tommy G. Thompson
Governor of Wisconsin
The Honorable Jim Hodges
Governor of South Carolina
The Honorable Jim Geringer
Governor of Wyoming
D-11
Distribution List
E. Native American Groups
Mr. Curtis Anderson
Tribal Chairperson
Las Vegas Paiute Colony
Ms. Geneal Anderson
Tribal Chairperson
Paiute Indian Tribes of Utah
Mr. Richard Arnold
Tribal Chairperson
Pahrump Paiute Tribe
Ms. Rose Marie Bahe
Tribal Chairperson
Benton Paiute Indian Tribe
Mr. Richard Boland
Chief Spokesperson
Timbisha Shoshone -
Restoration Project
Death Valley Land
Ms. Carmen Bradley
Tribal Chairperson
Kaibab Band of Southern Paiutes
Mr. Kevin Brady, Sr.
Tribal Chairperson
Yomba Shoshone Tribe
Ms. Gjrjle Dunlap
Tribal Chairperson
Chemehuevi Indian Tribe
Mr. Daniel Eddy, Jr.
Tribal Chairperson
Colorado River Indian Tribes
Ms. Pauline Esteves
Tribal Chairperson
Timbisha Shoshone Tribe
Mr. Mervin Hess
Tribal Chairperson
Bishop Paiute Indian Tribe
Mr. Jesse Leeds
Organization Chairperson
Las Vegas Indian Center
Mr. Frederick I. Marr
Counsel to the Timbisha Shoshone Tribe
Mr. Tim Thompson
Tribal Chairperson
Duckwater Shoshone Tribe
Ms. Roseanne Moose
Tribal Chairperson
Big Pine Paiute Tribe of the Owens Valley
Ms. Wendy Stine
Tribal Chairperson
Fort Independence Indian Tribe
Mr. Ron Apadaca
Tribal Chairperson
Ely Shoshone Tribe
Mr. Eugene Tom
Tribal Chairperson
Moapa Paiute Indian Tribe
Ms. Sandra J. Yonge
Interim Tribal Chairperson
Lone Pine Paiute-Shoshone Tribe
Mr. Darryl Bahe
Tribal Representative
Benton Paiute Indian Tribe
Ms. Lila Carter " '
Tribal Representative '
Las Vegas Paiute Colony
Ms. Eldene Cervantes
Tribal Representative
Paiute Indian Tribes of Utah
Mr. Jerry Charles
Tribal Representative
Ely Shoshone Tribe
Mr. David L. Chavez
Tribal Representative
Chemehuevi Indian Tribe
D-12
Distribution List
Mr. Lee Chavez
Tribal Representative
Bishop Paiute Indian Tribe
Mr. Donald J. Cloquet
Organization Representative
Las Vegas hidian Center
Ms. Betty L. Cornelius
Tribal Representative
Colorado River Indian Tribes
Ms. Charlotte Domingo
Tribal Representative
Paiute Indian Tribes of Utah
Mr. Maurice Frank-Churchill
Tribal Representative
Yomba Shoshone Tribe
Ms. Grace Goad
Tribal Representative
Timbisha Shoshone Tribe
Ms. Lalovi Miller
Tribal Representative
Moapa Paiute Indian Tribe
Mr. Vernon J. Miller
Tribal Representative
Fort Independence Indian Tribe
Ms. Bertha Moose
Tribal Representative
Big Pine Paiute Tribe of the Owens Valley
Ms. Gaylene Moose
Tribal Representative
Big Pine Paiute Tribe of the Owens Valley
Ms. Priscilla Naylor
Tribal Representative
Fort Independence Indian Tribe
Raymond Gonzales, Sr.
Chairman
Elko Band Council
Ms. Vivienne-Caron Jake
Tribal Representative
Kaibab Band of Southern Paiutes
Gilford Jim
Chairman
Battle Mountain Band Council
Ms. Rachel Joseph
Tribal Representative
Lone Pine Paiute-Shoshone Tribe
Ms. Michelle Saulque
Tribal Representative
Benton Paiute Indian Tribe
Ms. Lawanda Laffoon
Tribal Representative
Colorado River Indian Tribes
Ms. Gevene E. Savala
Tribal Representative
Kaibab Band of Southern Paiutes
Mr. Charles W. Lynch
Ms. Cynthia V. Lynch
Tribal Representative
Pahrump Paiute Tribe
Mr. Rudie Macias
Tribal Representative
Chemehuevi Indian Tribe
Stacy Stahl
Tribal Chairperson
Yerington Tribal Council
Darryl Crawford
Executive Director
Nevada Indian Environmental Coalition
Inter-Tribal Council of Nevada
Mr. Calvin Meyers
Tribal Representative
Moapa Paiute Indian Tribe
Steve Poole
Nevada Indian Environmental Coalition
Inter-Tribal Council of Nevada
D-13
Distribution List
Julie A. Gallardo Brian Wallace
Vice-Chairperson Chairman
Wells Band Council Washoe Tribal Council
Ernestine Coble Alvin James
Fort McDermitt Paiute-Shoshone Tribe Pyramid Lake Paiute Tribe
MRS Project Office
William Rosse Sr.
Helen Snapp Western Shoshone Nation
Fort McDermitt Paiute Shoshone Tribe
D-14
Distribution List
F. Environmental and Public Interest Groups
F.1 NATIONAL
Ms. Maureen Eldredge
Program Director
Alliance for Nuclear Accountability
Washington, DC
Ms. Susan Gordon
Director
Alliance for Nuclear Accountability
Seattle, WA
Ms. Karen Walls
Legislative Research Assistant
American Public Power Association
Washington, DC
Ms. Beth Gal legos
Citizens Against Contamination
Commerce City, CO
Mr. Toney Johnson
Citizens Against Nuclear Trash
Homer, LA
Ms. Janet Greenwald
Citizens for Alternatives to Radioactive
Dumping (CARD)
Albuquerque, NM
Dr. Mildred McClain
Citizens for Environmental Justice, Inc.
Savannah, GA
Mr. Jay Coghlan
Program Director
Concerned Citizens for Nuclear Safety
Santa Fe, NM
Ms. Lesley Jackson
Director
Council of Energy Resource Tribes
Denver, CO
Mr. Seth Kirshenberg
Executive Director
Energy Communities Alliance
Washington, DC
Mr. Fred Krupp
Executive Director
National Headquarters
Environmental Defense Fund, Inc.
New York, NY
Mr. Daniel Kirshner
Senior Analyst
West Coast Office
Environmental Defense Fund, Inc.
Oakland, CA
Mr. Chuck Broscious
Executive Director
Environmental Defense Institute
Troy, ID
Dr. Brent Blackwelder
President
Friends of the Earth
Washington, DC
Mr. Tom Carpenter
Government Accountability Project
Seattle, WA
Mr. Tom Clements
Nuclear Control Institute
Washington, DC
Mr. Tom Goldtooth
National Coordinator
Indigenous Environmental Network
Bemidji, MN
Mr. Arjun Makhijani, Ph.D.
President
Institute for Energy and Environmental Research
(lEER)
Takoma Park, MD
Ms. Bonnie Burgess
League of Women Voters
Washington, DC
D-15
Distribution List
Mr. Daniel Taylor
Executive Director
California State Office
National Audubon Society
Sacramento, CA
Ms. Jo Ann Chase
Executive Director
National Congress of American Indians
Washington, DC
Ms. Libby Fayad
Counsel
National Parks and Conservation Association
Washington, DC
Mr. Jerry Pardilla
National Tribal Environmental Council
Albuquerque, NM
Mr. Mark Van Putten
President and Chief Executive Officer
National Wildlife Federation
Vienna, VA
Ms. Gail Small
Native Action
Lame Deer, MT
Ms. Jill Kennay
Resources Manager
Natural Land Institute
Rockford, IL
Dr. Thomas V. Cochran
Director, Nuclear Programs
Natural Resources Defense Council, Inc.
Washington, DC
Ms. Kafi Watlington-MacLeod
Natural Resources Defense Council
Los Angeles, CA
Ms. Maggie Coon
Director of Government and Community
Relations
The Nature Conservancy
Arlington, VA
Mr. John Humke
Director of Agency Relations
Western Regional Office
The Nature Conservancy
Boulder, CO
Mr. Steven DoUey
Research Director
Nuclear Control Institute '
Washington, DC
Mr. Ralph Hutchison
Coordinator
Oak Ridge Environmental Peace Alliance
Oak Ridge, TN
Mr. Robert Tiller
Director of Security Programs
Physicians for Social Responsibility
Washington, DC
Mr. David Gulp
Plutonium Challenge
Washington, DC
Ms. Christine Chandler ,
Responsible Environmental Action League
Los Alamos, NM
Mr. Tom Marshall
Rocky Mountain Peace and Justice Center
Boulder, CO
Mr. Scott Denman
Executive Director
Safe Energy Communication Council
Ms. Vemice Miller
Natural Resources Defense Council
New York, NY
Mr. Jim Bloomquist
Senior Field Representative
Southern CA/NV/HI Office
Sierra Club
Los Angeles, CA
D-16
Distribution List
Ms. Beatrice Brailsford
Program Director
Snake River Alliance
Pocatello, ID
Mr. Richard Moore
Southwest Network for Environmental and
Economic Justice
Albuquerque, NM
Mr. Don Hancock
Southwest Research and Information Center
Albuquerque, NM
Ms. Marylia Kelley
Tri-Valley CAREs
Livermore, CA
Mr. Alden Meyer
Director, Government Relations
Union of Concerned Scientists
Washington, DC
Ms. Rebecca Stanfield
Staff Attorney
U.S. Public Interest Research Group
Washington, DC
F.2 STATE AND LOCAL
Ms. Kaitlin Backlund
Executive Director
Citizen Alert
Reno, NV
Ms. M. Lee Dazey
Northern Nevada Director
Citizen Alert
Reno, NV
Ms. Allie Smith
Citizen Alert
Las Vegas, NV
Mr. Hal Rogers
Co-Chair, Northern Nevada
The Study Committee
Dayton, NV
Ms. Jackie Cabasso
Executive Director
Western States Legal Foundation
Oakland, CA
Ms. Diane Jackson
Administrative Assistant
Ecology and Economics Research Department
The Wilderness Society
Washington, DC
Mr. Theodore Webb
Peace Action
Sacramento, CA
Ms. Shelby Jones
Solar Presents
Sacramento, CA
Ms. Bemice Kring
Grandmothers for Peace
Sacramento, CA
Ms. Shiela Baker
Nuclear Waste Information Committee
San Louis Obispo, CA
Mr. Bill Vasconi
Co-Chair, Southern Nevada
The Study Committee
Las Vegas, NV
Ms. Judy Treichel
Nevada Nuclear Waste Task Force
Las Vegas, NV
Grace Potorti
Rural Alliance For Military Accountability
Reno, NV
D-17
Distribution List
G. Other Groups and Individuals
Ms. Janice Owens
Director
Nuclear Waste Programs
National Association of Regulatory Utility
Commissioners
Mr. Joe Colvin
President and Chief Executive Officer
Nuclear Energy Institute
Ms. Angie Howard
Senior Vice President
Industry Communications
Nuclear Energy Institute
Mr. Marvin Fertel
Senior Vice President
Nuclear Infrastructure Support and International
Programs
Nuclear Energy Institute
Mr. Ralph Beedle
Senior Vice President
Nuclear Generation
Nuclear Energy Institute
Mr. John Kane
Vice President
Government Affairs
Nuclear Energy Institute
Mr. Theodore Garrish
Vice President
Legislative Affairs
Nuclear Energy Institute
Mr. Steven Kraft
Director
Nuclear Fuel Management
Nuclear Energy Institute
Mr. Scott Peterson
Senior Director
External Communications
Nuclear Energy Institute
Mr. Steven Kerekes
Section Manager
Media Relations
Nuclear Energy Institute
Mr. Steven Unglesbee
Manager
Media Relations
Nuclear Energy Institute
Dr. Klaus Stezenbach
Director
Harry Reid Center for Environmental Studies
University of Nevada, Las Vegas
Dr. Don Baepler
Director of Museum
Harry Reid Center for Environmental Studies
University of Nevada, Las Vegas
Dr. Stephen Wells
President
Desert Research Institute
Mr. Rex Massey
RMA Research
Mr. Chris Binzer ''" ' •
Meridian Center '
Ms. Ginger Swartz
Swartz and Associate
Mr. Ralph Andersen
Project Manager
Plant Support
Nuclear Energy Institute
D-18
Distribution List
H. Reading Rooms and Libraries
Peter Chamberlin
Inyo County Yucca Mountain Repository
Assessment Office
Independence, CA
Annette Ross
U. S. Department of Energy
Public Reading Room
Oakland, CA
Sarah Manion
National Renewable Energy Laboratory
Public Reading Room
Golden, CO
Ann Smith
Rocky Flats Public Reading Room
Westminster, CO
Nancy Mays/Laura Nicholas
Atlanta Support Office
U.S. Department of Energy
Public Reading Room
Atlanta, GA
Joel W. Seymour/Carol M. Franklin
Southeastern Power Administration
U.S. Department of Energy
Reading Room
Elberton, GA
Adrien Taylor
Boise State University Library
Government Documents
Boise, ID
Brent Jacobson/Gail Willmore
Idaho Operations Office
Department of Energy
Public Reading Room
Idaho Falls, ID
John Shuler
Chicago Operations Office
Document Department
University of Illinois at Chicago
Chicago, IL
Deanna Harvey
Strategic Petroleum Reserve Project
Management Office
U.S. Department of Energy
SPRPMO/SEB Reading Room
New Orleans, LA
Alan Kalt
Churchill County
Fallon, NV
Dennis Bechtel
Clark County
Las Vegas, NV
Aimee Quinn
Government Publications
Dickenson Library
University of Nevada, Las Vegas
Las Vegas, NV
Tony Cain
Esmeralda County
Repository Oversight Program
Goldfield, NV
Leonard Fiorenzi
Eureka County
Courthouse Annex
Eureka, NV
Tammy Manzini
Lander County
Austin, NV
Eve Culverwell
Lincoln County
Caliente, NV
Jackie Wells
Mineral County
Hawthorne, NV
Heather Elliot
Nevada State Clearinghouse
Department of Administration
Carson City, NV
D-19
Distribution List
Les Bradshaw
Nye County
c/o Department of Natural Resources and
Federal Facilities
Pahrump, NV
University of Nevada, Reno
The University of Nevada Libraries
Business and Government Information Center
Reno, NV
Debra Kolkman
White Pine County
Ely, NV
Beatty Yucca Mountain Science Center
Beatty, NV
Las Vegas Yucca Mountain Science Center
Las Vegas, NV
Pahrump Yucca Mountain Science Center
Pahrump, NV
Shawna Schwartz
Albuquerque Operations Office
US Department of Energy
Contract Reading Room
Kirtland Air Force Base
Albuquerque, NM
Gary Stegner
Femald Area Office
U.S. Department of Energy
Public Information Room
Cincinnati, OH
Josh Stroman
Bartlesville Project Office/National Institute for
Petroleum and Energy Research
BPO/NIPER Library
U.S. Department of Energy
Bartlesville, OK
Pam Bland
Southwestern Power Administration
U.S. Department of Energy
Tulsa, OK
Jean Pennington
Bonneville Power Administration
U.S. Department of Energy
Portland, OR
Ann C. Dunlap
Pittsburgh Energy Technology Center
U.S. Department of Energy
Pittsburgh, PA
David Darugh
Savannah River Operations Office
Gregg-Graniteville Library
University of South Carolina-Aiken
Aiken, SC
Lester Duncan
University of South Carolina
Thomas Cooper Library
Documents/Microforms Department
Columbia, SC
Amy Rothrock/Teresa Brown
Oak Ridge Operations Office
U.S. Department of Energy
Public Reading Room
American Museum of Science and Energy
Oak Ridge, TN
Stephen Short
Southern Methodist University
Central Union Libraries Fondren Library
Government Information
Dallas, TX
Walter Jones
University of Utah
Marriott Library
Special Collections
Salt Lake City, UT
Carolyn Lawson
Headquarters Office
U.S. Department of Energy
Washington, DC
Tommy Smith
OCRWM National Information Center
Washington, DC
Terri Traub
Richland Operations Center
U.S. Department of Energy
Public Reading Room
Richland, WA
D-20
Appendix E
Environmental Considerations for
Alternative Design Concepts and
Design Features for the Proposed
Monitored Geologic Repository
at Yucca Mountain, Nevada
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
TABLE OF CONTENTS
Section Page
E.l Introduction E-1
E.1.1 Objective E-1
E.1.2 Background E-1
E.1.3 Scope E-3
E.2 Design Features and Alternatives E-3
E.2.1 Barriers to Limit Release and Transport of Radionuclides E-3
E.2. 1.1 Ceramic Coatings E-3
E.2.1. 1.1 Potential Benefits E-3
E.2. 1.1.2 Potential Environmental Considerations E-4
E.2.1.2 Drip Shields E-4
E.2.1.2.1 Potential Benefits E^
E.2. 1.2.2 Potential Environmental Considerations E-4
E.2.1.3 Backfill E^
E.2.1.3.1 Potential Benefits E-4
E.2. 1.3.2 Potential Environmental Considerations E-5
E.2. 1.4 Waste Package Corrosion-Resistant Materials E-5
E.2.1.4.1 Potential Benefits E-5
E.2. 1.4.2 Potential Environmental Considerations E-5
E.2.1.5 Richards Barrier E-5
E.2.1.5.1 Potential Benefits E-6
E.2. 1.5.2 Potential Environmental Considerations E-6
E.2. 1.6 Diffusive Barrier Under the Waste Package E-6
E.2.1.6.1 Potential Benefits E-6
E.2. 1.6.2 Potential Environmental Considerations E-6
E.2. 1.7 Getter Under Waste Packages E-7
E.2.1.7.I Potential Benefits E-7
E.2. 1.7.2 Potential Environmental Considerations E-7
E.2. 1.8 Canistered Assemblies E-7
E.2.1.8.1 Potential Benefits E-7
E.2. 1.8.2 Potential Environmental Considerations E-7
E.2.1.9 Additives and Fillers E-8
E.2.1.9.1 Potential Benefits E-8
E.2. 1.9.2 Potential Environmental Considerations E-8
E.2.1. 10 Ground Support Options E-8
E.2.1. 10.1 Potential Benefits E-9
E.2. 1.10.2 Potential Environmental Considerations E-9
E.2.2 Repository Designs to Control Heat and Moisture E-9
E.2.2.1 Design Alternative 1, Tailored Waste Package Spatial Distribution E-9
E.2.2.1.1 Potential Benefits E-9
E.2.2. 1.2 Potential Environmental Considerations E-9
E.2.2.2 Design Alternative 2, Low Thermal Load E-10
E.2.2.2.1 Potential Benefits E-10
E.2.2.2.2 Potential Environmental Considerations E-10
E.2.2.3 Design Alternative 3, Continuous Postclosure Ventilation E-10
E.2.2.3.1 Potential Benefits E-11
E.2.2.3.2 Potential Environmental Considerations E-U
E.2.2.4 Design Alternative 6, Viability Assessment Reference Design E-1 1
E.2.2.5 Design Alternative 7, Viability Assessment Reference Design with Options E-1 1
E-iii
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
Section page
E.2.2.6 Aging and Blending of Waste E-11
E.2.2.6.1 Potential Benefits E-11
E.2.2.6.2 Potential Environmental Considerations E-12
E.2.2.7 Continuous Preclosure Ventilation E-12
E.2.2.7.1 Potential Benefits E-12
E.2.2.7. 2 Potential Environmental Considerations E-12
E.2.2.8 Drift Diameter E-13
E.2.2.8.1 Potential Benefits E-13
E.2.2.8. 2 Potential Environmental Considerations E-13
E.2.2.9 Drift Spacing and Waste Package Spacing E-14
E.2.2.9.1 Potential Benefits E-14
E.2.2.9.2 Potential Environmental Considerations E-14
E.2.2.10 Near-Field Rock Treatment E-15
E.2.2.10.1 Potential Benefits E-15
E.2.2.10.2 Potential Environmental Considerations E-15
E.2.2.I1 Surface Modification - Alluvium Addition E-15
E.2.2.I1.1 Potential Benefits E-15
E.2.2.1I.2 Potential Environmental Considerations E-15
E.2.2.12 Surface Modification - Drainage E-16
E.2.2.12.1 Potential Benefits E-16
E.2.2.12.2 Potential Environmental Considerations E-16
E.2.2.13 Higher Thermal Loading E-17
E.2.3 Repository Designs to Support Operational and/or Cost Considerations E-17
E.2.3.1 Design Alternative 4, Enhanced Access E-17
E.2.3.1.1 Potential Benefits E-17
E.2.3. 1.2 Potential Environmental Considerations E-17
E.2.3. 2 Design Alternative 5, Modified Waste Emplacement Mode E-17
E.2.3.2.1 Potential Benefits E-18
E.2.3.2.2 Potential Environmental Considerations E-18
E.2.3. 3 Design Alternative 8, Modular Design (Phased Construction) E-18
E.2.3.3.1 Potential Benefits E-18
E.2.3. 3.2 Potential Environmental Considerations E-18
E.2.3.4 Rod Consolidation E-18
E.2.3.4.1 Potential Benefits E-19
E.2.3.4.2 Potential Environmental Considerations E-19
E.2.3.5 Timing of Repository Closure E-19
E.2.3.5.1 Potential Benefits E-19
E.2.3. 5. 2 Potential Environmental Considerations E-20
E.2.3. 6 Maintenance of Underground Features and Ground Support E-20
E.2.3.6.1 Potential Benefits E-20
E.2.3.6.2 Potential Environmental Considerations E-20
E.2.3.7 Waste Package Self-Shielding E-20
E.2.3.7.1 Potential Benefits E-21
E.2.3. 7. 2 Potential Environmental Considerations E-21
E.2.3. 8 Repository Horizon Elevation E-21
E.2.3. 8.1 Potential Benefits E-21
E.2.3. 8.2 Potential Environmental Considerations E-21
E.3 Enhanced Design Alternatives E-22
E.3.1 Enhanced Design Alternative I E-22
E.3. 2 Enhanced Design Alternative II E-22
E.3. 3 Enhanced Design Alternative III E-23
E-iv
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
Section Page
E.3.4 Enhanced Design Alternative FV E-24
E.3.5 Enhanced Design Alternative V E-24
Reference E-25
E-v
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
APPENDIX E. ENVIRONMENTAL CONSIDERATIONS FOR
ALTERNATIVE DESIGN CONCEPTS AND DESIGN FEATURES FOR
THE PROPOSED MONITORED GEOLOGIC REPOSITORY
AT YUCCA MOUNTAIN, NEVADA
E.1 Introduction
E.1.1 OBJECTIVE
This appendix discusses design features and alternatives for a repository at Yucca Mountain in Nevada
that were under consideration by the U.S. Department of Energy (DOE) in the winter of 1998 and early
1999. It represents a forward look at how the repository design might evolve to incorporate these and/or
other features into a reference design that could be submitted in a repository license application. This
appendix also addresses how this design evolution might affect parameters important to the assessment of
environmental impacts. The design features and alternatives analyzed as part of the Yucca Mountain Site
Characterization Project were conceptual in nature (that is, not developed or analyzed in detail). This
appendix presents a qualitative description of the design features and alternatives and a brief assessment
of factors associated with each that could cause changes to the environmental impacts analyzed in this
environmental impact statement (EIS). This assessment generally indicates that the EIS reasonably
represents the foreseeable evolutions in repository design related to environmental impact considerations
and bounds potential impacts. Possible design evolutions that occur after DOE issues this Draft EIS will
be factored into the Final EIS, as appropriate, and any such refined design concepts will be carried
forward to license application if Yucca Mountain is determined to be a suitable site for a repository.
E.1 .2 BACKGROUND
DOE has completed the Viability Assessment of a Repository at Yucca Mountain (DOE 1998, all). The
Viability Assessment included a preliminary design concept (referred to as the Viability Assessment
reference design throughout this appendix), which presented preliminary design concepts for the
repository surface facilities, underground facilities, and waste packages. The Viability Assessment
reference design is the same as the high thermal load implementing alternative in the EIS.
Technical work associated with the Viability Assessment and the Viability Assessment reference design
was not intended to support the selection of a repository design concept or specific alternative for
licensing. Rather, the Viability Assessment identified areas requiring further study to determine site
suitability to support a Site Recommendation and a License Application for a repository at Yucca
Mountain. One area of further study and evaluation identified in the Viability Assessment was the
assessment of alternative repository design features and concepts. The License Application Design
Selection Process was established to study a broad range of alternative design concepts and design
features to support the selection of the design to be incorporated into a license application.
The License Application Design Selection Process used a multistep approach for evaluating a selected set
of features and alternatives against several criteria, including postclosure waste isolation performance,
preclosure performance, assurance of safety, engineering acceptance, operations and maintenance,
schedule, cost, and environmental considerations. In the first step, features and alternatives are evaluated
against these criteria. Following this initial evaluation, enhanced design alternatives (which provide a
unique approach to repository design and rely on the attributes of selected design features) were
developed. In the development of enhanced design alternatives, there were no limitations placed on the
development team to restrict consideration of features and alternatives to those on the initially selected
list. From the inception of the License Application Design Selection Process, additional or evolved
E-1
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
alternatives were expected to result. The process called for ranking of the enhanced design alternatives
against a selected set of criteria using decision analysis methods. At the time of development of this
appendix, enhanced design alternatives that were not part of the Viability Assessment had been
developed, but documentation of that development and ranking had not been completed. Therefore, the
information presented in this appendix is preliminary and based on both observations of the process and
informal discussions with License Application Design Selection Process participants. This appendix will
be revised as necessary to incorporate the final results of the License Application Design Selection
Process. For the purposes of the License Application Design Selection Process, the following terms were
defined:
• Design Feature. A design feature is a particular element or attribute of the repository design for
which postclosure performance could be evaluated independently of a specific repository design
alternative (fully developed design concept) or other design features. An individual design feature
could encompass separate discrete concepts or a continuous range of parametric values. Design
features can be added singularly or in combination to a design alternative. A design feature could
theoretically be applied to any design alternative, although logical compatibility and expected
postclosure waste isolation performance enhancement might be evident only when applied to
particular design alternatives. Section E.2 of this appendix discusses the design features that were
considered in the License Application Design Selection Process.
• Design Alternative. Each design altemative represents a fundamentally different conceptual design
for the repository, which could potentially stand alone as the license application repository design
concept. A design altemative can define major sections or the entire repository design. Design
alternatives are distinguished from design features by their complexity and their inclusion of several
features. Furthermore, a number of attributes are required to distinguish one design altemative from
another. While not mutually exclusive, design alternatives represent diverse and independent
methods of accomplishing the repository mission. Section E.2 discusses the design altematives that
were considered in the License Application Design Selection Process.
• Enhanced Design Alternative. Enhanced design altematives are combinations (and/or variations)
of one or more design altemative and design feature. While an enhanced design altemative could be
made up of any conceivable combination of design altematives and design features, enhanced design
altematives selected for further evaluation are those combinations that include mutually compatible
attributes and expected postclosure waste isolation performance characteristics that exceed those of
the basic design altematives. In other words, the enhanced design altematives are all improvements
to the design altematives in the first phase of the License Application Design Selection Process,
including the Viability Assessment reference design. Other considerations in developing the
enhanced design altematives include the compatibility of the features and altematives; the
developmental, operational, and maintenance simplicity of the resulting combination; and the ability
of the set of enhanced design altematives to address the entire set of design features and altematives
under consideration.
Recommendations for the repository design concept that resulted fi-om the License Application Design
Selection Process will be part of a technical report scheduled for completion after this appendix was
prepared. The design concept to be carried forward is expected to be one of the five enhanced design
altematives currently identified or minor variations of one of those enhanced design altematives.
Section E.3 of this appendix discusses the enhanced design altematives that are the subject of
consideration in the License Application Design Selection Process.
E-2
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
E.I. 3 SCOPE
This appendix discusses the evolution of the EIS repository design concept to the concept that will
ultimately be submitted as part of the license application for the Yucca Mountain repository, should the
site be approved. The discussion is broken down into three basic categories that reflect the potential types
of benefits from the design features and alternatives under consideration. The benefits that could be
derived from each of the features and alternatives are not necessarily limited to the categorization
presented, and some features and alternatives could fit into more than one category. However, the three
categories were chosen to facilitate an understanding of the design evolution process that is presented in
the main body of the EIS. Section E.2 discusses the set of selected design features and alternatives.
The categories, as presented in Sections E.2.1 through E.2. 3, are Barriers to Limit Release and Transport
of Radionuclides; Repository Designs to Control Thermal/Moisture Environment; and Repository
Designs to Support Operational and Cost Considerations. Within each category, the text includes
descriptions of the features and alternatives, explanations of why each feature/alternative was considered,
and discussions of the potential for environmental impacts associated with each feature/alternative.
Section E.3 presents the five enhanced design alternatives that were considered in the first phase of the
License Application Design Selection Process to develop a design concept for the proposed Yucca
Mountain Repository that was an improvement over the Viability Assessment reference design. This
improvement could take many forms, including enhanced licensibility, reduced uncertainty, and ease of
construction and operation. The five enhanced design alternatives represent five complete basic design
concepts that evolved from consideration of the features and alternatives discussed in Section E.2. The
enhanced design alternatives were selected to represent the potential differences in waste isolation
performance among differing repository designs. The participants in the License Application Design
Selection Process determined that a major factor in selecting the final design for the Yucca Mountain
Repository would be the thermal loading of the repository. As such, the five enhanced design alternatives
represent a range of thermal loads from 40 metric tons of heavy metal (MTHM) per acre to 150 MTHM
per acre. Important differences between the enhanced design alternatives and the Viability Assessment
reference design include differences in waste package materials and the addition of a drip shield to each
of the enhanced design alternatives. Each of the enhanced design alternatives was selected to improve on
the Viability Assessment reference design from a waste isolation performance perspective. As was the
case with the basic design features and alternatives discussed in Section E.2, there is the potential for
environmental impacts associated with the enhanced design alternatives.
•' • -fw. ir-rfiUyj--
^<E.2 Design Features and Alternatives
E.2.1 BARRIERS TO LIMIT RELEASE AND TRANSPORT OF RADIONUCLIDES
E.2.1. 1 Ceramic Coatings
A thin coating [1.5 millimeters (0.06 inch) or more] of a ceramic oxide on the outer surface of the waste
package could increase the life of the waste package by slowing the rate at which the waste package will
corrode. Candidate materials for the ceramic coating are magnesium aluminate spinel, aluminum oxide,
titanium oxide, and zirconia-yttria. Spinel is the leading alternative.
E.2.1 .1 .1 Potential Benefits
The ceramic coating could increase waste package life and repository waste isolation performance by
reducing corrosion of the waste package surface and, therefore, delaying the release of radionuclides.
E-3
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
E.2.1.1.2 Potential Environmental Considerations
There are no significant environmental considerations associated with ceramic coatings.
E.2.1.2 Drip Shields
Drip shields would provide a partial barrier by diverting infiltrating water away from waste packages in
an emplacement drift. Drip shields could be metal (for example, Alloy-22, a nickel-chromium-
molybdenum alloy, or titanium-7, a titanium metal alloyed with 0.15 percent palladium) or ceramic-
coated metal. One option is to place drip shields under backfill; another is to place the drip shields over
the backfill. Drip shields could be implemented with or without backfill.
If the drip shield was placed under backfill, it would fit over the entire length of each waste package,
configured to the outer diameter with an unspecified clearance between drip shield and waste package,
and enclosed at each end. Backfill, which would be emplaced during the repository's closure, would be
comprised of a heaped, single-layered material that covers the waste package and drip shield to some
unspecified depth. Another form of backfill, the Richards Barrier, could also be used. Backfill and
Richards Barriers are discussed later in this appendix.
The drip shield, as used in the second option, is formed to the approximate backfill surface profile and
placed atop the backfill (or Richards Barrier). With this option, the drip shield is placed in conjunction
with the placement of backfill at the closure of the repository.
E.2.1.2.1 Potential Benefits
Drip shields are intended to enhance long-term repository performance by reducing waste package
corrosion and extending waste package life.
E.2.1 .2.2 Potential Environmental Considerations
Additional labor hours would be required for the generation and placement of backfill material, and
industrial accidents could increase proportionately. Although drip shields would be emplaced remotely,
there could be some incidental radiological doses to workers.
Drip shields of titanium-7, Alloy-22, or other corrosion-resistant material would increase the demand for
such materials. Costs for repository closure would increase due to the cost of procuring and installing the
drip shields.
E.2.1 .3 Backfill
At repository closure, loose, dry, granular material such as sand or gravel would be placed over the waste
packages in a continuous, heaped pile. Other materials for backfill, such as crushed rock and depleted
uranium, may be evaluated in the future.
E.2.1 .3.1 Potential Benefits
Backfill would provide protection of waste packages and drip shields (if placed over the drip shields)
from rockfall. It could protect against corrosion of the waste packages by (1) potentially capturing the
corrosive salts of various soluble chemicals that might enter with water intrusion, (2) retarding advective
jflow, and/or (3) increasing the temperature of the emplacement drift to decrease relative humidity.
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Environmental Considerations for Alternative Design Concepts and
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E.2.1 .3.2 Potential Environmental Considerations
Additional workers would be needed, and there would be a potential increase for industrial accidents
because of the additional operations. Although backfill would be placed remotely, there could be some
incidental radiological doses to workers.
E.2.1 .4 Waste Package Corrosion-Resistant Materials
The Viability Assessment reference design for the waste package uses two concentric barrier layers: an
outer 100-millimeter (3.9-inch)-thick A516 carbon steel structural corrosion-allowance material, and an
inner 20-millimeter (0.8-inch)-thick nickel-based alloy-22 corrosion-resistant material. These two
barriers would be expected to provide substantially complete containment of the waste for the lifetime
goals established in the Viability Assessment; however, a waste package with the capability to provide
substantially complete containment for a significantly extended lifetime would be more desirable.
A variation of the waste package design would replace the corrosion-allowance barrier with a second
corrosion-resistant barrier. This design would provide in-depth defense if the second corrosion-resistant
barrier was independent of the first (for example, made of a different metal or ceramic). A number of
configurations of waste package containers with two corrosion-resistant materials were analyzed,
including designs with an inner layer of titanium and outer layer of nickel-based Alloy-22, with a
combined thickness of about 55 millimeters (2.2 inches).
E.2.1 .4.1 Potential Benefits
Longer waste package lifetimes would lead to improved long-term waste isolation performance of the
repository.
E.2.1 .4.2 Potential Environntental Considerations
The addition of a second independent corrosion-resistant layer would prolong waste package lifetimes,
resulting in delay and minimization of potential groundwater contamination.
Radiological dose to workers would increase without compensating changes in operating procedures,
because the total thickness of the waste package container could be less than the Viability Assessment
reference design. Appropriate shielding might have to be provided for the workers engaged in waste
package handling and emplacement operations. However, there would be a potential increased
occupational dose to the workers because the calculated dose rates at the waste package surface would be
higher.
E.2.1 .5 Richards Barrier
A Richards Barrier would be formed by placing two layers of backfill over the emplaced waste packages
at closure. The barrier would consist of a coarse-grained, sand-sized material underlying a fine-grained,
sand-sized material. Both materials would be placed as a continuous, heaped pile extending along the
alignment of the waste packages. A variety of materials could be used for both layers, including depleted
uranium as a coarse-grained material.
The Richards Barrier would be designed to divert water that might enter the emplacement drifts away
from the waste packages by transferring the vertical migration of water seepage laterally along the
interface between the two layers. The particle size distribution, shape, and porosity of material in the two
k
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Environmental Considerations for Alternative Design Concepts and
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layers would provide a permeability difference and would cause the upper layer to channel water seepage
along the boundary of the lower layer.
E.2.1 .5.1 Potential Benefits
The Richards Barrier would delay the transport of water to the waste packages, thereby delaying waste
package corrosion and improving long-term repository performance.
E.2.1 .5.2 Potential Environmental Considerations
Dust and equipment emissions could be a concern during the placement phase of the Richards Barrier.
If the chosen coarse material was depleted uranium, there would be an increase in radon emissions.
Uranium might also lead to an increase in the contamination of groundwater because the uranium in the
Richards Barrier would not be contained or restricted by other engineered barriers. Radiation exposure
would also have to be considered in design and operations of depleted uranium handling.
Additional workers would be needed during closure to implement this design feature, and there would be
an increased potential for industrial accidents. Although personnel would not be in the drifts, there might
be some incidental radiation dose to workers outside the drifts; therefore, additional shielding might be
required for personnel.
E.2.1 .6 Diffusive Barrier Under the Waste Package
A diffusive barrier would consist of loose, dry, granular material placed in the space between each waste
package and the bottom of the emplacement drift to form a restrictive barrier to seepage. Below a critical
seepage flux, water would disperse throughout the porous medium of the diffusive barrier, providing both
lateral vertical dispersion and thereby slowing the fluid movement to the natural environment.
Radionuclides, which might be released from breached waste packages, could become solubilized or
suspended within the seepage flow and be retarded by the porous material forming the barrier.
The diffusive barrier could be anything from common sand to gravel-size material without any special
qualifications to mineralogy, grain size distribution, shape, or density. Depleted uranium could also be
used. The diffusive barrier would be installed prior to waste emplacement.
E.2.1 .6.1 Potential Benefits
Improved waste isolation performance could be achieved by slowing radionuclide movement to the
natural environment.
E.2.1 .6.2 Potential Environmental Considerations
If the diffusive barrier material were depleted uranium, there would be increased radon emissions and
increased radiological dose to workers. There could be an increase in the contamination of groundwater
because the uranium would not be contained or restricted by other engineered barriers.
Additional workers would be needed to construct the diffusive barrier; therefore, there would be a
proportional increase in the potential for industrial accidents.
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Environmental Considerations for Alternative Design Concepts and
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E.2.1 .7 Getter Under Waste Packages
A getter would be a fine-grained material [either phosphate rock (apatite) or iron oxide (hematite,
geothite, etc.] with an affinity for radionuclides. This material would be placed in the invert recess below
the waste packages prior to waste emplacement.
E.2.1. 7.1 Potential Benefits
A getter material below the waste packages could improve long-term waste isolation through retardation
of radionuclide movement from the repository drifts.
E.2.1 .7.2 Potential Environmental Considerations
Additional workers would be needed to place the getter material in the drifts; therefore, there would be a
proportional increase in the potential for industrial accidents.
E.2.1 .8 Canistered Assemblies
Placing spent fuel assemblies in canisters at the Waste Handling Building before inserting them into
waste packages would provide an additional barrier and further limit mobilization of radionuclides if the
waste package is breached. The canisters would be fabricated from a corrosion-resistant material (for
example, Alloy-22 or a zirconium alloy). There are three general concepts for the placement of fuel
assemblies in canisters:
• Rectangular canisters designed to hold individual fuel assemblies: these canisters could be placed
into a waste package with a basket containing neutron absorber and aluminum thermal shunts, similar
to the current basket designs.
• Rectangular canisters designed to hold a few fuel assemblies: these canisters could have neutron
absorber between assemblies and fit into a basket containing neutron absorber and aluminum thermal
shunts.
• Large circular canister designed to hold multiple fuel assemblies and fit one per waste package: the
canister would have an internal basket with neutron absorber, aluminum thermal shunts, and fuel
tubes, similar to previous canistered fuel waste package designs.
E.2.1 .8.1 Potential Benefits
Placing spent fuel assemblies in canisters before inserting them into waste packages would provide an
additional barrier and limit mobilization of radionuclides in breached waste packages.
E.2.1 .8.2 Potential Environmental Considerations
Use of this feature could cause an increase in the size of the Waste Handling Building and require
additional workers. There would be an increase in operations and a possible increase in the number of
lifts required per fuel assembly. This increase could be as much as one extra lift per assembly (canister),
due to the moving of the canister to the waste package, which would lead to the potential for greater
exposure to radiation for workers.
Implementation of this feature could increase the amount of rejected materials due to faulty welding,
potentially generating more low-level radioactive waste and/or solid waste.
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Environmental Considerations for Alternative Design Concepts and
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E.2.1.9 Additives and Fillers
Additives and fillers are materials that could be placed into waste packages (in addition to those normally
required for the basket material) to fill the basket and waste form void spaces. The additives and fillers
would:
• Sorb radionuclides and retard their release from a breached waste package
• Sorb boron neutron absorber that might be released from corrosion of the borated stainless steel
absorber plates
• Displace moderator from the interior of the waste package to provide additional defense-in-depth for
nuclear criticality control
Potential additives and fillers would be oxides of iron and aluminum. These materials could be placed
within the waste package as a powder or as shot following loading of the waste form, or integrated into
the basket design.
E.2.1 .9.1 Potential Benefits
Additives and fillers could improve long-term repository performance by retardation of release of
radionuclides to the groundwater and could also improve long-term criticality control.
E.2.1 .9.2 Potential Environmental Considerations
Adding additives and fillers would make it more difficult to remove spent nuclear fuel assemblies from
waste packages following retrieval, if necessary. Operations would have to include the additional step of
removing this material before removal of the fuel.
E.2.1 .10 Ground Support Options
Ground support in the repository ensures drift stability before closure. Selection of ground support
options could affect repository waste isolation performance. Considerations of ground support options
include functional requirements for ground support, the use of either concrete or steel-lined systems, and
the feasibility of using an unlined drift ground support system with grouted rock bolts.
A concrete lining has been studied for its structural/mechanical behavior and subjected to the load
conditions expected of emplacement drifts. However, a number of postclosure performance assessment
issues related to the presence of concrete within the emplacement drift environment have been identified.
An all-steel ground support system (for example, steel sets with partial or full steel lagging) has been
considered to be a viable ground support candidate for emplacement drifts. Use of an all-steel lining
system would provide a means of limiting or eliminating the introduction of cementitious materials (that
is, concrete, shotcrete, or grout), including organic compounds into the emplacement drift environment.
The potential for corrosion of steel subjected to the emplacement drift environment is a concern with this
system. Another concern is the interaction of steel ground supports with waste package materials.
For an unlined drift scenario, rockbolts and mesh could be considered as permanently maintainable
ground support. Design and performance advantages associated with the use of rockbolts as permanent
ground support for emplacement drifts include durability and longevity of this system. A postclosure
concern would be the suitability of cementitious grout, which would be used for installing rockbolts.
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Environmental Considerations for Alternative Design Concepts and
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E.2.1 .1 0.1 Potential Benefits
Safety during emplacement and potential retrieval would be enhanced by use of appropriate ground
supports. Long-term repository performance could be improved by reducing or delaying damage to
canisters from rockfall, because damaged areas would be locations for enhanced corrosion even if the
canister was not breached by the rockfall.
E.2.1 .1 0.2 Potential Environmental Considerations
The choice of ground support options does not significantly impact any environmental consideration
except for long-term repository waste isolation performance.
E.2.2 REPOSITORY DESIGNS TO CONTROL HEAT AND MOISTURE
E.2.2.1 Design Alternative 1, Tailored Waste Package Spatial Distribution
Tailored spatial distributions of waste packages within the repository block emplacement drifts could
improve the postclosure waste isolation performance of the repository. The EIS design assumes the
various waste package types would be emplaced on a random basis, modified only to meet the areal mass
loading requirement of 25 to 85 MTHM per acre and the commercial fuel cladding and drift wall thermal
goals of 350°C and 2(X)°C (662°F and 392°F), respectively. There are three different methods of spatial
distribution under review, including:
• Distribution of waste packages as a function of infiltrating water percolation rate within various
regions of the repository block. Higher heat-producing packages would be placed in areas with
higher percolation rates.
• Distribution of commercial spent nuclear fuel waste package types as a function of the distance to the
water table and/or unsaturated zone zeolite content. Waste packages with radionuclides with the
highest tendency to travel would be placed furthest from the water table, and waste packages with
radionuclides with a higher tendency to be sorbed would be placed above areas with the highest
zeolite content.
• Grouping waste package types into categories of hot, medium, and cold waste packages to even out
the temperature differences across the repository.
E.2.2.1 .1 Potential Benefits
Tailoring spatial distribution of the waste packages within the repository block might improve the
performance of waste packages by delaying and reducing contact of water and/or increasing sorption of
released radionuclides by zeolites in the unsaturated zone. This form of distribution has the potential to
improve repository waste isolation performance.
E.2.2.1. 2 Potential Environmental Considerations
Larger surface storage facilities could be needed to allow appropriate selection of waste packages for the
desired spatial distribution. However, if the retrieval pad can be used for this purpose, no additional land
would be needed.
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Environmental Considerations for Alternative Design Concepts and
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E.2.2.2 Design Alternative 2, Low Tliermal Load
The low thermal load design alternative would limit the temperature of the drift wall and host rock. It
would cause less thermal change in the host rock than the Viability Assessment reference design.
Limiting temperature rise would also reduce the uncertainty in predicting several processes, and thermal,
chemical, mechanical, and hydrological effects would be easier to describe because coupling of these
effects would extend over a smaller region than the Viability Assessment reference design. In this
evaluation, a low thermal load refers to 40 MTHM per acre.
• Option 1. The waste package spacing would be the same as the spacing of the drifts, creating a
square area between waste packages. The spacing of waste packages would be farther apart than in
the Viability Assessment reference design. This option is the equivalent of the low thermal load
implementing alternative analyzed in the EIS.
• Option 2. The spacing of the waste packages within the drifts would be 9 meters (30 feet) as in the
Viability Assessment reference design, but drift spacing is increased to about 90 meters (300 feet).
This can be compared to 28 meters (92 feet) for the Viability Assessment reference design.
• Option 3. This option consists of a greater number of smaller waste packages than in Option 1 or 2,
and spacing of waste packages within the drifts is similar to Option 2. Drift spacing and excavated
rock volume are about the same as for Option 1.
E.2.2.2.1 Potential Benefits
The primary benefit would be the reduction in uncertainties associated with higher thermal loads and the
elevated temperature of the host rock. Lower repository temperatures could also potentially reduce waste
package material corrosion rates.
E.2.2.2.2 Potential Environmental Considerations
Options 1 and 3 would result in generation of more excavated rock compared to the Viability Assessment
reference design, and therefore requires a larger area for storage/disposal of excavated rock. Subsurface
costs would increase. Option 2 would result in less volume of excavated rock than Option 1 or 3.
E.2.2.3 Design Alternative 3, Continuous Postclosure Ventilation I't" 3^^'
Under this alternative there would be continuous ventilation of the emplacement drifts during the
postclosure period. Ventilation would occur by natural ventilation pressure induced by the difference in
air density between hot and cool areas. Three primary options were considered:
• Closed loop airways connected underground but sealed to the surface
• Open loop airways where the primary airways stay open and in which the repository drifts are open to
exchange air with the atmosphere; two additional ventilation shafts would be needed
• Open/closed loop ventilation where primary airways would be sealed, but drifts would be located
very close to a system of tunnels open to the atmosphere
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Environmental Considerations for Alternative Design Concepts and
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E.2.2.3.1 Potential Benefits
Postclosure ventilation would increase the removal of moisture from air around the waste packages for a
period of time (estimated to be 1,000 to 2,000 years for the closed loop system), but moisture would
eventually reestablish itself. Reduced moisture could improve performance by retarding waste package
corrosion.
E.2.2.3.2 Potential Environmental Considerations
Excavated rock piles would increase in size in proportion to the increase in drift excavation required.
Additional shafts would result in additional surface disturbed areas (small, relative to the Viability
Assessment reference design). Additional occupational exposure to radon-222 associated with excavation
would occur.
Overall, work force would increase by less than 10 percent, as would associated impacts such as industrial
accidents.
E.2.2.4 Design Alternative 6, Viability Assessment Reference Design
The Viability Assessment reference design is equivalent to the high thermal load alternative evaluated in
the EIS.
E.2.2.5 Design Alternative 7, Viability Assessment Reference Design with Options
The Viability Assessment reference design with options was considered as a design alternative in the
License Application Design Selection Process. The Viability Assessment reference and design is
analyzed in detail in the EIS. Options considered include ceramic coatings, drip shields, and backfill (see
Sections E.2.1.1, E.2.1.2, and E.2.1.3, respectively).
E.2.2.6 Aging and Blending of Waste
Pre-emplacement aging and blending of wastes provides mechanisms for managing the thermal output of
a waste package and the total thermal energy that must be accommodated by the repository.
Aging the waste before emplacement results in less variable (over time) thermal output of the waste
packages and lower waste package temperatures. Aging could be performed at the repository, at the
reactor sites, or at other locations.
Blending would allow a more uniform heat output from the waste packages. Blending would be
accomplished by selecting waste forms for insertion in waste packages based on their heat output to
minimize the variability in the thermal energy of each waste package.
E.2.2.6.1 Potential Benefits
Aging would reduce the temperature increase expected at the surface above the repository because the
total heat load of the repository would be decreased. Lower heat output could also result in a smaller
repository footprint by allowing more dense waste emplacement schemes without violating waste package
or drift wall temperature goals. Both blending and aging reduce the variability of the temperature
distribution in the repository, and drifts might be spaced more closely. Lower and equalized temperatures
could improve structural stability of the drifts. Aging and blending would improve waste package
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Environmental Considerations for Alternative Design Concepts and
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Stability (reducing rockfall-induced damage and corrosion) and improve long-term repository
performance.
E.2.2.6.2 Potential Environmental Considerations
The blending feature might require a significantly larger storage pool size. This would increase the size
of the pool storage building, and result in correspondingly higher costs. The Viability Assessment
reference design staging pools have the capacity for about 300 MTHM. This would be reconfigured and
expanded to allow for storage of up to 6,500 MTHM. Expanded pool storage would require additional
resources (steel, concrete, gravel and asphalt, fuel, electricity and water for construction and operation,
but the increases would not be significant (about 10 percent). Waste generation would also increase.
During operations, use of well water will increase by about 15 percent. Well water is used to replace
evaporative losses in the pools. Land use does not increase, hicreases in worker population mean an
increase in the potential for industrial accidents. Cumulative annual dose to workers would increase
slightly, but the average dose to workers would not increase.
If aging is done at the Yucca Mountain site, a surface storage facility would be required. The effects of
the aging feature are identical to the retrieval contingency discussed in the EIS because the same size
storage facility/pad would be needed. The retrieval contingency assumes a surface storage facility able to
handle the entire repository inventory.
E.2.2.7 Continuous Preclosure Ventilation
Continuous preclosure ventilation would provide increased air flow in the emplacement drifts compared
to the reference design preclosure ventilation rate of 0.1 cubic meter (3.5 cubic feet) per second. The
system would be shut off at closure.
Additional excavation would be required for an additional exhaust main. The actual number of
emplacement drifts would not change, but the layout of drifts would vary slightly to accommodate the
additional ventilation shafts. The sizes of the shafts would have to be increased and more would need to
be added. Access drifts and additional connections would have to be added between the exhaust mains
and the shafts.
E.2.2.7.1 Potential Benefits
Continuous ventilation in the preclosure period could reduce the rock wall and air temperature. It could
also remove enough moisture to reduce the length of time the waste packages are exposed to temperature/
moisture conditions that could result in higher corrosion rates. The removal of moisture also would
increase the stability of the ground-support system. In addition, with lower drift temperatures retrieval
would be easier.
E.2.2.7.2 Potential Environmental Considerations
Additional drifts and intake and exhaust shafts would be required to handle the additional airflow
quantities, resulting in additional excavated rock. Additional shaft locations would disturb land surface in
the limited locations available to place the' shafts, and roads would have to be constructed to the shaft
sites. Additional shafts and night lighting at the top of the mountain might be visible from off the Yucca
Mountain site.
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Environmental Considerations for Alternative Design Concepts and
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The changes in repository ventilation would increase emissions of naturally occurring radon-222 and its
radioactive decay products in the air exhausted from the subsurface. Power requirements could increase
substantially during emplacement operations and postclosure monitoring.
The number of workers would increase by less than 10 percent, with an attendant increase in the potential
for industrial accidents.
Closure would be more difficult because there would be additional openings to seal.
E.2.2.8 Drift Diameter
The emplacement drift diameter is a secondary design feature because the diameter is determined by a
number of primary design features. The size of the emplacement drift could directly affect design
considerations such as opening stability (rockfall potential), the extent of the mechanically induced
disturbed zone, and the amount and location of seepage into the drifts.
The drift diameter for the Viability Assessment reference design is 5.5 meters (18 feet). A range of drift
diameters is being considered [from 3.5 meters (11 feet) to 7.5 meters (25 feet)].
E.2.2.8.1 Potential Benefits
A smaller diameter drift is inherently more stable and could reduce the need for ground-support systems,
potentially reducing costs. The smaller drift diameter would also be less susceptible to water seepage. A
larger diameter allows for other modes of emplacement, such as horizontal or vertical borehole
emplacement. Both of these emplacement modes would reduce the potential for damage to waste
packages from rockfall, therefore potentially improving long-term performance of the repository.
E.2.2.8.2 Potential Environmental Considerations
An increase in drift diameter could increase the potential for rockfall (both size and frequency) and
decrease the overall opening stability. Rockfall could breach waste packages or cause lesser damage to
the packages, providing locations for accelerated corrosion. Also, the larger the drift diameter, the more
vulnerable it would be to water entry from seepage flow.
A smaller drift diameter would be inherently more stable in highly jointed rock and a decreased rockfall
size would be anticipated. A change to a smaller diameter could allow modification to the ground-
support system with possible elimination of a full circle drift liner. Although a smaller drift diameter
would be less susceptible to seepage, the smaller diameter drift might result in short-term increases of
temperature, which could affect the characteristics of potential groundwater movement.
Increasing the emplacement drift diameter would result in an increase in the quantity of excavated rock
and increased use of equipment and materials, higher releases of radon-222, and lower ventilation air
velocity. The lower air velocity would result in greater quantities of radon-222 and dust during
development, an important consideration for preventing suspension of respirable silica dust.
A smaller drift diameter, although reducing the potential of radon-222 releases, might not be able to
provide the quantities of air necessary for ventilation without raising velocities to undesirable levels.
Increased drift diameter would require more workers for tunnel boring machine operations, excavated
rock handling, ground-support installation and finishing works, surface equipment operators, and
maintenance. A decrease in the drift diameter would have an opposite affect on the worker requirements;
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Environmental Considerations for Alternative Design Concepts and
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that is, with a larger drift diameter, the additional excavation work would produce an increase in worker
accidents. Larger tunnel boring machines could require substantially more electrical power.
E.2.2.9 Drift Spacing and Waste Pacl<age Spacing
In repository design, thermal load refers to a density at which the waste packages will be emplaced in the
repository. The Viability Assessment reference design involves emplacement of waste packages in drifts
in a horizontal mode, and thermal load is directly related to the emplacement drift and waste package
spacing. The Viability Assessment reference design used a spacing of 28 meters (92 feet) between drifts.
For a given drift spacing, emplacement of waste packages can be arranged by using point load (waste
package spacing determined based on individual waste package characteristics, such as mass content or
equivalent heat output of each waste package), or line load [waste packages are emplaced nearly end to
end that is, with a O.I -meter (0.3-foot) gap with no considerations of individual waste package
characteristics].
The point load approach was used for the Viability Assessment reference design. Waste-package spacing
was determined based on mass content of waste packages, to achieve an overall area mass loading of
85 MTHM per acre for commercial spent nuclear fuel.
The line load method would be expected to provide a more intense and uniform heat source along the
length of emplacement. An increase in emplacement drift spacing would be required in conjunction with
line loading to maintain a constant overall thermal loading density (for example, 85 MTHM per acre).
E.2.2.9. 1 Potential Benefits
The line load approach would keep the emplacement drifts hot and dry longer and would decrease the
amount of water that could contact waste packages. Consequently, waste package performance could be
improved. The line load approach would also reduce the number of emplacement drifts needed for waste
emplacement. However, the concentrated heat load in the drifts could require continuous ventilation of
emplacement drifts to meet the near-field temperature requirements. Continuous ventilation is discussed
in Section E.2.2.7.
E.2.2.9.2 Potential Environmental Considerations
Line loading would require excavation of about 30 fewer emplacement drifts, with correspondingly less
excavated rock, dust, and pollutants from diesel- and gasoline-powered equipment and vehicles.
Decreased excavation would also reduce radon-222 release in the underground facility. However,
decreasing the waste package spacing would result in potentially large increases in the rock temperatures
in and near the emplacement drifts. This could create the need for continuous ventilation of emplacement
drifts, which could increase emissions of naturally occurring radon-222 and its radioactive decay products
in the air exhausted from the subsurface.
The reduction in total work and material requirements would be expected to be linearly proportional to
the reduction in required drift length. Fewer work hours would also result in less potential for industrial
accidents during construction. Decreased emplacement drift excavation would reduce the demand for
electric power, equipment fuel, construction materials, and site services. However, the higher drift
temperature associated with the line load option could require continuous ventilation of emplacement
drifts.
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Environmental Considerations for Alternative Design Concepts and
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E.2.2.10 Near-Field Rock Treatment
Near-field rock treatment involves injection of a grout material into the cracks in a portion of the rock
above each emplacement drift to reduce the hydraulic conductivity of the treated rock. Injection would
start at least 6 meters (20 feet) above the drift crown and would form a zone at least 4 meters (13 feet)
thick, extending at least 6 meters on each side of the drift. Injection would be through holes 2.5 to
5 centimeters (1 to 2 inches) in diameter drilled from inside each drift prior to waste emplacement.
Injection pressures would not exceed a certain minimum pressure, selected to limit rock fracturing or joint
opening.
The candidate materials include Portland cement grout, sodium silicate, bentonite (a clay), and calcite.
E.2.2.10.1 Potential Benefits
Reducing the hydraulic conductivity of the rock would improve long-term repository performance by
reducing or retarding postclosure water seepage into the drifts.
E.2.2.10.2 Potential Environmental Considerations
Installation of the grout material would require additional labor hours, with an associated change in the
potential for industrial accidents.
E.2.2.11 Surface Modification - Alluvium Addition
Covering the surface of Yucca Mountain above the repository footprint with alluvium could decrease the
net infiltration of precipitation water into the repository by increasing evapotranspiration. To cover the
mountain with alluvium, the surface of the mountain would be modified to prevent the alluvium from
washing away. Ridge tops on the eastem flank of Yucca Mountain would be removed and the excavated
rock placed in Solitario Canyon and in Midway Valley or used to fill the alluvium borrow pit. The
maximum slope of the ground surface remaining would be approximately 10 percent. Alluvium
[approximately 2 meters (7 feet) thick] would be placed on the new surface and vegetation would be
established. New haul roads to move the necessary materials would have to be constructed.
E.2.2.1 1 .1 Potential Benefits
Reduced net infiltration would improve long-term repository performance. However, there is uncertainty
about the permanence of both the vegetation and the alluvium that would be added to the surface of
Yucca Mountain.
E.2.2.1 1 .2 Potential Environmental Considerations
Approximately 8 square kilometers (2,000 acres) on Yucca Mountain would be resloped and covered.
The excavated material would cover 4.8 square kilometers (1,200 acres) in the fill area in Solitario
Canyon. The borrow pit would be about 5.2 square miles (1,300 acres). Additional access roads would
also be needed. Yucca Crest would be lower by approximately 30 to 60 meters (98 to 197 feet) the ridges
on the east side of Yucca Crest would be lowered by as much as 80 meters (262 feet). Quantities of
material to be moved would include:
• Total rock cut from Yucca Mountain 220 million cubic meters (17,600 acre-feet)
• Total alluvium removed from the alluvium borrow pit (probably in Midway Valley) about 22 million
cubic meters (17,600 acre-feet)
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Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
The operation would be equivalent to a major, large-scale open pit mining operation. It would likely
require a labor force of about 75 people per shift. There would be an increase in the potential for
industrial accidents because of the additional work. Generation of particulate emissions (fugitive dust)
and gaseous criteria pollutant emissions from vehicles would increase.
There would be alterations to natural drainage; however, the potential for flooding would not increase
with proper design.
The view to and from Yucca Mountain would be altered. Mining operations at the top of the mountain
would be visible for some distance, and the mountain would be considerably lower. Vegetation would be
restored because the design requires vegetation as part of the evapotranspiration process. The operation
would be carried out on three shifts, and night lighting on the top of the mountain could be visible to the
public.
E.2.2.12 Surface Modification - Drainage
Surface modification could reduce infiltration at the surface of the mountain. Net infiltration into Yucca
Mountain could be significantly decreased if the thin alluvium layer over the footprint of the repository
were removed to promote rapid runoff of the surface water. It has been shown that where the alluvium is
thin, it retains the surface water and allows it to infiltrate into the unsaturated zone. Where bedrock is
exposed on slopes, the water runs off rapidly and net infiltration is very small or reduced to zero.
The thin alluvium layer would be stripped from the topographic surface above the repository footprint and
a 300-meter (984-foot) buffer surrounding it.
E.2.2.12.1 Potential Benefits
Reduced infiltration would result in improved long-term repository waste isolation. However, there is
uncertainty about the permanence of alluvium removal. In addition, while infiltration might be reduced
on the top of the mountain, infiltration could increase in other areas because of the higher volumes of
surface water runoff.
E.2.2.12.2 Potential Environmental Considerations
The amount of land modified to improve drainage would be approximately 1,1(X) acres, located mainly on
the eastern flank of Yucca Mountain. Additional road construction would also be required. The removed
alluvium, about 2. 1 million cubic meters (2.7 million cubic yards), would be placed in Midway Valley.
There would be alterations to natural drainage, and the increased runoff could increase the potential for
flooding. The landforms would be changed only slightly because of the thin [less than 0.5-meter
(1.6-foot) thick] alluvium that would be removed. Any existing vegetation on the side of the ridges
would be removed during the process of alluvium removal. Bare bedrock would be exposed, which
would discourage vegetation from growing except from cracks in the rock.
Additional workers would be required, and there would be an accompanying increase in the potential for
industrial accidents.
Night lighting would be needed to support this operation that could be visible from off the site.
E-16
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
E.2.2.13 Higher Thermal Loading
Higher thermal loading would keep the drift temperature above the boiling point longer, thereby
minimizing the amount of moisture around the waste package during a longer postclosure period. The
higher thermal loading could also have adverse effects on the surrounding rock. This feature could also
be combined with aging to achieve greater mass loading per acre of repository area.
Higher thermal loads could be achieved by either decreasing drift spacing, by placing waste packages
closer together in the drift, or by a combination of drift spacing and waste package spacing. In all three
cases, the increased number of waste packages in a given area would result in a higher thermal load to a
given area of the repository.
The benefits and environmental considerations associated with this feature would be similar to those
discussed under Drift Spacing and Waste Package Spacing (Section E.2.2.9).
E.2.3 REPOSITORY DESIGNS TO SUPPORT OPERATIONAL AND/OR COST
CONSIDERATIONS
E.2.3.1 Design Alternative 4, Enhanced Access
The purpose of the enhanced access design would be to provide additional shielding around the waste
package to allow for personnel accessibility during waste package loading, transfer to the drift,
emplacement, and performance confirmation. Shielding would lower the dose rate to less than 25
millirem per hour. Enhanced access could be provided by:
• Additional shielding integral to the waste package
• Supplemental (separate from the waste package) shielding in the emplacement drifts only
• Portable shielding for personnel to access the drift
E.2.3.1. 1 Potential Benefits
The major benefit of these three options would be to provide access to the emplacement drifts so
personnel could carry out performance confirmation activities. Enhanced access designs could also offer
increased access for maintenance and ease of operations, and the potential elimination of some remote
handling equipment. If shielding were left in place at closure, it could provide additional protection for
waste packages from rock falls.
E.2.3. 1.2 Potential Environmental Considerations
Increased personnel access would increase occupational exposure, even with the additional shielding.
Enhanced access would decrease the number of observation and performance confirmation drifts needed,
and slightly decrease the volume of excavated rock piles.
The addition of shielding to waste packages would result in increased materials usage. Shielding
materials could be steel, concrete, magnetite concrete (concrete with iron shot included), or Ducrete®
(concrete with depleted uranium included).
E.2.3.2 Design Alternative 5, Modified Waste Emplacement Mode
In a modified waste emplacement design, unshielded waste packages would be emplaced in a
configuration in which the repository's natural or engineered barriers would provide shielding. Examples
E-i7
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
include placing waste packages in boreholes drilled into the floor or wall of emplacement drifts, in
alcoves off the emplacement drifts, in trenches at the bottom of the emplacement drifts, or in short cross
drifts excavated between pairs of excavated drifts. In each case, some type of cover plug would be used
to shield radiation in the emplacement drifts.
Unshielded waste packages, which in some designs would have a smaller capacity than specified in the
Viability Assessment reference design, would be used.
E.2.3.2.1 Potential Benefits
Natural or engineered barriers would enhance human access, reduce performance confirmation costs, and
facilitate conducting inspections and maintaining ground support. Retrieval operations would also be
easier because of easier access.
E.2.3.2.2 Potential Environmental Considerations
The footprint of the repository would not change, but the amount of excavated rock would increase. The
vertical borehole emplacement concept would generate the most additional excavated rock. Peak power
consumption would increase substantially because of the use of additional boring machines.
E.2.3.3 Design Alternative 8, Modular Design (Phased Construction)
Modular design is an alternative that could reduce annual expenditures during construction if annual
funding is constrained below that required for the Viability Assessment reference design. This alternative
would include staged modular construction of repository surface and subsurface facilities.
The modularized Waste Handling Building would be designed to handle specific types of waste forms
and quantities. The modular concept would include one Waste Handling Building completed in modular
phases or two separate buildings constructed in sequence.
E.2.3.3.1 Potential Benefits
The primary benefit would be leveled cash flow during construction.
E.2.3.3.2 Potential Environmental Considerations
The dual buildings would increase the overall size of the Waste Handling Building by an estimated
10 percent. The Radiologically Controlled Area could increase by about 10 percent or less. Operating
times (years of operation) would be extended and operations would be at a lower rate.
Some options would involve receipt of spent nuclear fuel from reactor sites prior to the start of
emplacement that could increase worker dose because it would have to be handled twice.
E.2.3.4 Rod Consolidation
Both pressurized-water reactor and boiling-water reactor fuel assemblies have fuel rods arranged in
regular square arrays with rod-to-rod separation maintained by the fuel assembly hardware. Rod
consolidation would involve eliminating this separation and bringing the fuel rods into close contact.
Reducing the volume taken up by fuel assemblies would allow the capacity of waste packages to be
increased and/or the size of waste packages to be reduced. Consolidation could be done at either the
current spent fuel storage locations or at the repository.
E-18
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
Rod consolidation would be accomplished by removing fuel rods from an assembly, repackaging the rods
in a denser arrangement in a suitable canister, and loading the new canister into a waste container. This
process could occur either in a pool or in a dry (hot cell) environment.
E.2.3.4.1 Potential Benefits
A reduced number or size of waste packages would be possible and could result in reduced emplacement
costs. If rod consolidation took place at the reactor sites, waste transportation requirements might be
reduced.
E.2.3.4.2 Potential Environmental Considerations
Because of the disassembly operations, the size of the Waste Handling Building would more than double
in area if rod consolidation were done at the repository. With the large number of fuel rod handling
operations in the hot cells, there would be a greater potential for radiological releases due to fuel handling
accidents (such as dropping a fuel rod/assembly).
The number of workers at the repository could increase if rod consolidation were performed at the
repository. With an increase in the number of fuel handling operations, the number of fuel handling
accidents would increase and result in a small increase in radiological exposure for onsite workers.
Approximately 10 to 40 kilograms (22 to 88 pounds) of leftover, nonfuel components from each as-
received fuel assembly would be packaged as Class C or Greater-Than-Class-C low-level wastes. In
addition, low-level waste would be generated by decontamination and disposal of equipment. Low-level
waste would be transported to the Nevada Test Site or other appropriate facility for disposal. Greater-
than-Class-C wastes could be disposed of offsite or in the repository with approval of the U.S. Nuclear
Regulatory Commission.
Waste packages containing consolidated fuel rods might result in higher cladding temperatures, which
could damage the cladding and have negative impacts on waste isolation performance.
E.2.3.5 Timing of Repository Closure
The first option assumes that the subsurface facilities would be fully maintained to the same level of
readiness during the 300-year period as planned for the 100-year period assumed for the Viability
Assessment reference design. There would be continuous ventilation during the entire 3(X)-year period.
The second option assumes the Nuclear Regulatory Commission would have approved completion of the
Performance Confirmation Program at the end of the first 1(X) years, and that continued access to the
emplacement drifts would no longer be required. The second option considers that ventilation,
maintenance, and repairs would be reduced to a minimum for cost considerations, but that temperatures
would be maintained at 50°C (122°F) or less for human access to the subsurface (nonemplacement)
facilities.
E.2.3.5.1 Potential Benefits
Extending the period before final closure would allow for reduction of waste package heat output,
extended monitoring, and extended retrieval period for the waste.
E-19
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
E.2.3.5.2 Potential Environmental Considerations
Delayed closure of the repository would lengthen the time that land would remain disturbed through the
occupation of surface facilities necessary to support extended operations from 100 to 300 years. It would
delay the reclamation of surface stockpiles retained for filling the mains, ramps, and shafts.
The release of radon-222 from excavations is proportional to time. Delayed closure from 100 to
300 years would increase the emissions of radon-222 by a factor of approximately 3.6.
The number of workers required for monitoring would not change. However, the number of labor hours
required, compared to the Viability Assessment reference design monitoring period, would be 3.6 times
the number required for closure at 100 years. The base case scenario requires the periodic retrieval of
waste packages for performance confirmation testing. An increase in the monitoring period from 76 to
276 years would increase radiation exposure due to increased waste package handling. More frequent
inspections would be likely during this extended period due to aging. Additionally, emplacement drifts
maintenance would require removal and re-emplacement of waste packages. An increased monitoring
period would increase the potential for industrial accidents and radiological exposure.
E.2.3.6 Maintenance of Underground Features and Ground Support
A maintenance program in the emplacement drifts would be needed to accommodate an extended long-
term repository service life and to reduce the risk of keeping the repository open for an additional
200 years. Repository emplacement drift ground support components would have to be designed and
maintained for a service life of greater than 300 years, including closure and retrieval times.
E.2.3.6.1 Potential Benefits
The benefits are the same as those listed in Section E.2.3.5.1
E.2.3.6.2 Potential Environmental Considerations
Some types of maintenance in the emplacement drifts would require retrieval of waste packages for
maintenance access. Blast cooling would be needed to lower the temperature to below 50°C for worker
access. There could be additional radiological exposure to workers.
E.2.3.7 Waste Package Self-Shielding
In the Viability Assessment reference design, handling of waste packages in the emplacement drifts
would be performed remotely, and human access to the emplacement drifts would be precluded when
waste packages are present. Waste package self-shielding would reduce the radiation in the drifts to
levels such that personnel access would be possible. This would allow direct access to the performance
confirmation instrumentation, and maintenance and repair in the drifts.
Self-shielding would be accomplished by adding a shielding material around the waste packages.
Candidate materials include A516 carbon steel, concrete with depleted uranium (Ducrete®), magnetite
concrete, and a composite material of boron-polyethylene and carbon steel.
The amount of shielding would depend on the target radiation dose level in the drift environment. For a
25-millirem-per-hour waste package contact dose, the estimated thickness of the concrete would be about
0.6 meter (2 feet). For higher contact doses, less shielding material would be required.
E-20
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
E.2.3.7.1 Potential Benefits
Monitoring, maintenance, and retrieval would be easier with contact handling of the waste packages.
E.2.3.7.2 Potential Environmental Considerations
Self-shielding could not be used with high thermal loading because the shielding would provide a thermal
barrier that would result in excessive fuel cladding temperature. Smaller waste packages would maintain
a constant outside diameter but would also require about four times as many waste packages and more
drifts. Radon-222 emissions would increase in proportion to the additional excavation.
Concrete shielding would be applied at the repository, and the number of workers would slightly increase,
as would the number of industrial accidents. There could be a reduction in radiological exposure to
workers during emplacement operations. The concrete shielding could degrade the long-term
performance of the waste packages.
E.2.3.8 Repository Horizon Elevation
This feature considers a two-level repository to increase repository capacity without moving out of the
characterized area.
One two-level concept would divide the Viability Assessment reference design layout along a north-south
axis and would relocate the western half above the eastem half. A second two-level concept would
duplicate the Viability Assessment reference design layout 50 meters (164 feet) above the current
footprint. The thermal loading of each level could be adjusted to increase the capacity.
E.2.3.8.1 Potential Benefits
There would be two potential advantages to repository long-term performance. Increased thermal load
would potentially enhance the umbrella effect (this could reduce the amount of water that could come in
contact with the waste package). There would also be added flexibility in emplacing waste packages on
the lower level, which could be shielded from moisture infiltration by the upper level horizon.
Retrieval could be accomplished more quickly due to the ability to operate two independent retrieval
operations at the same time.
E.2.3.8.2 Potential Environmental Considerations
The first two-level concept could use slightly less land area to store excavated rock because less material
would be excavated. The second two-level concept could double the excavation and double the excavated
rock volume that would require storage.
Surface soil temperatures could increase due to locating waste closer to the surface and/or increasing
thermal loading per acre.
Construction of the full size footprint two-tier repository would require slightly less than double the
number of workers and a longer construction period, with associated changes in the potential for
industrial accidents. Power consumption would approximately double.
E-21
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
E.3 Enhanced Design Alternatives
Enhanced Design Alternatives are combinations of the alternatives and design features described in
preceding sections. These concepts were developed to cover a range of potential repository designs as
part of the License Application Design Selection Process described in Section E.1.2. Enhanced Design
Alternatives are intended to be improvements to the basic design alternatives discussed in Section E.2.
Five Enhanced Design Alternatives are described below, along with the design concepts that led to their
development. Potential benefits and environmental considerations are discussed in the sections above
dealing with the design alternative and design features incorporated into each Enhanced Design
Alternative.
At the time of development of this appendix, the Enhanced Design Alternatives discussed below had
been developed, but documentation of the Enhanced Design Alternative development process was
forthcoming. That documentation was scheduled to be complete in May 1999. The Enhanced Design
Alternatives described in the following sections are preliminary and based on observations of the License
Application Design Selection Process and informal discussions with process participants.
E.3.1 ENHANCED DESIGN ALTERNATIVE I
Enhanced Design Alternative I is a low-temperature design intended to remove uncertainties and
modeling difficulties associated with above-boiling temperatures. Lower temperatures would mean less
disturbance of the subsurface and limit the combined effects of thermal, hydrological, and geochemical
processes that are more pronounced in above-boiling-temperature environments.
The goals of Enhanced Design Alternative I are to keep the drift wall temperature below the boiling point
of water and the commercial fuel cladding temperature below 350°C (662°F). This would be achieved for
the Enhanced Design Alternative I design by limiting areal mass loading to 45 MTHM per acre,
increasing the size of the repository to 6 square kilometers (1,500 acres), and using smaller waste
packages. Drift spacing would be 43 meters (141 feet) between drift centerlines, with an average end-to-
end waste package spacing of 3 meters (10 feet). Preclosure ventilation would use two intake and three
exhaust shafts.
The waste package design for this Enhanced Design Alternative would consist of two layers, with
Alloy-22 on the outside and 316L stainless steel (nuclear grade) on the inside. Flexible waste package |
spacing would be used to control the drift temperature. Blending would be used to reduce the maximum 1
thermal output of a waste package to 6.7 kilowatts. To optimize selection of waste for emplacement,
additional surface storage capacity above and beyond that in the Viability Assessment reference design
would be necessary. A 2-centimeter (0.8-inch)-thick titanium-7 drip shield, to be placed over the waste
package just prior to closure, is included in this design to provide defense in depth.
I
This design allows human access using blast cooling and portable shielding [15 centimeters (6 inches)
stainless steel and 7.5 centimeters (3 inches) borated polyethylene].
The major disadvantage of this design is that it uses all of the available space in the upper repository
block. Another disadvantage is that it uses smaller waste packages, requiring about 6,000 more waste
packages than other Enhanced Design Alternatives.
E.3.2 ENHANCED DESIGN ALTERNATIVE II
Enhanced Design Alternative II is a moderate temperature design intended to keep commercial fuel J
cladding temperature below 350°C (662°F) and to keep the boiling fronts from merging in the rock walls
E-22
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
between the drifts. Keeping a non-boiling area between the drifts ensures that there would be sufficient
area between the drifts that would be below the boiling point to allow water to drain. The areal mass
loading could be up to 60 MTHM per acre and still achieve these goals.
The waste package design would consist of two layers with Alloy-22 on the outside and 316L stainless
steel on the inside. Blending would be used to reduce the maximum heat output of a waste package to
9 kilowatts. The emplacement area would be 4.3 square kilometers (1,064 acres), and the waste package
design would be the same as for Enhanced Design Alternative I. The Enhanced Design Alternative II
design would use closely spaced waste packages, line loading, and a drift spacing of 8 1 meters (266 feet).
To optimize selection of waste for emplacement, additional surface storage capacity above and beyond
that in the Viability Assessment reference design would be necessary. This design also includes backfill,
a 2-centimeter (0.8-inch)-thick titanium-7 drip shield placed just prior to closure, as in Enhanced Design
Alternative I. Continuous ventilation would be used for the 50-year preclosure period.
An advantage of this design is that it would reduce or avoid uncertainties associated with the thermal
period or thermal pulse where large quantities of water could pool above the repository area. The cooler
pillars between the drifts would allow for drainage of waters. However, an uncertainty is that the
drainage of water has not been demonstrated. Another advantage is that the design provides flexibility for
modification to either a hotter or cooler design.
E.3.3 ENHANCED DESIGN ALTERNATIVE III
Enhanced Design Alternative III is a high thermal load design. The goals are to keep the drift wall
temperatures below 200°C (329°F), the commercial fuel cladding temperature below 350°C (662°F), and
to ensure that the waste package surface temperature cools to below 80°C (176°F) before the relative
humidity at the waste package surface rises above 90 j)ercent. These goals would be met with an 85
MTHM per acre loading, close [0.1 meter, (0.3 foot)] spacing of line-loaded waste packages, and a drift
spacing of 56 meters (184 feet).
Two different waste packages are considered (Enhanced Design Alternatives Illa and nib). The
Enhanced Design Alternative Ilia waste package would use a two-layer design with 2 centimeters
(0.8-inch) of Alloy-22 over 5 centimeters (2 inches) of 3 16L stainless steel (as in Enhanced Design
Alternatives I, II, and V). The Enhanced Design Alternative Illb waste package design would use a waste
package with an outer layer of 2.2 centimeters (0.9 inch) of Alloy-22 over 1.5 centimeter (0.6 inch) of
titanium-7 that have been shrink-fitted together, and a 4-centimeter (1.6-inch) inner layer of 316L
stainless steel that would fit loosely (gap of 4 millimeters or less) inside the Alloy-22/titanium-7 shell.
Blending would not be used in Enhanced Design Alternative III. However, preclosure ventilation of at
least 5 cubic meters (177 cubic feet) per second would be needed for a minimum of 50 years to achieve
the temperature goals of this Enhanced Design Alternative. This would require two intake and three
exhaust shafts in addition to the access tunnels. Enhanced Design Alternative HI also includes a
titanium-7 drip shield.
The advantage of Enhanced Design Alternative III is that the surface of the waste package is predicted to
cool below 80°C (176°F) before the relative humidity exceeds 90 percent, thus avoiding the worst of the
corrosive, warm-moist environment after closure. The disadvantages are the uncertainties connected with
temperatures over 100°C (212°F).
E-23
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
E.3.4 ENHANCED DESIGN ALTERNATIVE IV
Enhanced Design Alternative FV is a shielded waste package design located entirely in the upper block
with a high thermal load (85 MTHM per acre). The goals of this Enhanced Design Alternative are to
keep the gamma radiation dose at the surface of the waste package below 200 millirem per hour, keep the
fuel cladding below 350°C (662°F), and keep the emplacement drifts dry for thousands of years.
The waste package would be 30-centimeter (12-inch)-thick A516 steel, and it would have an integral filler
that acted as a sponge for oxygen. Waste packages would be line-loaded with a separation of 0.1 meter
(0.3 feet). Continuous ventilation at 2 to 5 cubic meters (71 to 177 cubic feet) per second would be
required for the 50-year preclosure period. Two intake and three exhaust shafts would be required in
addition to the access tunnels. Human access would require blast cooling to reduce temperatures in the
drift using a portable 5-centimeter (2-inch)-thick borated polyethylene neutron shielding over the waste
packages. Backfill material and drip shields are used in this Enhanced Design Alternative.
The Enhanced Design Alternative IV waste packages would weigh 18,140 metric tons (20 tons) more
than those used with other Enhanced Design Alternatives. Since this Enhanced Design Alternative
requires a hot postclosure environment to be successful, it would be necessary to manage the waste
stream to ensure uniform heat in the repository. Backfill would be placed at closure.
If this design concept does not properly control temperature and relative humidity to protect the drip
shield, the carbon steel waste packages would be expected to fail much earlier than the waste packages in
the other Enhanced Design Alternatives.
E.3.5 ENHANCED DESIGN ALTERNATIVE V
Enhanced Design Alternative V is a very high thermal load alternative (150 MTHM per acre) and covers
the smallest area [168 square kilometers (420 acres)] of the five Enhanced Design Alternatives. The
purpose of the very high thermal load is to provide a hot, dry drift environment for thousands of years and
avoid extended periods of warm, moist conditions. The goals of this Enhanced Design Alternative were
to have drift wall temperatures less than 225 °C (437°F) to maintain stability, commercial fuel cladding
temperature less than 350°C, and to keep the drift dry for several thousand years.
Waste blending would be required so that waste temperatures were all within 20 percent of the average.
Waste packages would be 2-centimeter (0.8-inch) Alloy-22 over 5-centimeter (2-inch) 316L stainless
steel, and they would be line loaded with a 0.1-meter (0.3-foot) spacing between waste packages. To
optimize selection of waste for emplacement, additional surface storage capacity above and beyond that
in the Viability Assessment reference design would be necessary. Drift spacing would be 32.4 meters
(106 feet). Preclosure ventilation would reduce air and drift temperatures and remove moisture from the
drifts. Four air shafts as well as three access tunnels would be needed. Titanium-7 drip shields would be
placed at the time of closure.
The advantage of this design is that it would be located entirely in the lower block of the repository,
where the percolation rate is less than half that in the upper block. However, access to the lower block
would require a third tunnel. In addition, postclosure conditions could lead to localized corrosion and
early failure of waste packages. The high temperatures also could create the possibility that the cladding
temperature goal would be exceeded for some waste packages.
E-24
Environmental Considerations for Alternative Design Concepts and
Design Features for the Proposed Monitored Geologic Repository at Yucca Mountain, Nevada
REFERENCE
DOE 1998 DOE (U.S. Department of Energy), 1998, Viability Assessment of a
Repository at Yucca Mountain, DOE/RW-0508, Office of Civilian
Radioactive Waste Management, Washington, D.C. [U.S.
Government Printing Office, MOL. 1998 1007.0027, Overview;
MOL. 1998 1007.0028, Volume 1; MOL. 1998 1007.0029, Volume 2;
MOL. 1998 1007.0030, Volume 3; MOL.19981007.0031, Volume 4;
MOL. 1998 1007.0032, Volume 5]
E-25
''rrr/7/hr>^
Appendix F
Human Health Impacts Primer
and Details for Estimating Healtii
Impacts to Workers from Yucca
Mountain Repository Operations
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
TABLE OF CONTENTS
Section Page
F. 1 Human Health Impacts from Exposure to Radioactive and Toxic Materials F-1
F.1.1 Radiation and Human Health F-1
F.1.1.1 Radiation F-1
F.1.1.2 Radioactivity, Ionizing Radiation, Radioactive Decay, and Fission F-2
F.1.1. 3 Exposure to Radiation and Radiation Dose F-3
F.1.1.4 Background Radiation from Natural Sources F-4
F.1.1.5 Impacts to Human Health from Exposure to Radiation F-4
F. 1.1.6 Exposures from Naturally Occurring Radionuclides in the Subsurface
Environment F-6
F.1.2 Exposure to Toxic or Hazardous Materials , F-7
F.1.3 Exposure Pathways F-10
F.2 Human Health and Safety Impact Analysis for the Proposed Action Inventory F-1 1
F.2. 1 Methodology for Calculating Occupational Health and Safety Impacts F-1 1
F.2.2 Data Sources and Tabulations F-12
F.2.2.1 Work Hours for the Repository Phases F-12
F.2.2.2 Workplace Health and Safety Statistics F-12
F.2.2.3 Estimates of Radiological Exposures F-16
F.2.3 Compilation of Detailed Results for Occupational Health and Safety Impacts F-19
F.2.3.1 Occupational Health and Safety Impacts During the Construction Phase F-19
F.2.3. 1.1 Industrial Hazards to Workers F-19
F.2.3.1.2 Radiological Health Impacts to Workers F-20
F.2.3.2 Occupational Health and Safety Impacts During the Operations Period F-21
F.2.3.2.1 Industrial Safety Hazards to Workers F-21
F.2.3.2.2 Radiological Health Impacts to Workers F-22
F.2.3.3 Occupational Health and Safety Impacts to Workers During the Monitoring
Period F-26
F.2.3.3. 1 Health and Safety Impacts to Workers from Workplace Industrial Hazards F-26
F.2.3. 3. 2 Radiological Health Impacts to Workers F-27
F.2.3.3.2.1 Surface Facility Workers F-27
F.2.3.3.2.2 Subsurface Facility Workers F-27
F.2.3.4 Occupational Health and Safety Impacts During the Closure Phase F-28
F.2.3.4.1 Health and Safety Impacts to Workers from Workplace Industrial Hazards F-28
F.2.3.4.2 Radiological Health hnpacts to Workers F-28
F.3 Human Health and Safety Analysis for Inventory Modules 1 and 2 F-31
F.3. 1 Methodology for Calculating Human Health and Safety Impacts F-32
F.3.2 Data Sources and Tabulations F-32
F.3.2. 1 Full-Time Equivalent Worker- Year Estimates for the Repository Phases for
Inventory Modules 1 and 2 F-32
F.3.2.2 Statistics on Health and Safety Impacts from Industrial Hazards in the Workplace F-34
F.3.2.3 Estimates of Radiological Exposure Rates and Times for Inventory
Modules 1 and 2 F-34
F.3.3 Detailed Human Health and Safety Impacts to Workers - Inventory Modules 1
and 2 F-34
F.3.3. 1 Construction Phase F-34
F.3.3. 1.1 Industrial Hazards to Workers F-34
F.3.3. 1.2 Radiological Health Impacts to Workers F-34
F.3.3.2 Operation and Monitoring Phase F-36
F.3.3.2.1 Health and Safety Impacts to Workers from Industrial Hazards F-36
F.3.3.2.2 Radiological Health Impacts to Workers F-36
F-iii
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Section Page
F.3.3.3 Closure Phase F-39
F.3.3.3.1 Health and Safety Impacts to Workers from Industrial Hazards F-39
F.4 Human Health and Safety Impact Analysis for the Retrieval Contingency F-40
F.4.1 Methodology for Calculating Human Health and Safety Impacts F-41
F.4.2 Data Sources and Tabulations F-41
F.4.2.1 Full-Time Equivalent Work- Year Estimates for the Retrieval Contingency F-41
F.4.2.2 Statistics on Health and Safety Impacts from Industrial Hazards in the Workplace F-42
F.4.2.3 Estimated Radiological Exposure Rates and Times for the Retrieval Contingency F-42
F.4.3 Detailed Results for the Retrieval Contingency F-43
F.4.3.1 Construction Phase F-43
F.4.3. 1. 1 Human Health and Safety Impacts to Workers from Industrial Hazards F-43
F.4.3.2 Operations Period F-43
F.4.3. 2.1 Health and Safety Impacts to Workers from Industrial Hazards F-43
F.4.3. 2. 2 Radiological Health and Safety Impacts to Workers F-44
References F-46
F-iv
Human Health Impacts Primer and Details for Estimating Health
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LIST OF TABLES
Table Page
F-1 Estimated full-time equivalent worker years for repository phases F-13
F-2 Health and safety statistics for estimating industrial safety impacts common to the
workplace F-14
F-3 Yucca Mountain Project worker industrial safety loss experience F-15
F-4 Correction factors and annual exposures from radon-222 and its decay products for
each of the project phases or periods under the Proposed Action F-17
F-5 Radiological exposure data used to calculate worker radiological health impacts F-18
F-6 Annual involved subsurface worker exposure rates from waste packages F-19
F-7 Industrial hazard health and safety impacts to surface facility workers during
construction phase F-20
F-8 Industrial hazard health and safety impacts to subsurface facility workers during
construction phase F-20
F-9 Radiological health impacts to subsurface facility workers from radon exposure
during construction phase F-21
F-10 Radiological health impacts to subsurface facility workers from ambient radiation
exposure during construction phase F-21
F-U Industrial hazard health and safety impacts to surface facility workers during waste
receipt and packaging period F-22
F-1 2 Industrial hazard health and safety impacts to subsurface facility workers during drift
development period F-22
F-13 Industrial hazard health and safety impacts to subsurface facility workers during
emplacement period F-23
F-14 Radiological health impacts to subsurface facility workers from waste packages
during emplacement period F-23
F-15 Radiological health impacts to subsurface facility workers from ambient radiation
during emplacement period F-24
F-1 6 Radiological health impacts to subsurface facility workers from ambient radiation
during drift development period F-24
F-17 Radiological health impacts to subsurface facility workers from airborne radon-222
during emplacement period F-25
F-18 Radiological health impacts to subsurface facility workers from airborne radon-222
during development period F-25
F-19 Industrial hazard health and safety impacts to surface facility workers during
decontamination period F-26
F-20 Industrial hazard health and safety impacts to surface facility workers during
monitoring period F-26
F-21 Industrial hazard health and safety impacts for subsurface facility workers during
monitoring period F-26
F-22 Radiological health impacts to surface facility workers during decontamination
period F-27
F-23 Radiological health impacts to subsurface facility workers during a 50-year work
period during a 76-year monitoring period F-27
F-24 Radiological health impacts to workers during a 26-year and a 276-year monitoring
period, dual-purpose canister packaging scenario F-28
F-25 Industrial hazard health and safety impacts to surface facility workers during closure
phase F-29
F-26 Industrial hazard health and safety impacts to subsurface facility workers during
closure phase F-29
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Human Health Impacts Primer and Details for Estimating Health
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Table Page
F-27 Radiological health impacts to subsurface facility workers from waste package
radiation exposures during closure phase F-30
F-28 Radiological health impacts to subsurface facility workers from ambient radiation
exposures during closure phase F-30
F-29 Radiological health impacts to subsurface facility workers from radon-222 exposure
during closure phase F-31
F-30 Expected durations of the Proposed Action and Inventory Modules 1 and 2 F-32
F-31 Full-time equivalent work years for various repository periods for Inventory Modules
land 2 F-33
F-32 Correction factors and annual exposures from radon-222 and its decay products for
the project phases or periods for Inventory Modules I and 2 F-34
F-33 Industrial hazard health and safety impacts to subsurface facility workers during
construction phase - Inventory Module I or 2 F-35
F-34 Radiological health impacts to subsurface facility workers from radon inhalation and
natural exposure for the construction phase - Inventory Modules 1 and 2 F-35
F-35 Industrial hazard health and safety impacts for surface facility workers during a 38-
year operations period by packaging option - Inventory Module 1 or 2 F-36
F-36 Industrial hazard health and safety impacts for subsurface facility workers for
development and emplacement period - Inventory Module 1 or 2 F-37
F-37 Industrial hazard health and safety impacts for subsurface facility workers during
monitoring period - Inventory Module 1 or 2 F-37
F-38 Industrial hazard health and safety impacts by packaging option to workers during
surface facility decontamination and monitoring period - Inventory Module 1 or 2 F-37
F-39 Radiological health impacts to surface facility workers for a 38-year operations
period - Inventory Module 1 or 2 F-38
F-40 Radiological health impacts to subsurface workers for emplacement and drift
development during operations period - Inventory Module 1 or 2 F-38
F-41 Radiological health impacts to surface facility workers for decontamination and
monitoring support - Inventory Module 1 or 2 F-39
F-42 Radiological health impacts to subsurface facility workers for a 62-year monitoring
period - Inventory Module 1 or 2 F-39
F-43 Industrial hazard health and safety impacts to surface workers during the closure
phase - Inventory Module 1 or 2 F-40
F-44 Health and safety impacts to subsurface facility workers from industrial hazards
during the closure phase - Inventory Module 1 or 2 F-40
F-45 Full-time equivalent work-year estimates for retrieval F-42
F-46 Statistics for industrial hazard impacts for retrieval F-42
F-47 Radiological doses and exposure data used to calculate worker exposures during
retrieval F-43
F-48 Industrial hazard health and safety impacts to workers during construction F-44
F-49 Industrial hazard health and safety impacts to surface facility workers during retrieval F-44
F-50 Industrial hazard health and safety impacts to subsurface facility workers during
retrieval F-45
F-51 Radiological health impacts to surface facility workers from waste handling during
retrieval F-45
F-52 Components of radiological health impacts to subsurface workers during retrieval for
the low thermal load scenario F-46
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LIST OF FIGURES
Figure
F-1 Sources of radiation exposure.
Page
...F-5
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APPENDIX F. HUMAN HEALTH IMPACTS PRIMER AND DETAILS FOR
ESTIMATING HEALTH IMPACTS TO WORKERS FROM YUCCA
MOUNTAIN REPOSITORY OPERATIONS
Section F. 1 of this appendix contains information that supports the estimates of human health and safety
impacts in this environmental impact statement (EIS). Specifically, Section F. 1 is a primer that explains
the natures of radiation and toxic materials, where radiation comes from in the context of the radiological
impacts discussed in this EIS, how radiation interacts with the human body to produce health impacts,
and how toxic materials interact with the body to produce health impacts. The remainder of the appendix
discusses the methodology that was used to estimate worker health impacts and the input data to the
analysis, and presents the detailed results of the analysis of worker health impacts.
Section F.2 discusses the methodology and data that the U.S. Department of Energy (DOE) used to
estimate worker health and safety impacts for the Proposed Action. It also discusses the detailed results
of the impact analysis.
Section F.3 discusses the methodologies and data that DOE used to estimate worker health and safety
impacts for Inventory Modules 1 and 2. It also discusses the detailed results of the impact analysis.
Section F.4 discusses the methodology and data that DOE used to estimate worker health and safety
impacts for retrieval, should such action become necessary. In addition, it discusses the detailed results
from the impact analysis.
Radiological impacts to the public from operations at the Yucca Mountain site could result from release
of naturally occurring radon-222 and its decay products in the ventilation exhaust from the subsurface
repository operations. The methodology and input data used in the estimates of radiological dose to the
public are presented in Appendix G, Air Quality. Outside of the radiation primer, health impacts to the
public are not treated in this appendix.
F.1 Human Health Impacts
from Exposure to Radioactive
and Toxic Materials
This section introduces the concepts of human
health impacts as a result of exposure to
radiation and potentially toxic materials.
F.1.1 RADIATION AND HUMAN
HEALTH
F.1 .1.1 Radiation
Radiation is the emission and propagation of
energy through space or through a material
in the form of waves or bundles of energy
called photons, or in the form of high-energy
subatomic particles. Radiation generally
results from atomic or subatomic processes
that occur naturally. The most common kind
of radiation is electromagnetic radiation.
RADIATION
Radiation occurs on Earth in many forms, either
naturally or as the result of human activities.
Natural forms include light, heat from the sun,
and the decay of unstable radioactive elements in
the Earth and the environment. Some elements
that exist naturally in the human body are
radioactive and emit ionizing radiation. They
include an isotope of potassium that is an
essential element for health and the elements of
the uranium and thorium naturally occurring
decay series. Human activities have also led to
sources of ionizing radiation for various uses,
such as diagnostic and therapeutic medicine and
nondestructive testing of pipes and welds.
Nuclear power generation produces ionizing
radiation as well as radioactive materials, which
undergo radioactive decay and can continue to
emit ionizing radiation for long periods of time.
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which is transmitted as photons. Electromagnetic radiation is emitted over a range of wavelengths and
energies. We are most commonly aware of visible light, which is part of the spectrum of electromagnetic
radiation. Radiation of longer wavelengths and lower energy includes infrared radiation, which heats
material when the material and the radiation interact, and radio waves. Electromagnetic radiation of
shorter wavelengths and higher energy (which are more penetrating) includes ultraviolet radiation, which
causes sunburn. X-rays, and gamma radiation.
Ionizing radiation is radiation that has sufficient energy to displace electrons from atoms or molecules to
create ions. It can be electromagnetic (for example. X-rays or gamma radiation) or subatomic particles
(for example, alpha and beta radiation). The ions have the ability to interact with other atoms or
molecules; in biological systems, this interaction can cause damage in the tissue or organism.
F.1.1.2 Radioactivity, Ionizing Radiation, Radioactive Decay, and Fission
Radioactivity is the property or characteristic of an unstable atom to undergo spontaneous transformation
(to disintegrate or decay) with the emission of energy as radiation. Usually the emitted radiation is
ionizing radiation. The result of the process, called radioactive decay, is the transformation of an
unstable atom (a radionuclide) into a different atom, accompanied by the release of energy (as radiation)
as the atom reaches a more stable, lower energy configuration.
Radioactive decay produces three main types of ionizing radiation — alpha particles, beta particles, and
gamma or X-rays — but our senses cannot detect them. These types of ionizing radiation can have
different characteristics and levels of energy and, thus, varying abilities to penetrate and interact with
atoms in the human body. Because each type has different characteristics, each requires different
amounts of material to stop (shield) the radiation. Alpha particles are the least penetrating and can be
stopped by a thin layer of material such as a single sheet of paper. However, if radioactive atoms (called
radionuclides) emit alpha particles in the body when they decay, there is a concentrated deposition of
energy near the point where the radioactive decay occurs. Shielding for beta particles requires thicker
layers of material such as several reams of paper or several inches of wood or water. Shielding irom
gamma rays, which are highly penetrating, requires very thick material such as several inches to several
feet of heavy material (for example, concrete or lead). Deposition of the energy by gamma rays is
dispersed across the body in contrast to the local energy deposition by an alpha particle. In fact, some
gamma radiation will pass through the body without interacting with it.
FISSION
Fission is the process whereby a large nucleus
(for example, uranium-235) absorbs a neutron,
becomes unstable, and splits into two fragments,
resulting in the release of large amounts of
energy per unit of mass. Each fission releases an
average of two or three neutrons that can go on to
produce fissions in nearby nuclei. If one or more
of the released neutrons on the average causes
additional fissions, the process keeps repeating.
The result is a self-sustaining chain reaction and a
condition called criticality. When the energy
released in fission is controlled (as in a nuclear
reactor), it can be used for various benefits such
as to propel submarines or to provide electricity
that can light and heat homes.
In a nuclear reactor, heavy atoms such as
uranium and plutonium can undergo another
process, caWedfission, after the absorption of a
subatomic particle (usually a neutron). In
fission, a heavy atom splits into two lighter
atoms and releases energy in the form of
radiation and the kinetic energy of the two
new lighter atoms. The new lighter atoms are
called fission products. The fission products
are usually unstable and undergo radioactive
decay to reach a more stable state.
Some of the heavy atoms might not fission
after absorbing a subatomic particle. Rather, a
new nucleus is formed that tends to be
unstable (like fission products) and undergo
radioactive decay.
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The radioactive decay of fission products and unstable heavy atoms is the source of the radiation from
spent nuclear fuel and high-level radioactive waste that makes these materials hazardous in terms of
potential human health impacts.
F.1.1.3 Exposure to Radiation and Radiation Dose
Radiation that originates outside an individual's body is called external or direct radiation. Such
radiation can come from an X-ray machine or from radioactive materials (materials or substances that
contain radionuclides), such as radioactive waste or radionuclides in soil. Internal radiation originates
inside a person's body following intake of radioactive material or radionuclides through ingestion or
inhalation. Once in the body, the fate of a radioactive material is determined by its chemical behavior and
how it is metabolized. If the material is soluble, it might be dissolved in bodily fluids and be transported
to and deposited in various body organs; if it is insoluble, it might move rapidly through the
gastrointestinal tract or be deposited in the lungs.
Exposure to ionizing radiation is expressed in terms of absorbed dose, which is the amount of energy
imparted to matter per unit mass. Often simply called dose, it is a fundamental concept in measuring and
quantifying the effects of exposure to radiation. The unit of absorbed dose is the rod. The different types
of radiation mentioned above have different effects in damaging the cells of biological systems. Dose
equivalent is a concept that considers (1) the absorbed dose and (2) the relative effectiveness of the type
of ionizing radiation in damaging biological systems, using a radiation-specific quality factor. The unit of
dose equivalent is the rem. In quantifying the effects of radiation on humans, other types of concepts are
also used. The concept of effective dose equivalent is used to quantify effects of radionuclides in the
body. It involves estimating the susceptibility of the different tissue in the body to radiation to produce a
tissue-specific weighting factor. The weighting factor is based on the susceptibility of that tissue to
cancer. The sum of the products of each affected tissue's estimated dose equivalent multiplied by its
specific weighting factor is the effective dose equivalent. The potential effects from a one-time ingestion
or inhalation of radioactive material are calculated over a period of 50 years to account for radionuclides
that have long half-lives and long residence time in the body. The result is called the committed effective
dose equivalent. The unit of effective dose equivalent is also the rem. Total effective dose equivalent is
the sum of the committed effective dose equivalent from radionuclides in the body plus the dose
equivalent from radiation sources external to the body (also in rem). All estimates of dose presented in
this environmental impact statement, unless specifically noted as something else, are total effective dose
equivalents, which are quantified in terms of rem or millirem (which is one one-thousandth of a rem).
More detailed information on the concepts of radiation dose and dose equivalent are presented in
publications of the National Council on Radiation Protection and Measurements (NCRP 1993, page
16-25) and the International Commission on Radiological Protection (ICRP 1991, page 4-1 1). The DOE
implementation guide for occupational exposure assessment (DOE 1998a, pages 3 to 11) also provides
additional information.
The factors used to convert estimates of radionuclide intake (by inhalation or ingestion) to dose are called
dose conversion factors. The National Council on Radiation Protection and Measurements and Federal
agencies such as the U.S. Environmental Protection Agency publish these factors (NCRP 1996, all;
Eckerman and Ryman 1993, all; Eckerman, Wolbarst, and Richardson 1988, all). They are based on
original recommendations of the International Commission on Radiological Protection (ICRP 1977, all).
The radiation dose to an individual or to a group of people can be expressed as the total dose received or
as a dose rate, which is dose per unit time (usually an hour or a year).
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Human Health Impacts Primer and Details for Estimating Health
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Collective dose is the total dose to an exposed population. Person-rem is the unit of collective dose.
Collective dose is calculated by summing the individual dose to each member of a population. For
example, if 100 workers each received 0.1 rem, then the collective dose would be 10 person-rem
(100x0.1 rem).
Exposures to radiation or radionuclides are often characterized as being acute or chronic. Acute
exposures occur over a short period of time, typically 24 hours or less. Chronic exposures occur over
longer times (months to years); they are usually assumed to be continuous over a period, even though the
dose rate might vary. For a given dose of radiation, chronic radiation exposure is usually less harmful
than acute exposure because the dose rate (dose per unit time, such as rem per hour) is lower, providing
more opportunity for the body to repair damaged cells.
F.1.1.4 Background Radiation from Natural Sources
Nationwide, on average, members of the public are exposed to approximately 360 millirem per year from
natural and manmade sources (Gotchy 1987, page 53). Figure F-1 shows the relative contributions by
radiation sources to people living in the United States (Gotchy 1987, page 55).
The estimated average annual dose rate from natural sources is only about 300 millirem per year. This
represents about 80 percent of the annual dose received by an average member of the U.S. public. The
largest natural sources are radon-222 and its radioactive decay products in homes and buildings, which
contribute about 200 millirem per year. Additional natural sources include radioactive material in the
Earth (primarily the uranium and thorium decay series, and potassium-40) and cosmic rays from space
filtered through the atmosphere. With respect to exposures resulting from human activities, medical
exposure accounts for 15 percent of the annual dose, and the combined doses from weapons testing
fallout, consumer and industrial products, and air travel (cosmic radiation) account for the remaining
3 percent of the total annual dose. Nuclear fuel cycle facilities contribute less than 0.1 percent (0.(X)5
millirem per year per person) of the total dose (Gotchy 1987, pages 53 to 55).
F.1.1.5 Impacts to Human Health from Exposure to Radiation
Chronic Exposure
Cancer is the principal potential risk to human health from exposure to low or chronic levels of radiation.
This EIS expresses radiological health impacts as the incremental changes in the number of expected fatal
cancers (latent cancer fatalities) for populations and as the incremental increases in lifetime probabilities
of contracting a fatal cancer for an individual. The estimates are based on the dose received and on dose-
to-health effect conversion factors recommended by the International Commission on Radiological
Protection (ICRP 1991, page 22). The Commission estimated that, for the general population, a collective
dose of 1 person-rem will yield 0.0005 excess latent cancer fatality. For radiation workers, a collective
dose of 1 person-rem will yield an estimated 0.0004 excess latent cancer fatality. The higher risk factor
for the general population is primarily due to the inclusion of children in the population group, while the
radiation worker population includes only people older than 18. These risk coefficients were adopted by
the National Council on Radiation Protection and Measurements in 1993 (NCRP 1993, page 3).
Other health effects such as nonfatal cancers and genetic effects can occur as a result of chronic exposure
to radiation. Inclusion of the incidence of nonfatal cancers and severe genetic effects from radiation
exposure increases the total change by a factor of 1.5 to 5, compared to the change for latent cancer
fatalities (ICRP 1991, page 22). As is the general practice for any DOE EIS, estimates of the total change
were not included in the Yucca Mountain EIS.
F-4
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
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Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Acute Exposure
Exposures to high levels of radiation at high dose rates over a short period (less than 24 hours) can result
in acute radiation effects. Minor changes in blood characteristics might be noted at doses in the range of
25 to 50 rad. The external symptoms of radiation sickness begin to appear following acute exposures of
about 50 to 100 rad and can include anorexia, nausea, and vomiting. More severe symptoms occur at
higher doses and can include death at doses higher than 200 to 300 rad of total body irradiation,
depending on the level of medical treatment received. Information on the effects of acute exposures on
humans was obtained from studies of the survivors of the Hiroshima and Nagasaki bombings and from
studies following a multitude of acute accidental exposures (Mettler and Upton 1995, pages 276 to 280).
Factors to relate the level of acute exposure to health effects exist but are not applied in this EIS because
expected exposures during normal operations for the Proposed Action (including transportation), and for
accident scenarios during the Proposed Action and the associated transportation activities, would be well
below 50 rem. See Appendix J for exposures from accident scenarios during transportation activities.
F.1.1.6 Exposures from Naturally Occurring Radionuclides in the Subsurface
Environment
The estimates of worker doses from inhalation of radon-222 and its decay products while in the
subsurface environment and from the ambient radiation fields in the subsurface environment were based
on measurements taken in the existing Exploratory Studies Facility drifts. The measurements and the
annual dose rates derived from them are discussed below.
Annual Dose Rate for Subsurface Facility Worker from Inhalation of Radon-222
The annual dose rate for a subsurface worker from inhalation of radon-222 and radon decay products was
estimated using site-specific measurements of the concentrations of radon-222 and its decay products in
the Yucca Mountain Exploratory Studies Facility drifts. Measurements were made at a number of
locations in the drifts (TRW 1999a, page 12). After examination of the data from various locations, the
measurements taken at the 5,035-meter (about 16,5(X)-foot) station in the main drift, with the ventilation
system operating, were determined to provide the best basis for estimating the concentration of radon-222
in the subsurface atmosphere during the various Yucca Mountain Repository phases (TRW 1999a, page
12). The measured concentrations ranged from 0.22 to 72 picocuries per liter, with a median value of 6.5
picocuries per liter.
For each project phase, the measured average value (6.5 picocuries per liter) was adjusted to take into
account the difference between the average air residence time in the repository at the time of
measurement of radon-222 concentration and the average air residence time for a specific project phase.
The average air residence time is the average volume being ventilated divided by the average ventilation
rate for a project phase. For example, an increased repository volume would result in an increased
average residence time as would a decrease in the ventilation flow rate.
Also considered were (1) the distribution of the measured values of the equilibrium fraction between
radon-222 and the decay products in the underground facility; this value ranged from 0.0022 to 0.44, with
a median of 0.14 (TRW 1999a, page 12); and (2) the number of hours an involved worker would be
underground, exposed to airborne radon. Based on a typical amount of time spent underground (about
6.5 hours per workday) (Jessen 1999, all), the yearly exposure time for involved workers would range
from 1,500 to 1,7(X) hours per year. The dose conversion factor for radon was taken from Publication 65
of the International Commission on Radiological Protection (ICRP 1994, page 24). This dose conversion
factor, which is 0.5 rem per working-level month for inhalation of radon decay products by workers,
corresponds to 0.029 millirem per picocurie per liter per hour for radon decay products in 100-percent
equilibrium (equilibrium factor of 1.0) with the radon-222 parent (ICRP 1994, page 5). For radon
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Human Health Impacts Primer and Details for Estimating Health
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products with a 0.14 equilibrium factor, the dose conversion factor would be 0.0041 miUirem per
picocurie per liter per hour.
The estimated baseline median dose to an involved worker in the Exploratory Studies Facility from
inhalation of radon and radon decay products was estimated to be approximately 60 millirem per year.
This estimate was used in calculating the worker dose estimates in this appendix. The estimated
5th-percentile dose is 2 millirem per year, and the 95th-percentile dose is 580 millirem per year. These
estimates were made using a Monte Carlo uncertainty analysis.
Annual Dose for Subsurface Facility Worker from Ambient External Radiation in Drifts
Workers in the underground facility would also be exposed to external radiation from naturally occurring
primordial radionuclides in the rock. Measured exposure rates for the underground facility ranged from
0.014 to 0.038 millirem per hour (TRW 1999a, page 12). As for inhalation dose estimates, an
underground exposure time of 1,500 to 1,700 hours per year was considered. The estimated baseline
median dose to an involved worker in the Exploratory Studies Facility from ambient external radiation
would be approximately 40 millirem per year. This estimate was used in this appendix for calculating the
worker dose estimates from ambient external radiation. The estimated 5th-percentile dose is 23 millirem
per year, and the 95th-percentile dose is 56 millirem per year. Like the radon dose estimates, these
estimates were made using a Monte Carlo uncertainty analysis.
F.I .2 EXPOSURE TO TOXIC OR HAZARDOUS MATERIALS
When certain natural or manmade materials or substances have harmful effects that are not random or do
not occur solely at the site of contact, the materials or substances are described as toxic. Toxicology is
the branch of science dealing with the toxic effects that chemicals or other substances might have on
living organisms.
Chemicals can be toxic for many reasons, including their ability to cause cancer, to harm or destroy tissue
or organs, or to harm body systems such as the reproductive, immune, blood-forming, or nervous
systems. The following list provides examples of substances that can be toxic:
• Carcinogens, which are substances known to cause cancer in humans or in animals. If cancers have
been observed in animals, they could occur in humans. Examples of generally accepted human
carcinogens include asbestos, benzene, and vinyl chloride (Kamrin 1988, pages 37 and 38 and
Chapter 6).
• Chemicals that controlled studies have shown to cause a harmful or fatal effect. Examples include
metals such as cadmium, lead, and mercury; strong acids such as nitric acid and sulfuric acid; some
welding fumes; coal dust; sulfur dioxide; and some solvents.
• Some biological materials, including various body fluids and tissues and infectious agents, are toxic.
Even though chemicals might be toxic, many factors influence whether or not a particular substance has a
toxic effect on humans. These factors include (1) the amount of the substance with which the person
comes in contact, (2) whether the person inhales or ingests a relatively large amount of the substance in a
short time (acute exposure) or repeatedly ingests or inhales a relatively small amount over a longer time
(chronic exposure), and (3) the period of time over which the exposure occurs.
Scientists determine a substance's toxic effect (or toxicity) by performing controlled tests on animals. In
addition to environmental and physical factors, these tests help establish three other important factors for
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measuring toxicity — dose-response relationship, threshold concept, and margin of safety. The dose-
response relationship relates the percentage of test animals that experience observable toxic effects to the
doses administered. After the administration of an initial dose, the dose is increased or decreased until, at
the upper end, all animals are affected and, at the lower end, no animals are affected. Thus, there is a
threshold concentration below which there is no effect. The margin of safety is an arbitrary separation
between the highest concentration or exposure level that produces no adverse effect in a test animal
species and the concentration or exposure level designated safe for humans. There is no universal margin
of safety. For some chemicals, a small margin of safety is sufficient; others require a larger margin.
Two substances in the rock at Yucca Mountain, crystalline silica and erionite, are of potential concern as
toxic or hazardous materials. Both of these naturally occurring compounds occur in the parent rock at the
repository site, and excavation activities could encounter them. The following paragraphs contain
additional information on these.
Crystalline Silica
Crystalline silica is a naturally occurring, highly structured form of silica (silicon dioxide, Si02). Because
it can occur in several different forms, including quartz, cristobalite, and tridymite, it is called a
polymorph. These three forms occur in the welded tuff parent rock at Yucca Mountain (DOE 1998b,
page 25). Crystalline silica is a known causative agent for silicosis, a destructive lung condition caused
by deposition of particulate matter in the lungs and characterized by scarring of lung tissue. It is
contracted by prolonged exposure to high levels of respirable silica dust or an acute exposure to even
higher levels of respirable silica dust (EPA 1996, Chapter 8). Accordingly, DOE considers worker
inhalation of respirable crystalline silica dust particles to be hazardous to worker health. Current
standards for crystalline silica have been established to prevent silicosis in workers.
Cristobalite has a lower exposure limit than does quartz. The limits for these forms of silica include the
Permissible Exposure Limits established by the Occupational Safety and Health Administration and the
Threshold Limit Value defined by the American Conference of Governmental Industrial Hygienists. The
Occupational Safety and Health Administration Permissible Exposure Limit is 50 micrograms per cubic
meter averaged over a 10-hour work shift. The American Conference of Governmental Industrial
Hygienists Threshold Limit Value is also 50 micrograms per cubic meter, but it is averaged over an
8-hour work shift (NJDHSS 1996, all). Thus, the two limits are essentially the same. In accordance with
DOE Order 440.1 A (DOE 1998a, page 5), the more restrictive value provided by the American
Conference of Governmental Industrial Hygienists will be applied. In addition, the National Institute for
Occupational Safety and Health has established Immediately-Dangerous-to-Life-and-Health
concentration limits at levels of 50,000 and 25,0(X) micrograms per cubic meter for quartz and
cristobalite, respectively (NIOSH 1996, page 2). These limits are based on the maximum airborne
concentrations an individual could tolerate for 30 minutes without suffering symptoms that could impair
escape from the contaminated area or irreversible acute health effects.
There is also evidence that silica may be a carcinogen. The International Agency for Research on Cancer
has classified crystalline silica and cristobalite as a Class I (known) carcinogen (LARC 1997, pages 205 to
210). The National Institute for Occupational Safety and Health considers crystalline silica to be a
potential carcinogen, as defined by the Occupational Safety and Health Administration's carcinogen
policy (29 CFR Part 1990). The National Institute for Occupational Safety and Health is reviewing data
on carcinogenicity, which could result in a revised limit for crystalline silica. The Environmental
Protection Agency has noted an increase in cancer risk to humans who have already developed the
adverse noncancer effects of silicosis, but the cancer risk to otherwise healthy individuals is not clear
(EPA 1996, pages 1 to 5).
F-8
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Because there are no specific limits for exposure of members of the public to crystalline silica, this
analysis used a comparative benchmark of 10 micrograms per cubic meter, based on a cumulative
lifetime exposure limit of 1,000 micrograms per (cubic meter multiplied by years). At this level, an
Environmental Protection Agency health assessment has stated that there is a less than 1 percent chance
of silicosis (EPA 1996, Chapter 1, page 5, and Chapter 7, page 5). Over a 70-year lifetime, this
cumulative exposure benchmark would correspond to an annual average exposure concentration of about
14 micrograms per cubic meter, which was rounded down to 10 micrograms per cubic meter to establish
the benchmark. Appendix G, Section G.l contains additional information on public exposure to
crystalline silica.
Samples of the welded tuff parent rock from four boreholes at Yucca Mountain have an average quartz
content of 15.7 percent, an average cristobalite content of 16.3 percent, and an average tridymite content
of 3.5 percent (DOE 1998b, page I-l). Worker protection during excavation in the subsurface would be
based on the more restrictive Threshold Limit Value for cristobalite. The analysis assumed that the parent
rock and dust would have a cristobalite content of 28 percent, which is the higher end of the concentration
range reported in TRW (1999b, page 4-81). Thus, the assumed percentage of cristobalite in dust probably
will overestimate the airborne cristobalite concentration. Also, studies of both ambient and occupational
airborne crystalline silica have shown that most of the airborne crystalline silica is coarse and not
respirable (greater than 5 micrometers aerodynamic diameter), and the larger particles will deposit rapidly
on the surface (EPA 1996, page 3-26).
Erionite
Erionite is a natural fibrous zeolite that occurs in the rock layers below the proposed repository level in
the hollows of rhyolitic tuffs and in basalts. It might also occur in rock layers above the repository level
but has not been found in those layers. Erionite is a rare tectosilicate zeolite with hexagonal symmetry
that forms wool-like fibrous masses (with a maximum fiber length of about 50 microns, which is
generally shorter than asbestos fibers). Erionite particles (ground to powder) resemble amphibole
asbestos fibers. Erionite fibers have been detected in samples of road dust in Nevada, and residents of the
Intermountain West could be exposed to fibrous erionite in ambient air (Technical Resources 1994,
page 134).
There are no specific limits for exposure to erionite. Descriptive studies have shown very high mortality
from cancer [malignant mesothelioma, mainly of the pleura (a lung membrane)] in the population of three
Turkish villages in Cappadocia where erionite is mined. The International Agency for Research on
Cancer has indicated that these studies demonstrate the carcinogenicity of erionite to humans. The
Agency classifies erionite as a Group 1 (known) carcinogen (LARC 1987, all).
Erionite could become a potential hazard during excavation of access tunnels to the lower block and to
offset Area 5 for the low and intermediate thermal load cases or during vertical boring operations
necessary to excavate ventilation shafts. DOE does not expect to encounter erionite layers during the
vertical boring operations, which would be through rock layers above known erionite layers, or during
excavation of access tunnels to the lower block or offset Area 5, where any identified layers of erionite
would likely be avoided (McKenzie 1998, all). In accordance with the Erionite Protocol (DOE 1995, all),
a task-specific health and safety plan would be prepared before the start of boring operations to identify
this material and prevent worker inhalation exposures from unconfmed material.
The Los Alamos National Laboratory is studying the mineralogy and geochemistry of the deposition of
erionite under authorization from the DOE Office of Energy Research. Laboratory researchers are
applying geochemical modeling so they can understand the factors responsible for the formation of zeolite
assemblages in volcanic tuffs. The results of this modeling will be used to predict the distribution of
F-9
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
erionite at Yucca Mountain and to assist in the planning of excavation operations so erionite layers are
avoided.
F.I .3 EXPOSURE PATHWAYS
Four conditions must exist for there to be a pathway from the source of released radiological or toxic
material to a person or population (Maheras and Thome 1993, page 1):
•
A source term: The material released to the environment, including the amount of radioactivity (if
any) or mass of material, the physical form (solid, liquid, gas), particle size distribution, and chemical
form
• An environmental transport medium: Air, surface water, groundwater, or a food chain
• An exposure route: The method by which a person can come in contact with the material (for
example, external exposure from contaminated ground, inmiersion in contaminated air or internal
exposure from inhalation or ingestion of radioactive or toxic material)
• A human receptor: The person or persons potentially exposed; the level of exposure depends on such
factors as location, duration of exposure, time spent outdoors, and dietary intake
These four elements define an exposure pathway. For example, one exposure scenario might involve
release of contaminated gas from a stack (source term); transport via the airborne pathway (transport
medium); external gamma exposure from the passing cloud (exposure route); and an onsite worker
(human receptor). Another exposure scenario might involve a volatile organic compound as the source
term, release to groundwater as the transport medium, ingestion of contaminated drinking water as the
exposure route, and offsite members of the public as the human receptors. No matter which pathway the
scenario involves, local factors such as water sources, agriculture, and weather patterns play roles in
determining the importance of the pathway when assessing potential human health effects.
Worker exposure to crystalline silica (and possibly erionite) in the subsurface could occur from a rather
unique exposure pathway. Mechanical drift excavation, shaft boring, and broken rock management
activities could create airborne dust comprising a range of particles sizes. Dust particles smaller than
10 micrometers have little mass and inertia in comparison to their surface area; therefore, these small
particles could remain suspended in dry air for long periods. Airborne dust concentrations could increase
if the ventilation system recirculated the air or if airflow velocity in the subsurface facilities became high
enough to entrain dust previously deposited on drift or equipment surfaces. As tunnel boring machines or
road headers break the rock from the working face, water would be applied to wet both the working face
and the broken rock to minimize airborne dust levels. Wet or dry dust scrubbers would capture dust that
was not suppressed by the water sprays. To prevent air recirculation, which would lead to an increase of
airborne dust loads, the fresh air intake and the exhaust air streams would be separated. Finally, the
subsurface ventilation system would be designed and operated to control ambient air velocities to
minimize dust reentrainment. If these engineering controls did not maintain dust concentrations below
the Threshold Limit Value concentration, workers would have to wear respirators until engineering
controls established habitable conditions.
F-10
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
F.2 Human Health and Safety Impact Analysis
for the Proposed Action Inventory
This section discusses the methodologies and data used to estimate industrial and radiological health and
safety impacts to workers that would result from the construction, operation and monitoring, and closure
of the Yucca Mountain Repository, as well as the detailed results from the impact calculations. Section
F.2. 1 describes the methods used to estimate impacts, Section F.2.2 contains tabulations of the detailed
data used in the impact calculations and references to the data sources, and Section F.2.3 contains a
detailed tabulation of results.
For members of the public, the EIS uses the analysis methods in Appendix K, Section K.2, to estimate
radiation dose from radon-222 and crystalline silica released in the subsurface ventilation system exhaust.
The radiation dose estimates were converted to estimates of human health impacts using the dose
conversion factors discussed in Section F.1.1.5. These impacts are expressed as the probability of a latent
cancer fatality for a maximally exposed individual and as the number of latent cancer fatalities among
members of the public within about 80 kilometers (50 miles) for the Proposed Action, the retrieval
contingency, and the inventory modules. The results are listed in Chapter 4, Section 4.1.7.
Health and safety impacts to workers have been estimated for two worker groups: involved workers and
noninvolved workers, hivolved workers are craft and operations personnel who would be directly involved
in activities related to facility construction and operations, including excavation activities; receipt, handling,
packaging, and emplacement of spent nuclear fuel and high-level radioactive waste material; monitoring of
conditions and performance of the waste packages; and those directly involved in closure activities.
Noninvolved workers are managerial, technical, supervisory, and administrative personnel who would not be
directly involved in construction, excavation, operations, monitoring, and closure activities. The analysis did
not consider project workers who would not be located at the repository site.
F.2.1 METHODOLOGY FOR CALCULATING OCCUPATIONAL HEALTH AND SAFETY
IMPACTS
To estimate the impacts to workers from industrial hazards common to the workplace, values for the
full-time equivalent work years for each phase of the project were multiplied by the statistic (occurrence
per 10,000 full-time equivalent work years) for the impact being considered. Values for the number of
full-time equivalent workers for each phase of the project are listed in Section F.2.2. 1. The statistics for
industrial impacts for each of the phases are listed in Section ¥.1.1.2 for involved and noninvolved
workers.
Two kinds of radiological health impacts to workers are provided in this EIS. The first is an estimate of
the latent cancer fatalities to the worker group involved in a particular project phase. The second is the
incremental increase in latent cancer fatalities attributable to occupational radiation for a maximally
exposed individual in the worker population for each project phase.
To calculate the expected number of worker latent cancer fatalities during a phase of the project, the
collective dose to the worker group, in person-rem, was multiplied by a standard factor for converting the
collective worker dose to projected latent cancer fatalities (see Section F.1.1.5). As discussed in
Section F.1.1.5, the value of this factor for radiation workers is 0.(X)04 excess latent cancer fatality per
person-rem of dose.
The collective dose for a particular phase of the operation is calculated as the product of the number of
full-time equivalent workers for the project phase (see Section F.2.2. 1), the average dose over the
exposure period, and the fraction of the working time that a worker is in an environment where there is a
F-11
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
source of radiation exposure. Values for exposure rates for both involved and noninvolved workers are
presented in Section F.2.2.3 as are the fractional occupancy factors. The calculation of collective dose to
subsurface workers from exposure to the radiation emanating from the loaded waste packages is an
exception. Collective worker doses from this source of exposure were calculated using the methodology
described in TRW (1999b, Tables G-1 and G-2). For the calculation of exposures, the estimated annual
radiation doses listed in TRW (1999b, Tables G-3, G-3a, G-4, and G-4a) for the various classes of
involved subsurface workers were used. The exposure values were multiplied by the craft manpower
distribution listed in TRW (1999b, Tables G-5, G-5a, G-5b, G-7, G-7a, and G-7b) for each of the
involved labor classes for a project phase to obtain an overall annual exposure. The annual exposures for
the labor classes were then summed to obtain the collective annual dose in person-rem to the involved
subsurface workers for each of the subsurface operational phases. The total collective dose was then
obtained by multiplying the annual collective dose by the length of the project phase.
To estimate the incremental increase in the likelihood of death from a latent cancer for the maximally
exposed individual, the estimated dose to the maximally exposed worker was multiplied by the factor for
converting radiation dose to latent cancers. The factor applied for workers was 0.0004 latent cancer
fatality per rem, as discussed above and in Section F. 1.1. 5. Thus, if a person were to receive a dose of
1 rem, the incremental increase in the probability that person would suffer a latent cancer fatality is 1 in
2,500 or 0.0004.
To estimate the dose for a hypothetical maximally exposed individual, the analysis generally assumed that
this individual would be exposed to the radiation fields (see Section F.2.2.3) over the entire duration of a
project phase or for 50 years, whichever would be shorter. Other sources of exposure while working
underground would be ambient radiation coming from the radionuclides in the drift walls and from
inhalation of radon-222 and its decay products. The radiation from the waste package is usually the
dominant component when these three dose contributors are added. Doses for the maximally exposed
subsurface worker were estimated by adding the three dose components because they would occur
simultaneously.
F.2.2 DATA SOURCES AND TABULATIONS
F.2.2.1 Work Hours for the Repository Phases
Table F-1 lists the number of workers involved in the various repository phases in terms of full-time
equivalent work years. Each full-time equivalent work year represents 2,(XX) work hours (the number of
hours assumed for a normal work year). The values were obtained from TRW (1999c, Section 6) and
from TRW (1999b, Section 6) for surface and subsurface workers, respectively.
F.2.2.2 Workplace Health and Safety Statistics
The analysis selected health and safety statistics for three impact categories — total recordable cases, lost
workday cases, and fatalities. Total recordable cases are occupational injuries or illnesses that result in:
•
•
•
Fatalities, regardless of the time between the injury and death, or the length of the illness
Lost workday cases, other than fatalities, that result in lost workdays
Nonfatal cases without lost workdays that result in transfer to another job, termination of
employment, medical treatment (other than first aid), loss of consciousness, or restriction of work or
motion
Diagnosed occupational illness cases that are reported to the employer but are not classified as
fatalities or lost workday cases
F-12
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
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F-13
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Lost workday cases, which are described above, include cases that result in the loss of more than half a
workday. These statistical categories, which have been standardized by the U.S. Department of Labor
and the Bureau of Labor Statistics, must be reported annually by employers with II or more employees.
Table F-2 summarizes the health and safety impact statistics used for this analysis.
Table F-2. Health and safety statistics for estimating industrial safety impacts common to the
workplace.'
Total recordable cases
incidents per
lOOFTEs"
Lost workday cases
per 100 FTEs
Fatalities per
100,000 FTEs
(involved and
noninvolved)*^
Data set for
TRCs and
Phase
Involved
Noninvolved
Involved
Noninvolved
LWCs"
Construction
Surface
6.1
3.3
2.9
1.6
2.9
(1)
Subsurface
6.1
3.3
2.9
1.6
2.9
(1)
Operation and Monitoring
Operation period
Surface
3
3.3
1.2
1.6
2.9
(3)
Subsurface - emplacement
Subsurface - drift
3
6.8
3.3
I.l
1.2
4.8
1.6
0.7
2.9
2.9
(3)
(2)
development
Monitoring period
Surface
3
3.3
1.2
1.6
2.9
(3)
Subsurface
3
3.3
1.2
1.6
2.9
(3)
Closure
Surface
6.1
3.3
2.9
1.6
2.9
(1)
Subsurface
6.1
3.3
2.9
1.6
2.9
(1)
a. See text below for source of data in Data Sets 1 , 2, and 3.
b. FTEs = full-time equivalent work years.
c. See the discussion about Data Set 4 for source of fatality statistic for normal industrial activities.
d. TRCs = total recordable cases; LWCs = lost workday cases.
Table F-2 cites three sets of statistics that were used to estimate total recordable cases and lost workday
cases for workers during activities at the Yucca Mountain site. In addition, there is a fourth statistic
related to the occupational fatality projections for the Yucca Mountain site activities. The source of
information from which the sets of impact statistics were derived is discussed below. All of the statistics
are based on DOE experience for similar types of activities and were derived from the DOE CAIRS
(Computerized Accident/Incident Reporting and Recordkeeping System) data base (DOE 1999, all).
Data Set 1, Construction and Construction-Like Activities
This set of statistics from the DOE CAIRS data base was applied to construction or construction-like
activities. Specifically, it was used for both surface and subsurface workers during the construction phase
and the closure phase (closure phase activities were deemed to be construction-like activities). The
statistics were based on a 6.75-year period (1992 through the third quarter of 1998).
For involved workers the impact statistic numbers were derived from the totals for all of the DOE
construction activities over the period. For noninvolved workers, the values were derived from the
combined government and services contractor noninvolved groups for the same period. The noninvolved
worker statistic, then, is representative of impacts for oversight personnel who would not be involved in
F-14
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
the actual operation of equipment or resources. The basic statistics derived from the CAIRS data base for
each of the groups include:
• Involved worker total recordable cases: 764 recordable cases for approximately 12,400 full-time
equivalent work years
• Involved worker lost workday cases: 367 lost workday cases for approximately 12,400 full-time
equivalent work years
• Noninvolved worker total recordable cases: 1,333 recordable cases for approximately 40,600 full-
time equivalent work years
• Noninvolved worker lost workday cases: 657 lost workday cases for approximately 40,600 full-time
equivalent work years
Data Set 2, Excavation Activities
This set of statistics was derived from experience at the Yucca Mountain Project over a 30-month period
(fourth quarter of 1994 though the first quarter of 1997). DOE selected this period because it coincided
with the exploratory tunnel boring machine operations at Yucca Mountain, reflecting a high level of
worker activity during ongoing excavation activities. This statistic was applied for the Yucca Mountain
Project subsurface development period, which principally involves drift development activities. The
Yucca Mountain Project experience from which the statistic is derived is presented in Table F-3. Stewart
(1998, all) contains the Yucca Mountain statistics, which were derived from the CAIRS data base (DOE
1999, all).
Table F-3. Yucca Mountain Project worker industrial safety loss experience.'
Factor
Value"
Basis
TRCs' per 100 FTEs"
Involved worker
Noninvolved worker
LWCs'per 100 FTEs
Involved worker
Noninvolved worker
Fatality rate occurrence per 100,000 FTEs
Involved worker
Noninvolved worker
6.8 56 TRCs for 825 construction FTEs
1.1 2.3 TRCs for 2,015 nonconstruction FTEs
4.8 40 LWCs for 825 construction FTEs
0.7 14 LWCs for 2,015 nonconstruction FTEs
0.0 No fatalities for 825 construction FTEs
0.0 No fatalities for 2,015 nonconstruction FTEs
a. Fourth quarter 1994 through first quarter 1997.
b. Source: Adapted from the CAIRS data base (DOE 1999, all) by Stewart (1998, all) for the fourth quarter of 1994 through
the first quarter of 1997.
c. TRCs = total recordable cases of injury and illness.
d. FTEs = full-time equivalent work years.
e. LWCs = lost workday cases.
Data Set 3, Activities Involving Work in a Radiological Environment
This set of statistics is from the DOE CAIRS data base (DOE 1999, all). In arriving at the statistics listed
in Table F-2, information from the Savannah River Site, the Hanford Site, and the Idaho National
Engineering and Environmental Laboratory was averaged individually for the 6.5 years from 1992
through the second quarter of 1998. The averages were then combined to produce an overall average.
The reason these three sites were selected as the basis for this set of statistics is that the DOE Savannah
River, Hanford, and Idaho National Engineering and Environmental Laboratory sites currently conduct
most of the operations in the DOE complex involving handling, sorting, storing, and inspecting spent
F-15
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
nuclear fuel and high-level radioactive waste materials, as well as similar activities for low-level
radioactive waste materials. The Yucca Mountain Repository phases for which this set of statistics was
applied included the receipt, handling, and packaging of spent nuclear fuel and high-level radioactive
waste in the surface facilities; subsurface emplacement activities; and surface and subsurface monitoring
activities, including decontamination of the surface facilities. These activities involve handling, storing,
and inspecting spent nuclear fuel and high-level radioactive waste, so the worker activities at the Yucca
Mountain site are expected to be similar to those cited above for the other sites in the DOE complex.
The basic statistics for the involved and noninvolved workers include:
• Involved worker total recordable cases: 1,246 for about 41,600 full-time equivalent work years
• Involved worker lost workday cases: 538 for about 41,600 full-time equivalent work years
• Noninvolved worker total recordable cases: 1,333 for about 40,600 full-time equivalent work years
• Noninvolved worker lost workday cases: 657 for about 40,600 full-time equivalent work years
Data Set 4, Statistics for Worlier Fatalities from Industhai Hazards
There have been no reported fatalities as a result of workplace activities for the Yucca Mountain project.
Similarly, there are no fatalities listed in the Mine Safety and Health Administration data base for stone
mining workers (MSHA 1999, all). Because fatalities in industrial operations sometimes occur, the more
extensive overall DOE data base was used to estimate a fatality rate for the activities at the Yucca
Mountain site. Statistics for the DOE facility complex for the 10 years between 1988 and 1997 were used
(DOE 1999, all). These fatality statistics are for both government and contractor personnel working in the
DOE complex who were involved in the operation of equipment and resources (involved workers). The
activities in the DOE complex covered by this statistic were governed by safety and administrative
controls (under the DOE Order System) that are similar to the safety and administrative controls that
would be applied for Yucca Mountain Repository work. These fatality statistics were also applied to the
noninvolved worker population because they are the most inclusive statistics in the CAIRS data base.
However, the statistics probably are conservatively high for the noninvolved worker group.
F.2.2.3 Estimates of Radiological Exposures
DOE considered the following potential sources of radiation exposure for assessing radiological health
impacts to workers:
• Inhalation of gaseous radon-222 and its decay products. Subsurface workers could inhale the
radon-222 present in the air in the repository drifts. Workers on the surface could inhale radon-222
released to the environment in the exhaust air from the subsurface ventilation system.
• External exposure of surface workers to radioactive gaseous fission products that could be released
during handling and packaging of spent nuclear fuel with failed cladding for emplacement in the
repository. Such impacts would be of most concern for the uncanistered shipping cask scenario.
• Direct external exposure of workers in the repository drifts as a result of naturally occurring
radionuclides in the walls of the drifts (primarily potassium-40 and radionuclides of the naturally
occurring uranium and thorium decay series).
• Extemal exposure of workers to direct radiation emanating from the waste packages containing spent
nuclear fuel and high-level radioactive waste either during handling and packaging (surface facility
workers) or after it is placed within the waste package (largely subsurface workers).
F-16
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Section F. 1.1.6 describes the approach taken to estimate exposures to workers as a result of release of
gaseous radon-222 from the drift walls to the subsurface atmosphere. For radon exposures to subsurface
workers, the analysis assumed a subsurface occupancy factor of 1 .0 for involved workers, an occupancy
factor of 0.6 for noninvolved workers for construction and drift development activities, and an occupancy
factor of 0.4 for noninvolved workers for emplacement, monitoring, and closure (Rasmussen 1998a, all;
Rasmussen 1999, all; lessen 1999, all).
As discussed in Section F. 1.1.6, the average concentration of radon-222 in the subsurface atmosphere
varies with the ventilation rate and repository volume. Table F-4 lists the correction factors (multipliers)
applied to the average value for the concentration of radon-222 measured in the Exploratory Studies
Facility for the Proposed Action.
Table F-4. Correction factors and annual exposures from radon-222 and its decay products for each of
the project phases or periods under the Proposed Action."*
Annual dose rate (millirem per year)
Correction factor
Thermal load scenario
Thermal load scenario
Project phase or period
High
Intermediate
Lx)w
High
Intermediate
Low
Construction
1.9
2.2
2.2
114
132
132
Drift development
0.6
0.6
0.6
36
36
36
Emplacement
1.1
1.5
2.9
66
90
174
Monitoring
3.2
4.1
4.4
192
246
264
Closure
3.2
4.1
4.4
192
246
264
Retrieval''
3.2
3.2
3.2
192
192
192
Based on the measured value of 60 rem per year corrected for repository volume and ventilation rate; see Section F. 1 . 1 .6
and Appendix G (Section G.2.3.1).
b. Multiplier for retrieval is not dependent on thermal load.
Appendix G, Section G.2.4.2 describes the approach taken to estimate source terms and associated doses
to workers from the potential release of gaseous fission products from spent nuclear fuel with failed
cladding.
Subsurface workers would also be exposed to background gamma radiation from naturally occurring
radionuclides in the subsurface rock (largely from the uranium-238 decay series radionuclides and from
potassium-40, both in the rock). DOE has based its projection of worker external gamma dose rates on
the data obtained during Exploratory Studies Facility operations (Section F. 1.1.6). The collective ambient
radiation exposures for subsurface workers were calculated assuming occupancy factors cited in the
previous paragraph for subsurface workers for emplacement and monitoring activities (Rasmussen 1998a,
all; Rasmussen 1999, all; lessen 1999, all).
Table F-5 lists dose rates in the fourth column for cases in which the annual full-time equivalent surface
worker exposure values vary with the shipping package scenario. The table also lists the sources from
which the data were obtained. The dose rates to subsurface workers from the radiation emitted from
waste packages would vary with the thermal load, as indicated in the fourth column of Table F-5.
Table F-6 lists the annual exposures to subsurface workers from radiation emanating from the waste
packages for the high, intermediate, and low thermal load scenarios, under the Proposed Action and
Module 1 and 2 inventories. Section F.3 discusses Inventory Modules 1 and 2.
F-17
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-5. Radiological exposure data used to calculate worker radiological health impacts (page 1 of 2).
Phase and worker
group
Exposure source' Occupancy factor
Annual dose
(millirem, except
where noted)
Annual full-time
equivalent workers'
UC DISP' DPC' Data source*
Construction
Surface
Involved
Radon-222 inhalation
1.0
Small relative to
subsurface worker
exposures
(h)
Noninvolved
Radon-222 inhalation
1.0
Small relative to
subsurface worker
exposures
(h)
Subsurface
Involved
Drift ambient
1.0
40
(1).(2)
Radon-222 inhalation
1.0
Table F-4
(2), Table F-4
Noninvolved
Drift ambient
0.6
40
(1), (2)
Radon-222 inhalation
0.6
Table F-4
(2), Table F-4
Operations and
monitoring
Surface handling
and loading
operations
Involved
Receipt, handling and
1.0
400
464
199
199
packaging of spent
100
297
228
244
(3)
nuclear fuel and high-
level radioactive
waste
Noninvolved
Receipt, handling and
1.0
25
175
150
149
(3)
packaging of spent
0
341
386
390
nuclear fuel and high-
level radioactive
waste
Surface monitoring
Involved only
Radon-222 inhalation
1.0
Small relative to
subsurface workers
(i)
Surface
decontamination
(postemplacement.
involved only)
External exposure
I.O
100
826
599
624
(4)
I.O
25
528
383
399
(4)
Subsurface
emplacement
Involved
Waste package
Varies, see Table F-6
Varies, see Table F-6
Table F-6
Drift ambient
1.0
40
(1),(2)
Radon-222
1.0
Table F-4
(2), Table F-4
Noninvolved
Waste package
0.04
0.1 millirem per hour
(5)
Drift ambient
0.4
40
(I), (2)
Radon-222 inhalation
0.4
Table F-4
(2), Table F-4
Subsurface drift
development
Involved
Drift ambient
1.0
40
(1).(2)
Radon-222 inhalation
1.0
Table F-4
(2), Table F-4
Noninvolved
Drift ambient
0.6
40
(1),(2)
Radon-222 inhalation
0.6
Table F-4
(2), Table F-4
Monitoring
Subsurface
Involved
Waste package
Varies, see Table F-6
Varies, see Table F-6
Table F-6
Drift ambient
1.0
40
(1).(2)
Radon-222 inhalation
1.0
Table F-4
(2), Table F-4
Noninvolved
Waste package
0.04
0. 1 millirem per hour
(5)
Drift ambient
0.4
40
(1).(2),(6)
Radon-222 inhalation
0.4
Table F-4
(2). (6),
Table F-4
F-18
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-5. Radiological exposure data used to calculate worker radiologica
health impacts (page 2 of 2).
Phase and worker
group
Exposure source'
Occupancy factor''
Annual dose
(millirem per year _
except where noted)
Annual full-time
equivalent workers"
UC DISP' DPC'
Data source*
Closure
Surface
Involved
1.0
Small relative to
subsurface worker
0)
Noninvolved
1.0
exposures
Small relative to
subsurface worker
0)
Subsurface
Involved
Noninvolved
Waste package
Drift ambient
Radon-222 inhalation
Waste package
Drift ambient
Radon-22 inhalation
Varies, see Table F-6
1.0
1.0
0.04
0.4
0.4
exposures
Varies, see Table F-6
40
Table F-4
0. 1 millirem per hour
40
Table F-4
Table F-6
(1).(2)
(2), Table F-4
(5)
(1).(2)
(2), Table F-4
b.
d.
e.
f.
Exposure sources include radiation from spent nuclear fuel and high-level radioactive waste packages to surface and subsurface
workers, the ambient exposure to subsurface workers from naturally occurring radiation in the drift walls, and internal exposures
from inhalation of radon-222 and its decay products in the drift atmosphere.
Fraction of 8-hour workday that workers are exposed.
Number of annual full-time equivalent workers for surface facility activities when number of workers would vary with shipping
package scenario.
UC = uncanistered packaging scenario.
DISP = disposable canister packaging scenario.
DPC = dual-purpose canister packaging scenario.
Sources:
(1) Section F.I. 1.6.
(2) Rasmussen (1998a, all).
TRW (1999c, Table 6-2).
Total employment for decontamination activities taken from TRW (1999c, Table 6-4). In Table 6-2 of TRW (1999c), the
distribution of involved workers for surface facility receipt, handling, and packaging phase between the 400 millirem per year
and 100 millirem per year cases is 61 percent and 39 percent, respectively. For decontamination operations it was assumed
that 69 percent of the involved worker population would receive 100 millirem per year and 39 percent of the involved worker
population would receive 25 millirem per year.
Rasmussen (1999, all).
Jessen (1999, all).
Comparison of information in Chapter 4, Table 4-2 (surface workers) and Table F-9 (subsurface workers).
Comparison of information in Chapter 4, Table 4-5 (surface workers) and Table F-27 (subsurface workers).
Comparison of information in Chapter 4, Table 4-7 (surface workers) and Table F-30 (subsurface workers).
(3)
(4)
(5)
(6)
Table F-6. Annual involved subsurface worker exposure rates from waste packages" (person-rem per
year).
Proposed Action
Inventory Modules
Project phase
High
Intermediate
Low
High
Intermediate
Low
Emplacement
Monitoring
Closure
10.1
7.2
12.5
10.2
7.2
12.5
5.6
4.1
7.4
10.2
7.2
12.5
10.2
7.8
12.5
6.0
5.6
7.4
a. Sources: individual exposure values from TRW (1999b, Appendix G, Tables G-3, G-3a, G-4, and G-4a).
b. Calculated annual exp)osures, Rasmussen (1999, all).
F.2.3 COMPILATION OF DETAILED RESULTS FOR OCCUPATIONAL HEALTH AND
SAFETY IMPACTS
F.2.3.1 Occupational Health and Safety Impacts During the Construction Phase
F.2.3.1.1 Industrial Hazards to Workers
Tables F-7 and F-8 list health and safety impacts from industrial hazards to surface and subsurface
workers, respectively, for construction activities.
F-19
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-7. Industrial hazard health and safety impacts to surface facility workers during construction
phase (44 months)."
Waste packaging scenario
Worker group
Uncanistered
Disposable canister Dual-purpose canister
Involved
Full-time equivalent work years'"
Total recordable cases
Lost workday cases
Fatalities
Noninvolved
Full-time equivalent work years
Total recordable cases
Lost workday cases
Fatalities
All workers (totals f
Full-time equivalent work years
Total recordable cases
Lost workday cases
Fatalities
2,380
150
70
0.07
900
30
15
0.03
3,280
180
85
0.10
1,650
100
50
0.05
630
21
10
0.02
2,280
120
59
0.07
1,760
110
50
0.05
670
22
11
0.02
2,420
130
63
0.07
a. Source: Impact rates from Table F-2.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
Table F-8. Industrial hazard health and safety impacts to subsurface facility workers during construction
phase (5 years).^
Thermal load scenario
Worker group
High
Intermediate
Low
Involved
Full-time equivalent work years""
2,300
2,460
2,460
Total recordable cases
140
150
150
Lost workday cases
68
72
72
Fatalities
0.07
0.07
0.07
Noninvolved
Full-time equivalent work years
600
600
600
Total recordable cases
20
20
20
Lost workday cases
10
10
10
Fatalities
0.02
0.02
0.02
All workers (totals f
Full-time equivalent work years
2,900
3,060
3,060
Total recordable cases
160
170
170
Lost workday cases
77
82
82
Fatalities
0.08
0.09
0.09
a. Source: Impact rates from Table F-2.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
F.2.3.1 .2 Radiological Health Impacts to Workers
Tables F-9 and F-IO list subsurface worker health impacts from inhalation of radon-222 in the subsurface
atmosphere and from ambient radiation exposure from radionuclides in the rock of the drift walls,
respectively. The radiological health impacts to surface workers from inhalation of radon-222 would be
small in comparison to those for subsurface workers; therefore, they were not tabulated in this appendix
(see Table F-5, Footnote h, for sources of exposure).
F-20
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-9. Radiological health impacts to subsurface facility workers from radon exposure during
construction phase."
Thermal load scenario
Worker group
High
Intermediate
Low
Involved
Full-time equivalent work years"*
2,300
2,460
2,460
Maximally exposed individual (MEI)
570
660
660
worker dose (millirem)
Latent cancer fatality probability for MEI
0.0002
0.0003
0.0003
Collective dose (person-rem)
260
320
320
Latent cancer fatality incidence
0.10
0.13
0.13
Noninvolved
Full-time equivalent work years
600
600
600
Maximally exposed individual (MEI)
430
500
500
worker dose (millirem)
Latent cancer fatality probability for MEI
0.0002
0.0002
0.0002
Collective dose (person-rem)
52
60
60
Latent cancer fatality incidence
0.02
0.02
0.02
All workers (totals f
Full-time equivalent work years
2,900
3,060
3,060
Collective dose (person-rem)
310
380
380
Latent cancer fatality incidence
0.12
0.15
0.15
a. Source: Exposure data from Table F-5.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
Table F-10. Radiological health impacts to subsurface facility workers from ambient radiation exposure
during construction phase.'
Thermal load scenario
Worker group
High
Intermediate
Low
Involved
Full-time equivalent work years'"
2,300
2,460
2,460
Maximally exposed individual (MEI)
200
200
200
worker dose (millirem)
Latent cancer fatality probability for MEI
0.00008
0.00008
0.00008
Collective dose (person-rem)
92
98
98
Latent cancer fatality incidence
0.04
0.04
0.04
Noninvolved
Full-time equivalent work years
600
600
600
Maximally exposed individual (MEI)
150
150
150
worker dose (millirem)
Latent cancer fatality probability for MEI
0.00006
0.00006
0.00006
Collective dose (person-rem)
18
18
18
Latent cancer fatality incidence
0.007
0.007
0.007
All workers ( totals f
Full-time equivalent work years
2,900
3,060
3,060
Collective dose (person-rem)
110
120
120
Latent cancer fatality incidence
0.04
0.05
0.05
a. Source: Exposure data from Table F-5.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
F.2.3.2 Occupational Health and Safety Impacts During the Operations Period
F.2.3.2.1 Industrial Safety Hazards to Workers
Tables F-1 1, F-1 2, and F-1 3 list estimated impacts for each worker group during waste receipt and
packaging, drift development, and emplacement activities during the operations period.
F-21
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-11. Industrial hazard health and safety impacts to surface facility workers during waste receipt
and packaging period (24 years)."
Worker group
Uncanistered
Waste packaging option
Disposable canister Dual-purpose canister
Involved
Full-time equivalent work years'"
17,500
Total recordable cases of injury and illness
520
Lost workday cases
210
Fatalities
0.51
Noninvolved
Full-time equivalent work years
13,150
Total recordable cases of injury and illness
430
Lost workday cases
210
Fatalities
0.38
All workers (totals f
Full-time equivalent work years
30,650
Total recordable cases of injury and illness
960
Lost workday cases
440
Fatalities
0.89
11,470
340
140
0.33
11,620
380
190
0.34
23,090
730
340
0.67
11,810
350
140
0.34
11,760
390
190
0.34
23,570
740
340
0.68
a. Source: Impact rates from Table F-2.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
Table F-12. Industrial hazard health and safety impacts to subsurface facility workers during drift
development period."
Thermal load scenario
High
Intermediate
Low
Worker group
(21 years)
(21 years)
(22 years)
Involved
Full-time equivalent work years'"
6,230
6,230
6,530
Total recordable cases of injury and illness
420
420
440
Lost workday cases
300
300
310
Fatalities
0.18
0.18
0.19
Noninvolved
Full-time equivalent work years
1,670
1,670
1,670
Total recordable cases of injury and illness
19
19
19
Lost workday cases
12
12
12
Fatalities
0.05
0.05
0.05
All workers (totals f
Full-time equivalent work years
7,900
7,900
8,210
Total recordable cases of injury and illness
440
440
460
Lost workday cases
310
310
330
Fatalities
0.23
0.23
0.24
a. Source: Impact rates from Tables F-2 and F-3.
b. Source: Table F- 1.
c. Totals might differ from sums due to rounding.
F.2.3.2.2 Radiological Health Impacts to Workers
Radiological health impacts to surface and subsurface facility workers for the operations period are the
sum of the estimates of impacts to surface facility workers and subsurface facility workers during
operation and monitoring (see Section F.2.3.3.2 for monitoring period).
• Table F-14 lists radiation dose to subsurface facility workers from radiation emanating from waste
packages during emplacement operations.
F-22
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-13. Industrial hazard health and safety impacts to
subsurface facility workers during emplacement peiiod/
For all thermal
Worker group
load scenarios
Involved
Full-time equivalent work years''
1,780
Total recordable cases of injury and illness
53
Lost workday cases
21
Fatalities
0.05
Noninvolved
Full-time equivalent work years
380
Total recordable cases of injury and illness
13
Lost workday cases
6
Fatalities
0.01
All workers (totals f
Full-time equivalent work years
2,160
Total recordable cases of injury and illness
66
Lost workday cases
29
Fatalities
0.06
a. Source: Impact rates from Table F-2.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
Table F-14. Radiological health impacts
emplacement period (24 years).'
to subsurface
facility
workers from waste packages during
Worker group
Thermal load scenario
High
Intermediate
Low
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEf
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
1,780
1,780
4,460
4,510
0.002
0.002
240
240
0.10
O.IO
380
380
190
190
0.00008
0.00008
3
3
0.001
0.001
2,160
2,160
240
250
0.10
0.10
1,780
2,490
0.001
140
0.05
380
190
0.00008
3
0.001
2,160
140
0.06
Source: Exposure data from Table F-5.
b. Source: Table F-1.
c. MEI = maximally exposed individual.
d. Totals might differ from suras due to rounding.
• Table F-1 5 lists radiation dose to subsurface workers from the ambient radiation in the drifts during
emplacement operations. Table F-16 lists radiation doses to subsurface facility workers from ambient
radiation during the drift development period.
• Table F-1 7 lists radiation dose to subsurface workers from inhalation of airborne radon-222 in the
drift atmosphere during emplacement operations. Table F-1 8 lists radiation dose to subsurface
workers from inhalation of airborne radon-222 during drift development operations.
F-23
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-15. Radiological health impacts to subsurface facility workers from ambient
radiation during emplacement period/
Worker group
Values are independent of
thermal load scenario
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI*^
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
1,780
960
0.0004
71
0.03
380
480
0.0002
8
0.003
2,160
79
0.03
a. Source: Exposure data from Table F-5.
b. Source: Table F-1.
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
Table F-16. Radiological health impacts to subsurface facility workers from ambient radiation during
drift development period."
Thermal load scenario
Worker group
High
(21 years)
Intermediate
(21 years)
Low
(22 years)
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI'
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals/
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
6,230
6,230
6,530
880
880
880
0.0004
0.0004
0.0004
250
250
260
0.10
0.10
0.10
1,670
1,670
1,670
660
660
660
0.0003
0.0003
0.0003
50
50
50
0.02
0.02
0.02
7,900
7,900
8,210
300
300
310
0.12
0.12
0.12
a. Source: Exposure data from Table F-5.
b. Source: Table F-L
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
F-24
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-17. Radiological health impacts to subsurface facility workers from airborne radon-222
during emplacement period."
Thermal load scenario
Worker group
High
Intermediate
Low
1,780
1,780
1,780
1,580
2,160
4,180
0.0006
0.0008
0.002
120
160
310
0.05
0.06
0.12
380
380
380
790
1,080
2,090
0.0003
0.0004
0.0008
13
17
33
0.005
0.007
0.01
2,160
2,160
2,160
130
180
340
0.05
0.07
0.14
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEf
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
a. Source: Exposure data from Table F-5.
b. Source: Table F-1.
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
Table F-18. Radiological health impacts to subsurface facility workers from airborne radon-222 during
development period.^
Thermal load scenario
Worker group
High
(21 years)
Intermediate
(21 years)
Low
(22 years)
Involved
Full-time equivalent work years'*
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEf
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers ( totals f
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
6,230
6,230
6,530
790
790
790
0.0003
0.0003
0.0003
220
220
240
0.09
0.09
0.09
1,670
1,670
1,670
590
590
590
0.0002
0.0002
0.0002
45
45
45
0.02
0.02
0.02
7,900
7,900
8,210
270
270
280
0.11
0.11
0.11
a. Source: Exposure data from Table F-5.
b. Source: Table F-L
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
F-25
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
i
F.2.3.3 Occupational Health and Safety Impacts to Workers During the Monitoring Period
F.2.3.3.1 Health and Safety Impacts to Workers from Workplace Industrial Hazards ■{
Health and safety impacts from industrial hazards common to the workplace for the monitoring period
consist of the following:
• Impacts to surface facility workers for the 3-year surface facility decontamination period (Table F-19)
• Impacts to surface facility workers for monitoring support activities (Table F-20)
• Impacts to subsurface facility workers for monitoring and maintenance activities (Table F-21)
Table F-19. Industrial hazard health and safety impacts to surface facility workers during
decontamination period/
Impact
Uncanistered Disposable canister Dual-purpose canister
Full-time equivalent work years
4,060
2,950
Total recordable cases of injury and illness
120
88
Lost workday cases
49
35
Fatalities
0.13
0.08
3,070
92
37
0.11
a. Source: Incident rate data from Table F-2.
b. Source: Table F-1.
Table F-20. Industrial hazard health and safety impacts to surface facility workers
during monitoring period."
Worker group
Phase
Annual
Full-time equivalent work years'"
2,660
35
Total recordable cases of injury and illness
80
1.1
Lost workday cases
32
0.42
Fatalities
0.08
0.001
a. Source: Impacts rates from Table F-2.
b. Source: Table F-1.
Table F-21. Industrial hazard health and safety impacts for subsurface facility workers during
monitoring period."
Thermal load scenario
Worker group
High
Intermediate
Low
Involved
Full-time equivalent work years'"
5,240
5,240
5,780
Total recordable cases of injury and illness
160
160
170
Lost workday cases
63
63
69
Fatalities
0.15
0.15
0.17
Noninvolved
Full-time equivalent work years
990
990
990
Total recordable cases of injury and illness
32
32
32
Lost workday cases
16
16
16
Fatalities
0.03
0.03
0.03
All workers ( totals f
Full-time equivalent work years
6,230
6,230
6,760
Total recordable cases of injury and illness
190
190
210
Lost workday cases
84
84
91
Fatalities
0.18
0.18
0.20
a. Source: Impacts rates from Table F-2.
b. Source: Table F-1.
c. Totals may differ from sums due to rounding.
For surface monitoring support activities, annual impact values are listed to facilitate the extrapolation of
the data for longer and shorter monitoring periods.
F-26
k
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
F.2.3.3.2 Radiological Health Impacts to Workers
F.2.3.3.2.1 Surface Facility Workers. During monitoring, surface workers would be involved in
two types of activities — decontamination for 3 years after the completion of emplacement and support of
subsurface monitoring for 76 years (starting at the end of emplacement). Surface workers providing
support to the subsurface activities would receive very little radiological dose in comparison to their
counterparts involved in subsurface monitoring activities. Therefore, radiological dose impacts were not
included for this group; they are estimated in Appendix G, Section G.2. Radiological health impact
estimates for the surface facilities decontamination activities are listed in Table F-22.
Table F-22. Radiological health impacts to surface facility workers during decontamination period.^
Worker group Uncanistered Disposable canister Dual-purpose canister
Full-time equivalent work years'" 4,060 2,950 3,070
Maximally exposed individual worker (millirem)" 300 300 300
Latent cancer fatality probabiUty for MEr* 0.0001 0.0001 0.0001
Collective dose (f)erson-rem) 290 210 220
Latent cancer fatality incidence 0.11 0.08 0.09
a. Source: Dose rate data from Table F-5.
b. Source: Table F-1.
c. Source: Based on Table F-4, maximum dose of 100 millirem per year for 3 years.
d. MEI = maximally exposed individual.
F.2.3.3.2.2 Subsurface Facility Woriters. Radiological health impacts to subsurface facility
workers during monitoring are listed in Table F-23. Maximum worker dose values in the table were
based on a maximum work period of 50 years on a monitoring assignment rather than a 76-year
monitoring period.
Table F-23. Radiological health impacts to subsurface facility workers during a 50-year work period
during a 76-year monitoring period.'
Thermal load scenario
Worker group
High
Intermediate
Low
Involved
Full-time equivalent work years'"
5,240
5,240
5,780
Dose to maximally exposed individual worker
16,240
18,940
17,610
(millirem)
Latent cancer fatality probability for MEP
0.006
0.008
0.007
Collective dose (person-rem)
1,760
2,050
2,060
Latent cancer fatality incidence
0.71
0.82
0.83
Noninvolved
Full-time equivalent work years
990
990
990
Dose to maximally exposed individual worker
6,200
7,550
8,000
(millirem)
Latent cancer fatality probability for MEI
0.003
0.003
0.003
Collective dose (person-rem)
120
150
160
Latent cancer fatality incidence
0.05
0.06
0.06
All workers {totalsf
Full-time equivalent work years
6,230
6,230
6,760
Collective dose (person-rem)
1,880
2,200
2,220
Latent cancer fatality incidence
0.75
0.88
0.89
a. Source: Exposure data from Table F-4.
b. Source: Table F-1.
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
F-27
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
In addition, DOE considered monitoring periods as short as 26 years and as long as 276 years.
Radiological health impacts for both of these monitoring periods were evaluated; the radiological health
impact estimates are listed in Table F-24. Doses to the maximally exposed worker were based on a
50-year employment period rather than the 276-year monitoring period.
Table F-24. Radiological health impacts to workers during a 26-year and a 276-year monitoring period,
dual-purpose canister packaging scenario." ___^
26 years 276 years
High Low High Low
thermal Intermediate thermal thermal Intermediate thermal
Group load thermal load load load thermal load load
Involved
Full-time equivalent work years
Dose to maximally exposed
individual worker (millirem)
Latent cancer fatality probability
for MEf
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed
individual worker (millirem)
Latent cancer fatality probability
forMEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
a. Sources: Tables F-1, F-4, and F-23.
b. MEI = maximally exposed individual.
F.2.3.4 Occupational Health and Safety Impacts During the Closure Phase
F.2.3.4.1 Health and Safety Impacts to Workers from Workplace Industrial Hazards
Health and safety impacts to workers from industrial hazards common to the workplace for closure are
listed in Table F-25 for surface facility workers and Table F-26 for subsurface facility workers.
F.2.3.4.2 Radiological Health Impacts to Workers
Radiological health impact to workers from closure activities are the sum of the following components:
• Radiological health impacts to subsurface workers from radiation emanating from the waste packages
during the closure phase (Table F-27)
• Radiological impacts to subsurface workers from the ambient radiation field in the drifts during the
closure phase (Table F-28)
• Radiological impacts to subsurface workers from inhalation of radon-222 in the drift atmosphere
during the closure phase (Table F-29)
F-28
1,790
1,790
1,980
19,040
19,040
20,980
8,440
9,850
9,160
16,240
18,940
17,610
0.003
0.004
0.004
0.006
0.008
0.007
600
700
710
6,400
7,430
7,500
0.24
0.28
0.28
2.6
3.0
3.0
340
340
340
3,590
3,590
3,590
3,220
3,930
4,160
6,200
7,550
8,000
0.001
0.002
0.002
0.002
0.003
0.003
42
51
54
450
540
570
0.02
0.02
0.02
0.18
0.22
0.23
2,130
2,130
2,320
22,630
22,630
24,570
640
750
760
6,850
7,970
8,073
0.26
0.30
0.30
2.7
3.2
3.2
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-25. Industrial hazard health and safety impacts to surface facility workers during closure phase.
Waste packaging option
Worker group
Uncanistered Disposable canister
Dual-purpose
canister
Involved
Full-time equivalent work years'"
Total recordable cases of injury and illness
Lost workday cases
Fatalities
Noninvolved
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
All workers (totals f
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
1,580
1,110
1,200
97
68
73
46
33
35
0.04
0.03
0.03
600
420
460
20
14
15
10
7
7
0.02
0.01
0.01
2,180
1,540
1,650
120
82
88
56
40
43
0.06
0.04
0.04
a. Source: Impact rates from Table F-2.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
Table F-26. Industrial hazard health and safety impacts to subsurface facility workers during closure
phase.^
Thermal load scenario
High
Intermediate
Low
Worker group
(6 years)
(6 years)
(15 years)
Involved
Full-time equivalent work years'"
1,310
1,310
3,270
Total recordable cases of injury and illness
80
80
200
Lost workday cases
39
39
96
Fatalities
0.04
0.04
0.09
Noninvolved
Full-time equivalent work years
260
260
660
Total recordable cases of injury and illness
9
9
22
Lost workday cases
4
4
11
Fatalities
0.01
0.01
0.02
All workers (totals f
Full-time equivalent work years
1,570
1,570
3,930
Total recordable cases of injury and illness
89
89
220
Lost workday cases
43
43
110
Fatalities
0.05
0.05
0.11
a. Source: Impact rates from Table F-2.
b. Source: Table F-1.
c. Totals might differ from sums due to rounding.
Because the surface facilities would be largely decontaminated at the beginning of the monitoring period
(the exception would be a small facility retained to handle an operations emergency), radiological health
impacts to surface facility workers during closure would be small in comparison to those to the subsurface
facility workers and so are not included here.
F-29
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-27. Radiological health impacts to subsurface facility workers from waste package radiation
exposures during closure phase." ^^
Thermal load scenario
Worker group
High
(5 years)
Intermediate
(6 years)
Low
(15 years)
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEf
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
1,310
1,310
650
650
0.0003
0.0003
75
75
0.03
0.03
260
260
48
48
0.00002
0.00002
2
2
0.0008
0.0008
1,570
1,570
77
77
0.03
0.03
3,270
960
0.0004
110
0.04
660
120
0.00005
5
0.002
3,930
115
0.05
a. Source: Exposure data from Table F-5.
b. Source: Table F-1.
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
Table F-28. Radiological health impacts to subsurface facility workers from ambient radiation exposures
during closure phase."
Thermal load scenario
Worker group
High
Intermediate
Low
(5 years)
(6 years)
(15 years)
1,310
1,310
3,270
240
240
600
0.0001
O.OOOI
0.0002
52
52
130
0.02
0.02
0.05
260
260
660
180
180
450
0.00006
0.00007
0.00018
8
8
20
0.003
0.003
0.008
1,570
1,570
3,930
60
60
150
0.02
0.02
0.06
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEf
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
a. Source: Exposure data from Table F-5.
b. Source: Table F-L
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
F-30
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-29. Radiological health impacts
closure phase.'
to subsurface
facility workers from radon-222 exposure during
Worker group
Thermal load scenario
High
(5 years)
Intermediate
(6 years)
Low
(15 years)
Involved
Full-time equivalent work years""
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEf
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
1,310
1,310
3,270
1,150
1,480
3,960
0.0005
0.0006
0.002
250
320
860
0.10
0.13
0.35
260
260
660
860
1,110
2,970
0.0003
0.0004
0.001
38
49
130
0.02
0.02
0.05
1,570
1,570
3,930
290
370
990
0.12
0.15
0.40
a. Source: Exposure data from Table F-5.
b. Source: Table F-1.
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
F.3 Human Health and Safety Impact Analysis
for Inventory Modules 1 and 2
DOE performed an analysis to estimate the occupational and public health and safety impacts from the
emplacement of Inventory Module 1 or 2. Module 1 would involve the emplacement of additional spent
nuclear fuel and high-level radioactive waste in the repository; Inventory Module 2 would emplace
commercial Greater-Than-Class-C waste and DOE Special-Performance-Assessment-Required waste,
which is equivalent to commercial Greater-Than-Class-C waste, in addition to the inventory from
Module 1. The volumes of Greater-Than-Class-C and Special-Performance-Assessment-Required waste
would be less than that for spent nuclear fuel and high-level radioactive waste (TRW 1999c, Table 3.1).
Waste packages containing these materials would be placed between the waste packages containing spent
nuclear fuel and high-level radioactive waste (see Chapter 8, Section 8.1.2.1).
With regard to estimating heath and safety impacts for the inventory modules, the characteristics of the
spent nuclear fuel and high-level radioactive waste were taken to be the same as those for the Proposed
Action, but there would be more material to emplace (see Appendix A, Section A.2). As described in
Appendix A, the radiological content of the Greater-Than-Class-C waste and Special-Performance-
Assessment-Required waste, which is the additional material in Module 2, is much less than that for spent
nuclear fuel and high-level radioactive waste. Therefore, the emplacement of the Module 2 material
would not meaningfully increase radiological impacts to workers over those estimated for the Module 1
inventory. Further, the facility design parameters, on which the impact estimates are based, are
extrapolations from existing designs and have some uncertainty associated with them [see, for example,
TRW (1999c), Section 6.2, first paragraph]. Therefore, separate occupational and public health and
safety impact analyses were not performed for Module 2 because the impacts for Inventory Modules 1
and 2 would not differ meaningfully.
F-31
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
The calculation of health and safety impacts to workers assumed that the throughput rate of materials for
the facility would remain the same as that assumed for the Proposed Action during repository operations
(that is, the 70,000-MTHM case). In addition, for the inventory modules the period of operations would
be extended to accommodate the additional materials, and the monitoring period would be reduced such
that the Yucca Mountain repository operations and monitoring activities would still occur in a 100-year
period. Table F-30 summarizes the expected lengths of the phases for Yucca Mountain Repository
activities for the inventory modules. These periods were used in the occupational and public health and
safety impact calculations.
Table F-30. Expected durations (years) of the Proposed Action and Inventory Modules 1 and 2.°
Construction
phase Operation and monitoring phase (2010-21 10) Closure phase
Inventory (2005-2010) Development'' Emplacement Monitoring Total (starts in 21 10)
Proposed Action 5 22 24 76 100' G-IS**
Module 1 or 2 5 36 38 62 100 13-27°
a. Sources: TRW (1999b, all); TRW (1999c, all); Jessen (1999, all).
b. Continuing subsurface construction (development) activities are concurrent with emplacement activities. i
c. Closure is assumed to begin 100 years following initial emplacement for the Proposed Action and Module 1 or 2 for the
evaluation of cumulative impacts.
d. 6, 6, and 15 years for the high, intermediate, and low thermal load scenarios, respectively.
e. 13, 17, and 27 years for the high, intermediate, and low thermal load scenarios, respectively.
This section discusses the methodologies and data used to estimate occupational radiological health and
safety impacts resulting from construction, operation and monitoring, and closure of the Yucca Mountain
Repository for Inventory Modules 1 and 2, and presents the detailed results. Section F.3.1 describes the
methods DOE used to estimate impacts. Section F.3.2 contains tabulations of the detailed data used in the
impact calculations and references to the data sources. Section F.3.3 contains detailed tabulations of
results.
F.3.1 METHODOLOGY FOR CALCULATING HUMAN HEALTH AND SAFETY IMPACTS
DOE used the methodology described in Section F.2.I to estimate health and safety impacts for the
inventory modules. This methodology involved assembling data for the number of full-time equivalent
workers for each repository phase. These numbers were used with statistics for the likelihood of an
impact (industrial hazards) or the expected dose rate in the worker environment to calculate health and
safety impacts. The way in which the input data was combined in the calculation of health and safety
impacts is described in more detail in Section F.2.I. Some of the input data for the calculations for the
inventory modules are different from those for the Proposed Action, as discussed in the next section.
F.3.2 DATA SOURCES AND TABULATIONS
F.3.2.1 Full-Time Equivalent Worker- Year Estimates for the Repository Phases for
Inventory Modules 1 and 2
The full-time equivalent work-year estimates for the inventory modules are different from those for the
Proposed Action. Table F-3 1 lists the number of full-time equivalent work years for the various
repository phases for the inventory modules. Each full-time equivalent work year represents 2,000 work
hours, the hours assumed to be worked in a normal work year.
This analysis divides the repository workforce into two groups — ^involved and noninvolved workers (see
Section F.2 for definitions of involved and noninvolved workers). It did not consider workers whose
place of employment would be other than at the repository site.
F-32
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
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F-33
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
F.3.2.2 Statistics on Health and Safety Impacts from Industrial Hazards in the Workplace
DOE used the same statistics for health and safety impacts from industrial hazards common to the
workplace that were used for the Proposed Action (70,000 MTHM) for analyzing the inventory module
impacts (see Table F-2).
F.3.2.3 Estimates of Radiological Exposure Rates and Times for Inventory
Modules 1 and 2
DOE used the values in Table F-5 (Proposed Action) for exposure rates, occupancy times, and the
fraction of the workforce that would be exposed to estimate radiological health impacts for the inventory
module cases, except for doses from the waste packages and from radon-222 inhalation for the subsurface
emplacement, monitoring, and closure phases. Annual exposures to subsurface workers for Inventory
Modules 1 and 2 from radiation emanating from the waste packages are listed as part of Table F-6.
Table F-32 lists annual dose rates from inhalation of radon-222 and its decay products. Section F. 1 . 1 .6
discusses the basis for the values in Table F-32.
Table F-32. Correction factors and annual exposures from radon-222 and its decay products for the
project phases or periods for Inventory Modules 1 and 2."
Correction factor Annual dose rate (millirem per year)
Subsurface project period
High
Intermediate
Low
High
Intermediate
Low
Construction
2.1
2.1
2.1
126
126
126
Drift development
0.6
0.6
0.6
36
36
36
Emplacement
2.0
1.7
3.5
120
120
210
Monitoring
4.2
2.7
4.1
252
160
246
Closure
4.2
2.7
4.1
252
160
246
a. Based on measured value of 60 millirem per year corrected for ref)ository volume and ventilation rate; see the discussions in
Section F.1.1.6 and Appendix G (Section G.2.3.1).
F.3.3 DETAILED HUMAN HEALTH AND SAFETY IMPACTS TO WORKERS - INVENTORY
MODULES 1 AND 2
F.3.3.1 Construction Phase
F.3.3.1 .1 Industrial Hazards to Workers
This section details health and safety impacts to workers from industrial hazards common to the
workplace for the construction phase. Impact values for surface workers are the same as those presented
for the Proposed Action in Table F-7. Impact values for subsurface workers are presented in Table F-33.
The subsurface impacts are independent of thermal load or packaging scenarios.
F.3.3.1 .2 Radiological Health Impacts to Workers
Table F-34 lists subsurface worker health impacts from inhalation of radon-222 and its decay products in
the subsurface atmosphere and from exposure to natural radiation from radionuclides in the drift walls.
The radiological health impacts to surface workers from inhalation of radon-222 and its decay products
would be small in comparison to those for subsurface workers; therefore, they are not tabulated here (see
Table F-5, Footnote h).
F-34
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-33. Industrial hazard health and safety
impacts to subsurface facility workers during
construction phase - Inventory Module 1 or 2."
Worker group
Impacts
Involved
Full-time equivalent work years'"
Total recordable cases of injury and illness
2,460
150
Lost workday cases
72
Fatalities
0.07
Noninvolved
Full-time equivalent work years
Total recordable cases of injury and illness
600
20
Lost workday cases
10
Fatalities
0.02
All workers (totals f
Full-time equivalent work years
Total recordable cases of injury and illness
3,060
170
Lost workday cases
82
Fatalities
0.09
a. Source: Impact rates from Table F-2.
b. Source: Table F-31.
c. Totals might differ from sums due to rounding.
Table r-34. Radiological health impacts to subsurface facility workers from radon inhalation and natural
exposure for the construction phase - Inventory Modules 1 and 2.'
Worker group
Involved
Full-time equivalent work years'^
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEF
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual worker (millirem)
Latent cancer fatality probability for MEI
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals)
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
Radon inhalation
Subsurface ambient
exposure
exposure
2,460
2,460
630
200
0.0002
0.00008
310
98
0.12
0.04
600
600
470
150
0.0002
0.00006
57
18
0.02
0.007
3,060
3,060
370
120
0.15
0.05
a. Sources: Table F-5 (ambient exposure); Table F-32 (exposure from radon inhalation).
b. Source: Table F-31.
c. MEI = maximally exposed individual.
d. Totals might differ from sums due to rounding.
F-35
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
F.3.3.2 Operation and Monitoring Phiase
F.3.3.2.1 Health and Safety Impacts to Workers from Industrial Hazards
This section details health and safety impacts to workers from industrial hazards common to the
workplace for the operation and monitoring phase. These impacts would consist of four components:
• Health and safety impacts to surface workers for operations (Table F-35)
• Health and safety impacts to subsurface workers for emplacement and for drift development
(Table F-36)
• Health and safety impacts to subsurface workers for the monitoring period (Table F-37)
• Health and safety impacts to surface workers for surface facility decontamination and monitoring
support (Table F-38) ;
Table F-35. Industrial hazard health and safety impacts for surface facility workers during a 38-year
operations period by packaging option - Inventory Module 1 or 2/
Worker group Uncanistered Disposable canister Dual-purpose canister
Involved
Full-time equivalent work years'"
Total recordable cases of injury and illness
Lost workday cases
Fatalities
Noninvolved
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
All workers (totals)'^
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
a. Source: Impact rates from Table F-2.
b. Source: Table F-3L
c. Totals might differ from sums due to rounding.
F.3.3.2.2 Radiological Health Impacts to Workers
This section details radiological health impacts to workers during the operation and monitoring phase for
the inventory modules. These impacts consist of four components:
• Radiological health impacts to surface workers during operations (Table F-39)
• Radiological health impacts to subsurface workers during operations (emplacement and drift
development) (Table F-40)
• Radiological health impacts to workers during surface facility decontamination and monitoring
support (Table F-41)
• Radiological health impacts to subsurface workers for the monitoring period (Table F-42)
27,700
18,160
18,700
830
540
560
360
240
240
0.80
0.53
0.55
20,820
18,390
18,620
680
600
610
340
300
300
0.60
0.53
0.54
48,530
36,560
37,320
1,520
1,150
1,170
700
530
540
1.4
LI
1.1
F-36
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-36. Industrial hazard health and safety impacts for subsurface facility workers for development
Inventory Module 1 or 2."
Intermediate thermal
Worker group
High thermal load
load
Low thermal load
Involved
Full-time equivalent work years'"
11,920
12,350
13,370
Total recordable cases of injury and illness
700
730
790
Lost workday cases
480
500
540
Fatalities
0.35
0.36
0.39
Noninvolved
Full-time equivalent work years
3,060
3,060
3,380
Total recordable cases of injury and illness
48
48
52
Lost workday cases
27
27
29
Fatalities
0.09
0.09
0.10
All workers (totals f
Full-time equivalent work years
14,980
15.410
16,750
Total recordable cases of injury and illness
750
780
850
Lost workday cases
500
530
570
Fatalities
0.42
0.45
0.49
a. Source: Impact rates from Tables F-2 and F-3.
b. Source: Table F-3 1 .
c. Totals might differ from sums due to rounding.
Table F-37. Industrial hazard health and safety impacts for subsurface facility workers during
monitoring period - Inventory Module 1 or 2.'
Intermediate
Worker group
High thermal load
thermal load
Low thermal load
Involved
Full-time equivalent work years''
4,280
4,710
5,950
Total recordable cases of injury and illness
130
140
180
Lost workday cases
55
61
77
Fatalities
0.12
0.14
0.17
Noninvolved
Full-time equivalent work years
810
810
1610
Total recordable cases of injury and illness
26
26
53
Lost workday cases
13
13
26
Fatalities
0.02
0.02
0.05
All workers (totals f
Full-time equivalent work years
5,080
5.520
7,560
Total recordable cases of injury and illness
160
170
230
Lost workday cases
68
74
100
Fatalities
0.15
0.16
0.22
a. Source: Impact rates from Table F-2.
b. Source: Table F-3 1.
c. Totals might differ from sums due to rounding.
Table F-38. Industrial hazard health and safety impacts by packaging option to workers during surface
facility decontamination and monitoring period - Inventory Module 1 or 2.'
Involved workers Uncanistered Disposable canister Dual-purpose canister
Full-time equivalent work years" 6,230 5,120 5,240
Total recordable cases of injury and illness 190 150 160
Lost workday cases 80 70 70
Fatalities 0,18 015 0^5
a Source: Impact rates from Table F-2.
b. Source: Table F-3 1.
F-37
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-39. Radiological health impacts to surface facility workers for a 38-year operations period -
Inventory Module 1 or 2."
Worker group
Uncanistered Disposable canister Dual-purpose canister
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual
worker (millirem)
Latent cancer fatality probability for
maximally exposed individual
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual
worker (millirem)
Latent cancer fatality probability for
maximally exposed individual
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals f
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
27,700
18,160
18,700
15,200
15,200
15,200
0.006
0.006
0.006
8,180
3,890
3,950
3.3
L6
1.6
20,820
18,390
18,620
950
950
950
0.0004
170
0.07
48,530
8,350
3.3
0.0004
140
0.06
36,560
4,030
1.6
0.0004
140
0.06
37,320
4,090
1.6
a. Source: Exposure data from Table F-5.
b. Source: Table F-31.
c. Totals might differ from sums due to rounding.
Table F-40. Radiological health impacts to subsurface workers for emplacement and drift development
during operations period - Inventory Module 1 or 2.^
Worker group
Intermediate thermal
High thermal load load
Low thermal load
Involved
Full-time equivalent work years'"
Dose to maximally exposed individual
worker (millirem)
Latent cancer fatality probability for
maximally exposed individual
Collective dose (person-rem)
Latent cancer fatality incidence
Noninvolved
Full-time equivalent work years
Dose to maximally exposed individual
worker (millirem)
Latent cancer fatality probability for
maximally exposed individual
Collective dose (person-rem)
Latent cancer fatality incidence
All workers (totals f
Full-time equivalent work years
Collective dose (person-rem)
Latent cancer fatality incidence
11,900
12,350
13,370
13,220
12,530
13,460
0.005
0.005
0.005
1,530
1,510
1,770
0.61
0.60
0.71
3,060
3,060
3,380
2,280
2,240
4,290
0.0009
190
0.08
14,980
1,720
0.69
0.0009
190
0.08
15,410
1,700
0.68
0.002
240
0.10
16,750
2,010
0.80
a. Source: Exposure data from Table F-4 except waste package exposures, which are from Table F-6.
b. Source: Table F-31.
c. Totals might differ from sums due to rounding.
F-38
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-41. Radiological health impacts to surface facility workers for decontamination and monitoring
support - Inventory Module 1 or 2.'
Involved workers Uncanistered Disposable canister Dual-purpose canister
Full-time equivalent work years'" 6,230 5,120 5,240
Dose to maximally exposed individual 300 300 300
worker (millirem)
Latent cancer fatality probability for 0.0001 O.OOOI 0.0001
maximally exposed individual
Collective dose (person-rem) 290 210 220
Latent cancer fatality incidence 0.11 0.08 0.09
a. Source: Exposure data from Table F-4.
b. Source: Table F-31.
Table F-42. Radiological health impacts to subsurface facility workers for a 62-year monitoring period -
Inventory Module I or 2.'
Intermediate thermal
Worker group High thermal load load Low thermal load
Involved
Full-time equivalent work years'" 4,280 4,710 5,950
Dose to maximally exposed individual 19,240 14,740 16,710
worker (millirem)
Latent cancer fatality probability for 0.008 0.006 0.007
maximally exposed individual
Collective dose (person-rem) 1,700
Latent cancer fatality incidence 0.68
Noninvolved
Full-time equivalent work years 8 10
Dose to maximally exposed individual 7,700
worker (millirem)
Latent cancer fatality probability for 0.003 0.002 0.003
maximally exposed individual
Collective dose (person-rem) 120
Latent cancer fatality incidence 0.05
All workers (totals f
Full-time equivalent work years 5,080
Collective dose (person-rem) 2,300
Latent cancer fatality incidence 0.92
a. Source: Exposure data from Table F-5 except for exposure from waste packages, which is from Table F-6.
b. Source: Table F-31.
c. Totals might differ from sums due to rounding.
F.3.3.3 Closure Phase
F.3.3.3.1 Health and Safety Impacts to Workers from Industrial Hazards
This section details health and safety impacts to workers from industrial hazards common to the
workplace for the closure phase. The impacts would consist of two components — impacts to surface
workers supporting the closure operations, and impacts to subsurface workers during the closure phase.
These impacts are listed in Tables F-43 and F-44, respectively.
1,440
2,050
0.58
0.82
810
1,610
5,450
7,550
88
240
0.04
0.10
5,520
7,560
2,050
2,470
0.82
3.0
F-39
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-43. Industrial hazard health and safety impacts to surface workers during the closure phase -
Inventory Module 1 or 2." ^^^^
Worker group Uncanistered Disposable canister Dual-purpose canister
Involved
Full-time equivalent work years'"
Total recordable cases of injury and illness
Lost workday cases
Fatalities
Noninvolved
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
All workers ( totals f
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
1,580
1,110
1,200
97
68
73
46
33
35
0.05
0.03
0.04
600
420
460
20
14
15
10
7
7
0.02
0.01
0.01
2,180
1,540
1,650
116
82
88
56
40
43
0.06
0.04
0.05
a. Source: Impact rates from Table F-2.
b. Source: Table F-3L
c. Totals might differ from sums due to rounding.
Table F-44. Health and safety impacts to subsurface facility workers from industrial hazards during the
closure phase - Inventory Module 1 or 2."
High Intermediate Low
Worker group thermal load thermal load thermal load
Involved
Full-time equivalent work years'"
Total recordable cases of injury and illness
Lost workday cases
Fatalities
Noninvolved
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
All workers (totals f
Full-time equivalent work years
Total recordable cases of injury and illness
Lost workday cases
Fatalities
a. Source: Impact rates from Table F-2.
b. Source: Table F-3 1 .
c. Totals might differ from sums due to rounding.
F.4 Human Health and Safety Impact Analysis
for the Retrieval Contingency
Nuclear Regulatory Commission regulations state that the period for which DOE must maintain the
ability to retrieve waste is at least 50 years after the start of emplacement operations [lOCFR 60.111(b)].
Although DOE does not anticipate retrieval and it is not part of the Proposed Action, the Department
would maintain the ability to retrieve the waste for at least 100 years and possibly for as long as 300 years
I
2,830
3,710
5,890
170
230
360
84
110
170
0.08
0.11
0.17
570
750
1,190
19
25
39
9
12
19
0.02
0.02
0.03
3,410
4,450
7,070
193
250
400
93
120
190
0.10
0.13
0.21
F-40
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
after the start of emplacement. Factors that could lead to a decision to retrieve the waste would be (1) to
protect the public health and safety or the environment or (2) to recover resources from spent nuclear fuel.
This EIS evaluates retrieval as a contingency action and describes potential impacts should it occur. The
analysis assumes that under this contingency DOE would retrieve all the waste associated with the
Proposed Action and would place it on surface storage pads pending future decisions about its ultimate
disposition.
The analysis of health and safety impacts to workers divided the retrieval period into two subperiods, as
follows:
• First, a construction subperiod in which DOE would (1) build the surface facilities necessary to
handle and enclose retrieved waste packages in concrete storage units in preparation for placement on
concrete storage pads, and (2) construct the concrete storage pads.
No radioactive materials would be involved in the construction subperiod, so health and safety
impacts would be limited to those associated with industrial hazards in the workplace. DOE expects
this subperiod would last 2 to 3 years, although construction of the concrete storage pads probably
would continue on an as-needed basis during most of the operations subperiod. The analysis assumed
a 3-year period.
• Second, an operations subperiod during which the waste packages would be retrieved and moved to
the Waste Retrieval Transfer Building. Surface facility workers would unload the waste package
from the transfer vehicle and place it on a concrete base. The package and concrete base would then
be enclosed in a concrete storage unit that would be placed on the concrete storage pad. The analysis
assumed an 1 1-year period.
This section discusses the methodologies and data used to estimate human health and safety impacts
resulting from the retrieval contingency. Section F.4. 1 describes the methods DOE used to estimate
impacts. Section F.4.2 contains tabulations of the detailed data used in the impact calculations and
references to the data sources. Section F.4.3 contains detailed tabulations of the results.
F.4.1 METHODOLOGY FOR CALCULATING HUMAN HEALTH AND SAFETY IMPACTS
DOE used the methodology summarized in Section F.2. 1 to estimate health and safety impacts for the
retrieval contingency. This involved assembling data for the number of full-time equivalent workers for
each retrieval activity. These numbers were used with statistics on the likelihood of an impact (industrial
hazards), or the estimated radiological dose rate in the worker environment, to calculate health and safety
impacts. The way in which the input data were combined to calculate health and safety impacts is
described in more detail in Section F.2.1. Some of the input data in the retrieval impact calculations are
different from those for the Proposed Action, as described in the next section.
F.4.2 DATA SOURCES AND TABULATIONS
F.4.2.1 Full-Time Equivalent Work- Year Estimates for the Retrieval Contingency
This analysis divides the repository workforce into two groups — involved and noninvolved workers (see
Section F.2 for definitions of involved and noninvolved workers).
Table F-45 lists the number of workers involved in the two subperiods of the retrieval operation and the
sources of the numbers. They are tabulated as full-time equivalent work years. Each full-time equivalent
F-41
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-45. Full-time equivalent work-year estimates for retrieval.
Length of subperiod Full-time equivalent
Subperiod and worker group (years) work years
Surface facilities, construction" 3
Involved 1,130
Noninvolved 430
Surface facilities, retrieval support 11
Involved 320
Noninvolved 870
Subsurface facility retrieval operations'^ 11
Involved 810
Noninvolved 180
Total 3,740
a. Source: TRW (1999c, Table 1-2).
b. Source: TRW (1999c, Table 1-3).
c. Source: TRW (1999b, Table 6.1.5.1-1).
I
work year represents 2,000 work hours, the hours assumed to be worked in a normal work year. The fulT
time equivalent work year estimates are independent of thermal load. _^
F.4.2.2 Statistics on Health and Safety Impacts from Industrial Hazards in the Workplace
For the retrieval contingency, DOE used the same set of statistics on health and safety impacts from
industrial hazards common to the workplace that were used for the Proposed Action (70,000 MTHM) (see
Table F-2). The specific statistics that were applied to the retrieval contingency subphases are listed in
Table F-46.
Table F-46. Statistics for industrial hazard impacts for retrieval.
Total recordable incidents
Lost workday cases
Fatalities
Subperiod and worker group
(rate per 100 FTEs)'
(rate
per lOOFTEs)
(rate per 100,000 FTEs)"
Construction, surface workers'^
2.9
Involved
6.1
2.9
Noninvolved
3.3
1.6
Retrieval, surface workers^
2.9
Involved
3.0
1.2
Noninvolved
3.3
1.6
Retrieval, subsurface workers^
2.9
Involved
3.0
1.2
Noninvolved
3.3
1.6
a. FTE = full-time equivalent work years.
b. Source: Data Set 4, Section F.2.2.
c. Source: Data Set 1, Section F.2.2.
d. Source: Data Set 3, Section F.2.2.
F.4.2.3 Estimated Radiological Exposure Rates and Times for the Retrieval Contingency
DOE used the same set of worker exposure rates and exposure times as those used for evaluating
radiological worker impacts for the Proposed Action. Table F-47 presents the specific application of this
data to the retrieval contingency subphases. The source of the information is also referenced. The rates
used in the analysis did not take into account radioactive decay for the period between emplacement and
retrieval.
F-42
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-47. Radiological doses and
exposure data used to calculate worker exposures during retrieval."
Occupancy factor for
Annual dose
Full-time
Subperiod and
Source of
exposure rate (fraction
(millirem, except
equivalent
worker group
exposure
of 8-hour workday)
where noted)
workers'"
Source"
Construction
Surface
Involved
None
Noninvolved
None
Operations
Surface
Involved
Waste package
1.0
400
13
(1)
Radiation
100
16
(1)
Noninvolved
1.0
25
22
(2)
0
57
(2)
Subsurface
Involved
Waste package
I.O
Variable
—
(3)
Radon-222
1.0
Table F-4
(5), Table F-4
Drift ambient
1.0
40
(4), (5)
Noninvolved
Waste package
0.04 (0.4 for 10% of
workers)
0.1 millirem per
hour
(7)
Radon-222
0.4
Table F-4
(6), Table F-4
Drift ambient
0.4
40
(4), (6)
a. External exposures include radiation from spent nuclear fuel and high-level radioactive waste packages to surface and subsurface workers,
the ambient exposure to subsurface workers from naturally occurring radiation in the drift walls, and subsurface worker exposure from
inhalation of radon-222.
b. Number of full-time equivalent workers by dose category for surface faciUty activities.
C. Sources:
(I) Adapted from TRW (1999c, Table 6.2) for waste receipt, handling, and packaging operations. Values are based on dose rate
distribution (fractions) fix)m TRW (1999c, Table 6.2) for involved workers for dual-purpose canister scenario adjusted for fewer
workers for retrieval. Forty-five percent of 29 involved workers would be in the 400-milUrem-per-year category and 55 percent would
be in the lOO-millirem-per-year category.
Adapted from TRW (1999c, Table 6.2) for waste receipt, handling, and packaging operations. Values based on dose rate distribution
(fractions) from TRW (1999c, Table 6.2) for noninvolved workers for dual-purpose canister scenario adjusted for fewer workers for
retrieval. Twenty-eight percent of the 79 workers would be in the 25-milhrem-per-year category and 72 percent would be in the
0-rem-per-year category.
Table F-4.
Section F. 1.1. 6.
Rasmussen (1998a, all).
Rasmussen(l999, all).
Rasmussen (1998b, all).
(2)
(3)
(4)
(5)
(6)
(7)
FAS DETAILED RESULTS FOR THE RETRIEVAL CONTINGENCY
F.4.3.1 Construction Phase
F.4.3.1 .1 Human Health and Safety Impacts to Workers from Industrial Hazards
The construction phase would entail only surface-facility activities. Table F-48 summarizes health and
safety impacts to workers from industrial hazards during construction. There would be no radiological
sources present during surface facility construction activities for retrieval and, hence, no radiological
health and safety impacts to workers.
F.4.3.2 Operations Period
F.4.3.2.1 Health and Safety Impacts to Workers from Industrial Hazards
Chapter 4, Table 4-47, summarizes health and safety impacts to workers from industrial hazards
common to the workplace for the retrieval operations period. The impacts in that table consist of two
F-43
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-48. Industrial hazard health and safety impacts to
workers during construction.'
Worker group
Impacts
Involved
Full-time equivalent work years'"
1,130
Total recordable cases of injury and illness
69
Lost workday cases
33
Fatalities
0.03
Noninvolved
Full-time equivalent work years
430
Total recordable cases of injury and illness
14
Lost workday cases
7
Fatalities
0.0 1
All workers (totalsf
Full-time equivalent work years
1,560
Total recordable cases of injury and illness
83
Lost workday cases
40
Fatalities
0.05
a. Source: Impact rates from Table F-46.
b. Source: Table F-45.
components — health impacts to surface workers and health impacts to subsurface workers. Tables F-49
and F-50 list health impacts from industrial hazards during retrieval operations for surface and subsurface
workers, respectively.
Table F-49. Industrial hazard health and safety impacts to
surface facility workers during retrieval."
Worker group
Impacts
Involved
Full-time equivalent work years'"
320
Total recordable cases of injury and illness
10
Lost workday cases
4
Fatalities
0.009
Noninvolved
Full-time equivalent work years
870
Total recordable cases of injury and illness
29
Lost workday cases
14
Fatalities
0.03
All workers (totalsf
Full-time equivalent work years
1,190
Total recordable cases of injury and illness
37
Lost workday cases
18
Fatalities
0.03
a. Source: Impact rates from Table F-46.
b. Source: Table F-45.
c. Totals might differ from sums due to rounding.
F.4.3.2.2 Radiological Health and Safety Impacts to Workers
Potential radiological health impacts to workers during the operations period of retrieval consist of the
following components:
• Impacts to surface facility workers involved in handling the waste packages and placing them in
concrete storage units
F-44
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-50. Industrial hazard health and safety impacts to
subsurface facility workers during retrieval."
Worker group
Impacts
Involved
Full-time equivalent work years'"
810
Total recordable cases of injury and illness
24
Lost workday cases
11
Fatalities
0.02
Noninvolved
Full-time equivalent work years
180
Total recordable cases of injury and illness
6
Lost workday cases
3
Fatalities
0.01
All workers {totals f
Full-time equivalent work years
990
Total recordable cases of injury and illness
30
Lost workday cases
13
Fatalities
0.03
a. Source: Impact rates from Table F-46.
b. Source: Table F-45.
c. Totals might differ from sums due to rounding.
• Impacts to subsurface facilities workers from direct radiation emanating from the waste packages
• Impacts to subsurface workers from inhalation of radon-222 in the atmosphere of the drifts
• Impacts to subsurface workers from ambient radiation from naturally occurring radionuclides in the
drift walls
Tables F-51 and F-52 list potential radiological health impacts for each of these component parts. The
impacts to subsurface workers only vary slightly (less than 2 percent) with thermal load and are highest
for the low thermal load. Thus, the values in Table F-52 for the low thermal load case, would produce the
largest impacts.
Table F-51. Radiological health impacts to surface facility workers from waste
handling during retrieval."
Worker group Impacts
Involved
Full-time equivalent work years'" 320
Maximally exposed individual dose (millirem) 4,400
Latent cancer fatality probability for maximally exposed individual 0.002
Collective dose (person-rem) 75
Latent cancer fatality incidence for overall worker group 0.03
Noninvolved
Full-time equivalent work years 870
Maximally exposed individual dose (millirem) 280
Latent cancer fatality probability for maximally exposed individual 0.0001
Collective dose (person-rem) 6
Latent cancer fatality incidence for overall worker group 0.002
All workers (totals f
Full-time equivalent work years 1,190
Collective dose (person-rem) 8 1
Latent cancer fatality 0.03
a. Source: Exposure rate data from Table F-47.
b. Source: Table F-45.
c. Totals might differ from sums due to rounding.
F-45
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Table F-52. Components of radiological health impacts to subsurface workers
during retrieval for the low thermal load scenario.
a,b
Source of
exposure
Waste
Radon-222
Group
packages
Ambient
inhalation
Total'
Involved
Full-time equivalent work years'"
840
840
840
840
Maximally exposed individual dose
4,400
440
2,110
6,950
(millirem)
Latent cancer fatality probability for
0.002
0.0002
0.0008
0.003
maximally exposed individual
Collective dose (person-rem)
200
33
160
390
Latent cancer fatality incidence for
0.08
0.01
0.06
0.16
overall worker group
Noninvolved
Full-time equivalent work years
180
180
180
180
Maximally exposed individual dose
88
220
1,060
1,370
(millirem)
Latent cancer fatality probability for
0.00004
0.00009
0.0004
0.000
maximally exposed individual
5
Collective dose (person-rem)
1
4
17
22
Latent cancer fatality incidence for
0.0004
0.001
0.007
0.009
overall worker group
All workers (totals f
Full-time equivalent work years
1,010
1,010
1,010
1,010
Collective dose (person-rem)
200
37
180
420
Latent cancer fatality incidence for
0.08
0.01
0.07
0.17
overall worker group
a. Source: Exposure data from Table F-47.
b. The variation in values among the thermal load scenarios was small. Therefore, only the
largest values (for the low thermal load) are listed.
c. Totals might differ from sums due to rounding.
d. Source: Table F-45.
REFERENCES
DOE 1995
DOE 1998a
DOE 1998b
DOE 1999
DOE (U.S. Department of Energy), 1995, YMP Erionite Control
Protocol, Office of Civilian Radioactive Waste Management, Yucca
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with DOE Order 440.1 Occupational Exposure Assessment, DOE G 440.1-
3, Department of Energy, Office of Worker Health and Safety,
Washington, D.C. [240305]
DOE (U.S. Department of Energy), 1998b, Air Quality Control Design
Analysis, BCADOOOOO-01717-0200-00008, Revision 00, Office of
Civilian Radioactive Waste Management, Washington, D.C.
[MOL. 19980729.0044]
DOE (U.S. Department of Energy), 1999, "CAIRS Database, DOE and
Contractor Injury and Illness Experience by Operation Type by Year and
Quarter, 1993 through 1998," http://tis.eh.doe.gov/cairs/cairs/dataqtr/
q984a.htm, May 22, Washington, D.C. [244036]
F-46
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Eckerman and Ryman 1993
Eckerman, Wolbarst, and
Richardson 1988
EPA 1996
Gotchy 1987
lARC 1987
lARC 1997
ICRP 1977
ICRP 1991
ICRP 1994
lessen 1999
Eckerman, K. P., and J. C. Ryman, 1993, External Exposure to
Radionuclides in Air, Water, and Soil, Exposure-to-Dose Coefficients for
General Application, Based on the 1987 Federal Radiation Protection
Guidance: Federal Guidance Report No. 12, EPA 402-R-93-081, Office
of Radiation and Indoor Air, U.S. Environmental Protection Agency,
Washington D.C. [225472]
Eckerman, K. P., A. B. Wolbarst, and A. C. B. Richardson, 1988, Limiting
Values of Radionuclide Intake and Air Concentration and Dose
Conversion Factors for Inhalation, Submersion, and Ingestion, Pederal
Guidance Report No. 1 1, EPA-520/1-88-020, U.S. Environmental
Protection Agency, Office of Radiation Programs, Oak Ridge National
Laboratory, Oak Ridge, Tennessee. [203350]
EPA (U.S. Environmental Protection Agency), 1996, Ambient Levels and
Noncancer Health Effects of Inhaled Crystalline and Amorphous Silica:
Health Issue Assessment, EPA/600/R-95/1 15, National Center for
Environmental Assessment, Office of Research and Development,
Washington, D.C. [243562]
Gotchy, R. L., 1987, Potential Health and Environmental Impacts
Attributable to the Nuclear and Coal Fuel Cycles, NUREG-0332, Office of
Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, D.C. [234603]
LARC (International Agency for Research on Cancer), 1987, Silica and
Some Silicates, World Health Organization, United Nations, Lyon,
Prance. [226502]
lARC (International Agency for Research on Cancer), 1997, lARC
Monographs on the Evaluation of Carcinogenic Risks to Humans, Silica,
Some Silicates, Coal Dust and p&ra-Aramid Fibrils, Volume 68, LARC
Working Group on the Evaluation of Carcinogenic Risks to Humans,
World Health Organization, United Nations, Lyon, France. [236833]
ICRP (International Commission on Radiological Protection), 1977,
Recommendations of the International Commission on Radiological
Protection, ICRP Publication 26, Pergamon Press, Elmsford, New York.
[221568]
ICRP (International Commission on Radiological Protection), 1991, 7990
Recommendations of the International Commission on Radiological
Protection, Publication 60, Volume 21, Numbers 1-3, Pergamon Press,
Elmsford, New York. [235864]
ICRP (International Commission on Radiological Protection), 1994,
Protection Against Radon-222 at Home and at Work, Publication 65,
Pergamon Press, Oxford, Great Britain. [236754]
lessen, J., 1999, "Pinal Closure Phase Years based on March 99 EP's,"
electronic communication to Dcenberry et al., Jason Technologies
Corporation, Las Vegas, Nevada. [MOL. 19990526.0030]
F-47
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Kamrin 1988
Maheras and Thome 1993
McKenzie 1998
Mettler and Upton 1995
MSHA 1999
NCRP 1987
NCRP 1993
NCRP 1996
NIOSH 1996
NJDHSS 1996
Rasmussen 1998a
Kamrin, M. A., 1988, Toxicology - A Primer on Toxicology Principles
and Applications: Indoor & Outdoor Air, Drinking Water, Food,
Workplace Environment, Lewis Publishers, Inc., Chelsea, Michigan.
[243888]
Maheras, S. J., and D. J. Thome, 1993, New Production Reactor
Exposure Pathways at the Idaho National Engineering Laboratory, NPR-
8957, EG&G Idaho, hic, Idaho Falls, Idaho. [243737]
McKenzie, D., 1998, "Erionite Encounters in Expanded Layouts,"
electronic mail to D. Walker (Jason Technologies Corporation),
December 21, Morrison Knudsen Corporation, Las Vegas, Nevada.
[MOL. 199905 11.0294]
Mettler, F. A., Jr., and A. C. Upton, 1995, Medical Effects of Ionizing
Radiation, Second Edition, W. B. Saunders Company, Philadelphia,
Pennsylvania. [244122]
MSHA (Mine Safety and Health Administration), 1999, "Table 6. -
Number of Contractor Injuries, Injury-Incidence Rates, Average Numbers
of Employees, and Employee Hours, by Work Location and Mineral
Industry," http://www.msha.gov.stats/wq964+06.htm, March 11,
Washington, D.C. [243568]
NCRP (National Council on Radiation Protection and Measurements),
1987, Ionizing Radiation Exposure of the Population of the United States:
Recommendations of the National Council on Radiation Protection and
Measurements, Report No. 93, Bethesda, Maryland. [229033]
NCRP (National Council on Radiation Protection and Measurements),
1993, Risk Estimates for Radiation Protection, Report No. 115, Bethesda,
Maryland. [232971]
NCRP (National Council on Radiation Protection and Measurements),
1996, Screening Models for Releases of Radionuclides to Atmosphere,
Surface Water, and Ground, Recommendations of the National Council
on Radiation Protection and Measurements, Report No. 123, Bethesda,
Maryland. [225158, Volume 1; 234986, Volume 2]
NIOSH (National Institute for Occupational Safety and Health), 1996,
"Silica, crystalline (as respirable dust), DDLH Documentation"
(downloaded from http://www.cdc.gov/niosh/idlh/14808607.html, April
8, 1999), Washington, D.C. [243424]
NJDHSS (New Jersey Department of Health and Senior Services), 1996,
"Hazardous Substance Fact Sheet - Silica, Cristobalite," Trenton, New
Jersey. [243425]
Rasmussen, D. G., 1998a, "Radiation exposure information," electronic
communication with J. lessen (Jason Technologies Corporation), July 22,
TRW Environmental Safety Systems Inc., Las Vegas, Nevada.
[MOL. 19990526.0029]
F-48
Human Health Impacts Primer and Details for Estimating Health
Impacts to Workers from Yucca Mountain Repository Operations
Rasmussen 1998b
Rasmussen 1999
Stewart 1998
Technical Resources 1994
TRW 1999a
TRW 1999b
TRW 1999c
Rasmussen, D. G., 1998b, "Radiation exposure information," electronic
communication with R. Orthen (Jason Technologies Corporation), July
29, TRW Environmental Safety Systems Inc., Las Vegas, Nevada.
[MOL. 199905 11.0386]
Rasmussen, D., 1999, "Additional matrix," electronic communication
with attachment to D. Walker (Jason Technologies Corporation), April
16, TRW Environmental Safety Systems Inc., Las Vegas, Nevada.
[MOL. 19990602.01 80]
Stewart B., 1998, "YMP EIS Information Request - CAIRS Statistics for
Construction and Non-Construction Activities During TBM Operations,"
electronic communication with attachment to J. Steinhoff (TRW
Environmental Safety Systems Inc.), December 17, Las Vegas, Nevada.
[MOL. 199905 11.0298]
Technical Resources, Inc., 1994, Seventh Annual Report on Carcinogens
1994, Rockville, Maryland. [243694]
TRW (TRW Environmental Safety Systems Inc.), 1999a, Environmental
Baseline File for Human Health, BOOOOOOOO-01717-5705-00114,
Revision 01, Las Vegas, Nevada. [MOL. 19990608.0035]
TRW (TRW Environmental Safety Systems Inc.), 1999b, Engineering
File - Subsurface Repository, BCAOOOOOO-017 17-5705-00005,
Revision 02 with DCNl, Las Vegas, Nevada. [MOL. 19990622.0202,
document; MOL. 19990621.0157, DCNl]
TRW (TRW Environmental Safety Systems Inc.), 1999c, Repository
Surface Design Engineering Files Report, BCBOOOOOO-017 17-5705-
00009, Revision 03, Las Vegas, Nevada. [MOL. 19990615.0238]
F-49
Appendix G
Air Quality
Air Quality
TABLE OF CONTENTS
Section Page
G.l Nonradiological Air Quality G-1
G.1.1 Computer Modeling and Analysis G-3
G.1.2 Locations of Hypothetically Exposed Individuals G-4
G.1.3 Meteorological Data and Reference Concentrations G-4
G.1.4 Construction Phase G-6
G.1.4.1 Fugitive Dust Emissions from Surface Construction G-6
G.1.4.2 Fugitive Dust from Subsurface Excavation G-8
G.1.4.3 Fugitive Dust from Excavated Rock Pile G-9
G. 1.4.4 Fugitive Dust from Concrete Batch Facility G-11
G. 1.4.5 Exhaust Emissions from Construction Equipment G-12
G. 1.4.6 Exhaust from Boiler G-14
G.1.5 Operation and Monitoring Phase G-15
G.1.5.1 Fugitive Dust from Concrete Batch Facility G-16
G.1.5.2 Fugitive Dust from Subsurface Excavation G-16
G.1.5.3 Fugitive Dust from Excavated Rock Pile G-16
G.1.5.4 Exhaust from Excavated Rock Pile Maintenance Equipment G-18
G.1.5.5 Exhaust from Boiler G-I9
G.1.6 Closure Phase G-20
G.1.6.1 Dust from Backfill Plant G-20
G.l. 6.2 Fugitive Dust from Concrete Batch Facility G-21
G.1.6.3 Fugitive Dust from Closure Activities G-22
G.1.6.4 Fugitive Dust from Excavated Rock Pile G-22
G.l. 6.5 Exhaust Emissions from Surface Equipment G-24
G.1.7 Retrieval Scenario G-25
G.1.7.1 Fugitive Dust from Construction of Retrieval Storage Facility G-26
G.1.7.2 Exhaust from Construction Equipment G-26
G.2 Radiological Air Quality G-27
G.2.1 Locations of Hypothetically Exposed Individuals and Populations G-28
G.2.2 Meteorological Data and Atmospheric Dispersion Factors G-30
G.2.3 Radiological Source Terms G-31
G.2.3. 1 Release of Radon-222 and Radon Decay Products from the Subsurface Facility G-3 1
G.2.3.2 Release of Radioactive Noble Gases from the Surface Facility G-35
G.2.4 Dose Calculation Methodology G-36
G.2.4.1 Dose to the Public G-36
G.2.4.2 Dose to Noninvolved Workers G-37
References G-38
G-iii
Air Quality
LIST OF TABLES
Table Page i
G-1 Criteria pollutants and regulatory limits G-2 I
G-2 Distance to the nearest point of unrestricted public access G-4 !
G-3 Unit release concentrations and direction to maximally exposed individual
location for 1 1 combinations of 4 release periods and 5 regulatory limit averaging
times G-5
G-4 Land area disturbed during the construction phase for each thermal load scenario G-7
G-5 Fugitive dust releases from surface construction G-7
G-6 Estimated fugitive dust air quality impacts from surface construction G-8
G-7 Fugitive dust releases from excavation activities G-8
G-8 Fugitive dust and cristobalite air quality impacts from excavation activities G-8
G-9 Active area of excavated rock pile during the construction phase G-10
G-10 Fugitive dust released from the excavated rock pile during the construction phase G-10
G- 1 1 Fugitive dust and cristobalite air quality impacts from the excavated rock pile A
during the construction phase G-11 «
G-12 Dust release rates for the concrete batch facility G-1 1
G-1 3 Dust release rates for the concrete batch facility during the operation and
monitoring phase G-12 —
G-14 Particulate matter air quality impacts from the concrete batch facility during the M^
construction phase G-12
G-15 Pollutant emission rates for construction equipment G-12
G-16 Amount of fuel consumed per year during the construction phase G-13
G-1 7 Pollutant release rates from surface equipment during the construction phase G-13
G- 1 8 Air quality impacts from construction equipment during the construction phase G-14
G- 1 9 Annual pollutant release rates for the South Portal Operations Area boiler G-14
G-20 Pollutant release rates from the boiler during the construction phase G-15
G-2 1 Air quality impacts from boiler pollutant releases from the South Portal
Operations Area during the construction phase G-15
G-22 Estimated active excavated rock pile area during subsurface excavation activities W
during the operation and monitoring phase G-16
G-23 Fugitive dust release rate from the excavated rock pile during the operation and
monitoring phase G-17
G-24 Fugitive dust and cristobalite air quality impacts from the excavated rock pile
during the operation and monitoring phase G-17
G-25 Annual amount of fuel consumed during the operation and monitoring phase G-1 8
G-26 Pollutant release rates from surface equipment during the operation and
monitoring phase G-18 ^
G-27 Air quality impacts from surface equipment during the operation and monitoring ^
phase G-1 9
G-28 Air quality impacts from boiler pollutant releases from both North and South
Portal Operations Areas G-20
G-29 Emission rates from a crushed stone processing plant G-21
G-30 Dust release rates from the backfill plant G-21
G-31 Particulate matter air quality impacts from backfill plant G-21
G-32 Dust release rates from the concrete batch facility during the closure phase G-22
G-33 Particulate matter air quality impacts from the concrete batch facility during the
closure phase G-22
G-34 Active excavated rock pile area during the closure phase G-22
G-35 Fugitive dust release rates from the excavated rock pile during the closure phase G-23
G-iv
Air Quality
Table Page
G-36 Fugitive dust and cristobalite air quality impacts from the excavated rock pile
during the closure phase G-23
G-37 Annual amount of fuel consumed during the closure phase G-24
G-38 Pollutant release rates from surface equipment during the closure phase G-24
G-39 Air quality impacts from surface construction equipment during the closure phase G-25
G-40 Fugitive dust release rates from surface construction of retrieval storage facility
and storage pad G-26
G-41 Fugitive dust air quality impacts from surface construction of the retrieval
storage facility and storage pad G-26
G-42 Pollutant release rates from surface equipment during the retrieval scenario G-27
G-43 Air quality impacts from surface equipment during the retrieval scenario G-27
G-44 Projected year 2000 population distribution within 80 kilometers of repository
site G-28
G-45 Noninvolved (surface) worker population distribution for Yucca Mountain
activities G-29
G-46 Distribution of repository subsurface exhaust ventilation air G-31
G-47 Atmospheric dispersion factors for potentially exposed individuals and
populations from releases at the repository site G-32
G-48 Estimated radon-222 releases for repository activities for the Proposed Action
inventory G-33
G-49 Estimated radon-222 releases for repository activities for Inventory Modules 1 or
2 G-35
G-50 Krypton-85 releases from surface facility handling activities for commercial
spent nuclear fuel during the operation and monitoring phase G-36
G-5 1 Factors for estimating dose to the public and noninvolved workers per
concentration of radionuclide in air for krypton-85 and radon-222 G-37
G-v
I
Air Quality
APPENDIX G. AIR QUALITY
Potential releases of nonradiological and radiological pollutants associated with the construction,
operation and monitoring, and closure of the proposed Yucca Mountain Repository could affect the air
quality in the surrounding region. This appendix discusses the methods and additional data and
intermediate results that the U.S. Department of Energy (DOE) used to estimate impacts from potential
releases to air. Final results are presented in Chapter 4, Section 4.1.2, and Chapter 8, Section 8.2.2.
Nonradiological pollutants can be categorized as hazardous and toxic air pollutants, criteria pollutants, or
other substances of particular interest. Repository activities would cause the release of no or very small
quantities of hazardous and toxic pollutants; therefore, these pollutants were not considered in the
analysis. Concentrations of six criteria pollutants are regulated under the National Ambient Air Quality
Standards (40 CFR Part 50) established by the Clean Air Act. This analysis evaluated releases and
potential impacts of four of these pollutants — carbon monoxide, nitrogen dioxide, sulfur dioxide, and
particulate matter with an aerodynamic diameter of 10 micrometers or less (PMio) — quantitatively. It
addresses the other two criteria pollutants — lead and ozone — and the concentration of particulate matter
with an aerodynamic diameter of 2.5 micrometers or less (PM25), qualitatively. In addition, this analysis
considers potential releases to air of cristobalite, a form of crystalline silica that can cause silicosis and is
a potential carcinogen. These pollutants could be released during all project phases. Section G.l
describes the methods DOE used to calculate impacts from releases of criteria pollutants and cristobalite.
Radionuclides that repository-related activities could release to the atmosphere include the noble gas
krypton-85 from spent nuclear fuel handling during the operation and monitoring phase, and naturally
occurring radon-222 and its decay products from ventilation of the subsurface facility during all project
phases. Other radionuclides would not be released or would be released in such small quantities they
would result in very small impacts to air quality. Such radionuclides are not discussed further in this
appendix. Section G.2 describes the methods DOE used to calculate impacts of radionuclide releases.
G.1 Nonradiological Air Quality
This section describes the methods DOE used to analyze potential impacts to air quality at the proposed
Yucca Mountain Repository from releases of nonradiological air pollutants during the construction,
operation and monitoring, and closure phases, and a retrieval scenario. It also describes intermediate
results for various repository activities. Table G-1 lists the six criteria pollutants regulated under the
National Ambient Air Quality Standards or the Nevada Administrative Code along with their regulatory
limits and the periods over which pollutant concentrations are averaged. The criteria pollutants addressed
quantitatively in this section are nitrogen dioxide, sulfur dioxide, particulate matter 10 micrometers or less
in aerodynamic diameter (PMio), and carbon monoxide. Lead was not considered further in this analysis
because there would be no airborne sources at the repository. Particulate matter 2.5 micrometers or less
in aerodynamic diameter (PM2.5) and ozone are discussed below, as is cristobalite, a mineral occurring
naturally in the subsurface rock at Yucca Mountain.
The U.S. Environmental Protection Agency revised the primary and secondary standards for particulate
matter in 1997 (62 FR 38652, July 18, 1997), establishing annual and 24-hour PM25 standards at 15
micrograms per cubic meter and 65 micrograms per cubic meter, respectively. Primary standards set
limits to protect public health, including the health of "sensitive" populations. Secondary standards set
limits to protect public welfare, including protection against decreased visibility, damage to animals,
crops, vegetation, and buildings. Because the new particulate standard will regulate PM2 5 for the first
time, the agency has allowed 5 years for the creation of a national monitoring network and the analysis of
collected data to help develop state implementation plans. The new PM2 5 standards have not been
implemented and the imposition of local area controls will not be required until 2(X)5. By definition,
PM2.5 levels can be no more than, and in the real world are always substantially less than, PMio levels. In
G-1
Air Quality
Table G-1. Criteria
polli
jtants and regulatory limits.
Regulatory
limit"
Micrograms per
Pollutant
Period Parts per million
cubic meter
Nitrogen dioxide
Annual 0.053
100
Sulfur dioxide
Annual 0.03
80
24-hour 0.14
365
3-hour 0.50
1,300
Carbon monoxide
8-hour 9
10,000
1-hour 35
40,000
PM,o
Annual
50
24-hour
150
PNlz.,"
Annual
15
24-hour
65
Ozone
8-hour 0.08
157
1-hour 0.12'
235
Lead
Quarterly
1.5
a. Sources: 40 CFR 50.4 through 50. 1 1 and Nevada Administrative Code 445B.391 .
b. Standard not yet implemented.
c. The 1 -hour standard does not apply to Nevada because the State was in
attainment when the 8-hour standard was adopted in July 1997.
general, PM2.5 levels would be approximately one-third of the PMio levels. As the analysis for PMio
shows, even the maximum PMio levels that could be generated by the Proposed Action are substantially
below the PM2.5 standards. Thus, although no detailed PM2 5 analysis has been conducted, the PMio
analysis can be regarded as a surrogate for a PM2.5 analysis and illustrates that potential PM2.5 levels
would be well below applicable regulatory standards.
The purpose of the ozone standard is to control the ambient concentration of ground-level ozone, not
naturally occurring ozone in the upper atmosphere. Ozone is not emitted directly into the air; rather, it is
formed when volatile organic compounds react in the presence of sunlight. Nitrogen dioxides are also
important precursors to ozone. Small quantities of volatile organic compounds would be released from
repository activities; the peak annual release would be about 540 kilograms (1,200 pounds) (TRW 1999a,
Table 6-2, page 75). Because Yucca Mountain is in an attainment area for ozone, the analysis compared
the estimated annual release to the Prevention of Significant Deterioration of Air Quality emission
threshold for volatile organic compounds from stationary sources (40 CFR 52.21). The volatile organic
compound emission threshold is 35,000 kilograms (77,000 pounds) per year, so the peak annual release
from the repository would be well below this level. Accordingly, the analysis did not address volatile
organic compounds and ozone further, although this does not preclude future, more detailed analyses if
estimates of volatile organic compound emissions change.
Cristobalite, one of several naturally occurring crystalline forms of silica (silicon dioxide), is a major
mineral constituent of Yucca Mountain tuffs (TRW 1999b, page 4-81). Prolonged high exposure to
crystalline silica can cause silicosis, a disease characterized by scarring of lung tissue. An increased
cancer risk to humans who already have developed adverse noncancer effects from silicosis has been
shown, but the cancer risk to otherwise healthy individuals is not clear (EPA 1996, page 1-5).
Cristobalite is principally a concern for involved workers because it could be inhaled during subsurface
excavation operations. Appendix F, Section F.l, contains additional information on crystalline silica.
While there are no limits for exposure of the general public to cristobalite, there are limits to workers for
exposure (29 CFR 1(X)0.1910). Therefore, this analysis used a comparative benchmark of 10 micrograms
per cubic meter, based on a cumulative lifetime exposure of 1,000 micrograms per cubic meter multiplied
by years (that is, the average annual exposure concentration times the number of years exposed). At this
level, an Environmental Protection Agency health assessment (EPA 1996, pages 1-5 and 7-5) states that
G-2
Air Quality
there is a less than 1 percent chance of silicosis. Over a 70-year lifetime, this cumulative exposure
benchmark would correspond to an annual average exposure concentration of about 14 micrograms per
cubic meter, which was rounded down to 10 micrograms per cubic meter to establish the benchmark.
Cristobalite would be emitted from the subsurface in exhaust ventilation air during excavation operations
and would be released as fugitive dust from the excavated rock pile, so members of the public and
noninvolved workers could be exposed. Fugitive dust from the excavated rock pile would be the largest
potential source of cristobalite exposure to the public. The analysis assumed that 28 percent of the
fugitive dust released from this rock pile and from subsurface excavation would be cristobalite, reflecting
the cristobalite content of the parent rock, which ranges from 18 to 28 percent (TRW 1999b, page 4-81).
Using the parent rock percentage probably overestimates the airborne cristobalite concentration, because
studies of both ambient and occupational airborne crystalline silica have shown that most of this airborne
material is coarse and not respirable and that larger particles will deposit rapidly on the surface (EPA
1996, page 3-26).
G.1.1 COMPUTER MODELING AND ANALYSIS
DOE used the Industrial Source Complex computer program to estimate the annual and short-term
(24-hour or less) air quality impacts at the proposed Yucca Mountain Repository. The Department has
used this program in recent EISs (DOE 1995, all; 1997a,b, all) to estimate nonradiological air quality
impacts. The program contains both a short-term model (which uses hourly meteorological data) and a
long-term model (which uses joint frequency meteorological data). The program uses steady-state
Gaussian plume models to estimate pollutant concentrations from a variety of sources associated with
industrial complexes (EPA 1995a, all). This modeling approach assumes that (1) the time-averaged
pollutant concentration profiles at any distance downwind of the release point may be represented by a
Gaussian (normal) distribution in both the horizontal and vertical directions; and (2) the meteorological
conditions are constant (persistent) over the time of transport from source to receptor. The Industrial
Source Complex program is appropriate for either flat or rolling terrain, and for either urban or rural
environments. The Environmental Protection Agency has approved this program for specific regulatory
applications. Input requirements for the program include source configuration and pollutant emission
parameters. The short-term model was used in this analysis to estimate all nonradiological air quality
impacts and uses hourly meteorological data that include wind speed, wind direction, and stability class to
compute pollutant transport and dispersion.
Because the short-term pollutant concenfrations were based on annual usage or release parameters,
conversion of annual parameter values to short-term values depended on the duration of the activity.
Many of the repository activities were assumed to have a schedule of 250 working days per year, so the
daily release would be the annual value divided by 250.
In many cases, site- or activity-specific information was not available for estimating pollutant emissions
at the Yucca Mountain site. In these cases, generic information was used and conservative assumptions
were made that tended to overestimate actual air concentrations.
As noted in Section G.l, the total nonradiological air quality impacts are described in Chapter 4, Section
4.1.2, for the Proposed Action and in Chapter 8, Section 8.2.2, for the inventory modules. These impacts
are the sum of air quality impacts from individual sources and activities that take place during each of the
project phases and that are discussed later in this section (for example, dust emissions from the concrete
batch facility during the construction phase). The maximum air quality impact (that is, air concentration)
resulting from individual sources or activities could occur at different land withdrawal area boundary
locations depending on the release period and the regulatory averaging time (see Section G. 1.3). These
maximums generally occur in a westerly or southerly direction. The total nonradiological air quality
impacts presented in Sections 4.1.2 and 8.2.2 are the sum of the calculated maximum concentrations
regardless of direction. Therefore, the values presented would be larger than the actual sum of the
G-3
Air Quality
concentrations for a particular distance and direction. This approach was selected to simplify the
presentation of air quality results.
G.I .2 LOCATIONS OF HYPOTHETIC ALLY EXPOSED INDIVIDUALS
The location of the public maximally exposed individual was determined by calculating the maximum
ground-level pollutant concentrations. Because unrestricted public access would be limited to the site
boundary, the analysis assumed that a hypothetical individual would be present at one point on the site
boundary during the entire averaging time of the regulatory limit (Table G-1).
Table G-2 lists the distances from the North and South Portals to the land withdrawal area boundary
where the analysis assumed members of the public would be present. The table does not list all directions
because the land withdrawal area boundaries would not be accessible to members of the public in some
directions (restricted access areas of the Nevada Test Site and Nellis Air Force Range). The distance to
the nearest unrestricted public access in these directions would be so large that there would be no air
quality impacts. For the east to south-southeast directions, the distances to the land withdrawal area
boundary would be large, but the terrain is such that plumes traveling in these directions tend to enter
Fortymile Wash and turn south. The analysis used the distance to the south land withdrawal area
boundary for those sectors.
Table G-2. Distance to the nearest point of unrestricted public
access (kilometers)."'''^
From North
From South
Direction
Portal
Portal
Northwest
14
15
West-northwest
12
12
West
11
11
West-southwest
14
12
Southwest
18
16
South-southwest
23
19
South
21
18
South-southeast''
21
18
Southeast''
21
18
a. Source: DOE (1997c, all).
b. Numbers are rounded to two significant figures.
c. To convert kilometers to miles, multiply by 0.6217.
d. Distances assumed to be the same as those to the south.
G.1.3 METEOROLOGICAL DATA AND REFERENCE CONCENTRATIONS
DOE estimated the concentrations of criteria pollutants in the region of the repository by using the
Industrial Source Complex program and site-specific meteorological data for 1993 to 1997 from air
quality and meteorology monitoring Site 1 (TRW 1999c, electronic addendum). Site 1 is less than 1
kilometer (0.6 mile) south of the proposed North Portal surface facility location. Similar topographic
exposure leads to similar prevailing northerly and southerly winds at both locations. DOE used Site 1
data because an analysis of the data collected at all the sites showed that site to be most representative of
the surface facilities (TRW 1999c, page 7). Wind speed data are from the 10-meter (33-foot) level, as are
atmospheric stability data, using the night-adjusted sigma-theta method (EPA 1987, pages 6-20 to 6-32).
Mixing height measurements were not available for Yucca Mountain so the analysis assumed a mixing
height of approximately 140 meters (470 feet), which is one-tenth of the 1,420 meters (4,700 feet)
mixing-layer depth for Desert Rock, Nevada. Desert Rock is the nearest upper air meteorological station,
about 44 kilometers (27 miles) east-southeast near Mercury, Nevada. The average mixing height at
Desert Rock was divided by 10 to simulate the mixing height during very stable conditions, which is
when the highest concentrations from a ground-level source would normally occur. All nonradiological
G-4
Air Quality
pollutant releases were assumed to come from ground-level point sources. Both of these conservative
assumptions, made because of a lack of site-specific information, tend to overestimate actual air
concentrations. Fugitive dust emissions could be modeled as an area source, but the distance from the
source to the exposure location would be large [more than 10 kilometers (6 miles)] so a point source
provides a good approximation. Some sources would have plume rise, such as boiler emissions, but this
was not considered because there is inadequate information to characterize the rise.
The analysis estimated unit release concentrations at the land withdrawal area boundary points of
maximum exposure for ground-level point-source releases. The concentrations were based on release
rates of 1 gram (0.04 ounce) per second for each of the five regulatory limit averaging times (annual,
24-hour, 8-hour, 3-hour, or 1-hour). Various activities at the Yucca Mountain site could result in
pollutants being released over four different periods in a 24-hour day [continuously, 8-hour, 12-hour (two
6-hour periods), or 3-hour]. Eleven combinations of release periods and regulatory limit averaging times
would be applicable to activities at the Yucca Mountain site.
The analysis assumed that the 8-hour pollutant releases would occur from 8 a.m. to 4 p.m. and to be zero
for all other hours of the day. Similarly, it assumed that the 3-hour releases would occur from 9 a.m. to
12 p.m. and to be zero for all other hours. The 12-hour release would occur over two 6-hour periods,
assumed to be from 9 a.m. to 3 p.m. and from 5 p.m. to 1 1 p.m.; other hours would have zero release.
Continuous releases would occur throughout the 24-hour day. The estimates of all annual-average
concentrations assumed the releases were continuous over the year.
Table G-3 lists the maximum unit release concentrations for the 1 1 combinations of the Yucca Mountain
site-specific release periods and regulatory limit averaging times. The analysis estimated the unit
Table G-3. Unit release concentrations (micrograms per cubic meter based on a release of 1 gram per
second) and direction to maximally exposed individual location for 1 1 combinations of 4 release periods
and 5 regulatory limit averaging times."
Direction from South Unit release Direction from North Unit release
Portal Operations area concentration Portal Operations Area concentration
Continuous release - annual average concentration (1995)
South-southeast 0.12
Continuous release - 24-hour average concentration (1993)
Southeast 1 .0
Continuous release - 8-hour average concentration (1995)
Southeast 3.0
Continuous release - 3-hour average concentration (1995)
West 6.1
Continuous release - 1-hour average concentration (1995)
West 18
8-hour release (8 a.m. to 4 p.m.) - 24-hour average concentration (1997)
West-southwest 0.19
8-hour release (8 a.m. to 4 p.m.) - 8-hour average concentration (1997)
West-southwest 0.57
8-hour release (8 a.m. to 4 p.m.) - 3-hour average concentration (1997)
West-southwest 1 .5
8-hour release (8 a.m. to 4 p.m.)- 1-hour average concentration (1997)
West-northwest 3.3
12-hour release (9 a.m. to 3 p.m. and 5 p.m. to 1 1 p.m.) - 24-hour average concentration (1997)
West 0.95
3-hour release (9 a.m. to 12 p.m.) - 24-hour average concentration (1997)
West-northwest 0.17
a. Numbers are rounded to two significant figures.
b. Number in parentheses is the year from 1993 through 1997 for which meteorological data would result in the highest unit
concentration.
South-southeast
0.099
West
0.95
Southeast
2.5
West
6.1
West
18
West-northwest
0.18
West-northwest
0.52
West-northwest
1.4
West-northwest
3.3
concentration (1997)
West
0.95
West-northwest
0.17
G-5
Air Quality
concentrations and directions using the meteorological data during a single year from 1993 through 1997
(TRW 1999c, electronic addendum) that would result in the highest unit concentration. For all years, the
unit release concentrations for a particular averaging time are within a factor of 2 of each other. Table
G-3 lists the 24-hour averaged concentration for the 3- and 12-hour release scenarios because the
activities associated with these scenarios would only release PMio, which has annual and 24-hour
regulatory limits. The estimated concentration at the point of exposure was calculated by multiplying the
estimated source release rate (presented for each source in the following sections) by the maximum unit
release concentration for that averaging period.
G.1.4 CONSTRUCTION PHASE
This section describes the method used to estimate air quality impacts dming the 5-year construction
phase. DOE would complete the surface facilities during the construction phase, as well as sufficient
excavation of the subsurface to support initial emplacement activities.
This analysis used calculations of the pollutant concentrations from various construction activities to
determine air quality impacts. To calculate these impacts, estimated pollutant emission rates discussed in
this section were multiplied by the unit release concentration (see Section G.1.3). This produced the
pollutant concentration for comparison to regulatory limits. Short-term pollutant emission rates and
concentrations were estimated using the method described in Section G.1.1.
The principal emission sources of particulates would be fugitive dust from construction activities on the
surface, excavation of rock from the repository, storage of material on the excavated rock pile, and dust
emissions from the concrete batch facility. The principal sources of nitrogen dioxide, sulfur dioxide, and
carbon monoxide would be fuel combustion in trucks, cranes, and graders and emissions from a boiler in
the South Portal Operations Area. Nitrogen dioxide, sulfur dioxide, and carbon monoxide would also be
emitted during maintenance of the excavated rock pile. The following sections describe these sources in
more detail.
G. 1.4.1 Fugitive Dust Emissions from Surface Construction
Fugitive dust would be generated during such construction activities as earth moving and truck traffic.
All surface construction activities and associated fugitive dust releases were assumed to occur during
250 working days per year with one 8-hour shift per day. The preferred method suggested by the
Environmental Protection Agency would be to break the construction activities into component activities
(for example, earth moving, truck traffic) and calculate the emissions for each component. However,
detailed information was not available for the construction phase, so a generic, conservative approach was
taken. The release rate of total susp)ended particulates (particulates with aerodynamic diameters of 30
micrometers or less) was estimated as 0.27 kilogram per square meter (1.2 tons per acre) per month (EPA
1995b, pages 13.2.3-1 to 13.2.3-7). This estimated emission rate for total suspended particulates was
based on measurements made during the construction of apartments and shopping centers.
The amount of PMio (the pollutant of interest) emitted from the construction of the Yucca Mountain
Repository probably would be less than 0.27 kilogram per square meter (1.2 tons per acre) per month
because many of the particulates suspended during construction would be at the larger end of the
30-micrometer range and would tend to settle rapidly (Seinfeld 1986, pages 26 to 31). Experiments on
dust suspension due to construction found that at 50 meters (160 feet) downwind of the source, a
maximum of 30 percent of the remaining suspended particulates at respirable height were in the PMio
range (EPA 1988, pages 22 to 26). Based on this factor, only 30 percent of the 0.27 kilogram per square
meter per month of total suspended particulates, or 0.081 kilogram per square meter (0.36 ton per acre)
per month, would be emitted as PMio from construction activities. Because the default emission rate was
based on continuous emissions over 30 days, the daily PMio emission rate would be 0.0027 kilogram per
square meter (0.012 ton per acre) per day, or 0.0(K)1 1 kilogram per square meter (0.00050 ton per acre)
G-6
Air Quality
per hour. Dust suppression activities would reduce PMio emissions; however, the analysis took no credit
for normal dust suppression activities.
The estimation of the annual and 24-hour average PMio emission rates required an estimate of the size of
the area to be disturbed along with the unit area emission rate [0.000 1 1 kilogram per square meter
(0.00050 ton per acre) per hour] times 8 hours of construction per day. The analysis estimated that
20 percent of the total disturbed land area would be actively involved in construction activities at any
given time. This was based on the total disturbed area at the end of the construction period divided by the
5 years construction activities would last. Table G-4 lists the total areas of disturbance at various
repository operation areas. The analysis assumed that the entire land area required for excavated rock
storage (for both the construction and operation phases) would be disturbed by excavated rock storage
preparation activities, although only a portion of it would be used during the construction phase. The
much larger volume of rock that DOE would remove during excavation for the low thermal load scenario
would require that the excavated rock pile not be in the South Portal Operations Area. Rather, it would be
about 5 kilometers (3 miles) east of the South Portal (TRW 1999b, pages 6-41 and 6-43). The excavated
rock could be piled higher in this location [to about 15 meters (50 feet)] than in the South Portal
Operations Area [where the piles could be no more than about 6 meters (20 feet) high], requiring less land
area under this option and making the area required for all three thermal load scenarios about the same.
Table G-5 lists fugitive dust emissions from surface construction; Table G-6 lists estimated air quality
impacts from fugitive dust as the pollutant concentration in air and as the percent of the applicable
regulatory limit.
Table G-4. Land area (square kilometers)" disturbed during the construction phase
for each thermal load scenario.''''^
Operations area
High
Intermediate
Low
North Portal and roads
0.62
0.62
0.62
South Portal
0.15
0.15
0.15
Ventilation shafts
0.02
0.02
0.06
Total excavated rock storage
1.0
1.2
1.1
Rail construction on site''
0.6
0.6
0.6
Totals"
2.4
2.6
2.6
Area disturbed per year
0.48
0.52
0.50
a. To convert square kilometers to acres, multiply by 247.1.
b. Numbers are rounded to two significant figures; therefore, totals might differ from sums of values.
c. Source: Jessen (1998, all).
d. Onsite rail line assumed to be 10 kilometers (6 miles) long and 0.06 kilometer (0.04 mile) wide.
Table G-5. Fugitive dust releases from surface construction (PMip).
Pollutant emission
Emission rate
Thermal load scenario
Period
(kilograms)''
(grams per second')
High
Annual
110,000 per year
3.4
24-hour
430 per day
15"
Intermediate
Annual
120,000 per year
3.6
24-hour
460 per day
16"
Low
Annual
120,000 per year
3.7
24-hour
460 per day
16"
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert grams per second to pounds per hour, multiply by 7.9366.
d. Based on an 8-hour release period.
Fugitive dust from construction would produce small offsite PMio concentrations. The annual and
24-hour average concentrations of PMio would be about 1 percent and about 2 percent, respectively, of
the regulatory limit for all three thermal load scenarios. The differences between the thermal load
G-7
Air Quality
Table G-6. Estimated fugitive dust air quality impacts (micrograms per
cubic meter) ffom surface construction (PMip).
Maximum
Regulatory
Percent
Thermal load scenario
Period
concentration*
limit
of limit"
High
Annual
0.41
50
0.83
24-hour
2.9
150
1.9
Intermediate
Annual
0.44
50
0.88
24-hour
3.0
150
2.0
Low
Annual
0.44
50
0.88
24-hour
3.1
150
2.0
a. Numbers are rounded to two significant figures.
scenarios would be very small; the high thermal load would have the smallest impacts due mainly to the
smaller area required for excavated rock storage.
For Modules 1 and 2, the same technique was used as for the Proposed Action, but the amount of land
disturbed would be about 1.1, 1.1, and 1.3 times larger than for the Proposed Action for the high,
intermediate, and low thermal load scenarios, respectively (lessen 1998, all). The increase in disturbed
land area would lead to estimated air quality impacts about 1.1, 1.1, and 1.3 times larger than the
Proposed Action for the high, intermediate, and low thermal load scenarios, respectively.
G.1.4.2 Fugitive Dust from Subsurface Excavation
Fugitive dust would be released during the excavation of rock from the repository. Subsurface excavation 1
activities would take place 250 days per year in three 8-hour shifts per day. Excavation would generate
dust in the tunnels, and some of the dust would be emitted to the surface atmosphere through the
ventilation system. DOE estimated the amount of dust that would be emitted by the ventilation system by
using engineering judgment and best available information (DOE 1998, page 37). Table G-7 lists the
release rates of PMio for excavation activities. Table G-8 lists estimated air quality impacts from fugitive
dust as pollutant concentration in air and percentage of regulatory limit.
Table G-7.
Fugitive dust releases from excavation activities (PMio).^
Period
Emission (kilograms)"" Emission rate (grams per second)*^
Annual
24-hour
920 per year 0.029
3.7 per day 0.043"
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert grams per second to pounds per hour, multiply by 7.9366.
d. Based on a 24-hour release period.
Table G-8. Fugitive dust (PMio) and cristobalite air quality impacts
(micrograms per cubic meter) from excavation activities.
Period
Maximum
concentration"
Regulatory
limit
Percent of
regulatory limit"
PMjo
Annual
24-hour
Cristobalite
Annual
0.0035
0.044
0.0010
50
150
10"
0.0070
0.029
0.010
a. Numbers are rounded to two significant figures.
b. This value is a benchmark; there is no regulatory limit for cristobalite. See
Section G.l.
G-8
Air Quality
Fugitive dust emissions from excavation operations would produce small offsite PMio concentrations.
Both annual and 24-hour average concentrations of PMio would be much less than 1 percent of the
regulatory standards. The highest estimated annual and 24-hour excavation rates, and hence the highest
estimated fugitive dust concentrations, would be the same for all three thermal load scenarios.
Dust generated during excavation would contain cristobalite, a naturally occurring form of crystalline
silica discussed in Section G.l. The analysis estimated the amount of cristobalite released by multiplying
the amount of dust released annually (shown in Table G-7) by the percentage of cristobalite in the parent
rock (28 percent). Table G-8 also lists the potential air quality impacts for releases of cristobalite from
excavation of the repository. Because there are no public exposure limits for cristobalite, the annual
average concentration was compared to a derived benchmark level for the prevention of silicosis, as
discussed in Section G.l. The offsite cristobalite concentration would be about 0.01 percent of this
benchmark.
The air quality impacts from fugitive dust emissions from excavation operations under the construction
phase would be the same for Modules 1 and 2 as for the Proposed Action.
G.1.4.3 Fugitive Dust from Excavated Rock Pile
The disposal and storage of excavated rock on the surface excavated rock pile would generate fugitive
dust. Dust would be released during the unloading of the excavated rock and subsequent smoothing of
the excavated rock pile, as well as by wind erosion of the material. DOE used the total suspended
particulate emission for active storage piles from a report by Cowherd, Muleski, and Kinsey (1988, pages
4-17 to 4-37) to estimate fugitive dust emission. The equation is:
E = 1.9 X (s H- 1.5) X [(365 - p) - 235] x (f ^ 15)
where E = total suspended particulate emission factor (kilogram per day per hectare
[1 hectare = 0.01 square kilometer = 2.5 acres])
s = silt content of aggregate (percent)
p = number of days per year with 0.25 millimeter or more of precipitation
f = percentage of time wind speed exceeds 5.4 meters per second (12 miles per hour)
at pile height
For this analysis, s is equal to 4 percent [no value was available for this variable, so the average silt
content of limestone quarrying material (EPA 1995b, page 13.2.4-2) was used], p is 37.75 (Fransioli
1999, all) and/is 16.5 (calculated from meteorological data used in the Industrial Source Complex
model). Thus, E is equal to 7.8 kilograms of total particulates per day per hectare (6.9 pounds per day per
acre). Only about 50 percent of the total particulates would be PMio (Cowherd, Muleski, and Kinsey
1988, pages 4-17 to 4-37); therefore, the emission rate for PMio would be 3.9 kilograms per day per
hectare (3.5 pounds per day per acre).
The analysis estimated fugitive dust from disposal and storage using the size of the area actively involved
in storage and maintenance. Only a portion of the excavated rock pile would be actively disturbed by the
unloading of excavated rock and the subsequent contouring of the pile, and only that portion would be an
active source of fugitive dust. The analysis assumed that the rest of the excavated rock pile would be
stabilized by either natural processes or DOE stabilization measures and would release small amounts of
dust.
DOE based its estimate of the size of the active portion of the excavated rock pile on the amount of
material it would store there each year. The volume of rock placed on the excavated rock pile from
excavation activities during the construction phase (TRW 1999b, page 6-7) was divided by the height of
the storage pile. The average height of the excavated rock pile would be about 6 meters (20 feet) for the
G-9
Air Quality
high and intermediate thermal load scenarios (TRW 1999b, page 6-42) and 15 meters (50 feet) for the low
thermal load scenario (TRW 1999b, page 6-43). Table G-9 lists the areas of the excavated rock pile and
the active portion for each thermal load scenario. The active area of the excavated rock pile was
estimated using the total area of the rock pile at the end of the construction phase divided by the number
of years of construction multiplied by 2 (Smith 1999, all). As noted in Section G.1.4.I, under the low
thermal load scenario the excavated rock pile would be several kilometers east of the South Portal
Operations Area. Under this option the pile could be higher in this location, allowing for a smaller area of
disturbance than for the excavated rock piles of the high and intermediate thermal load scenarios in the
South Portal Operations Area.
Table G-9. Active area (square kilometers)* of excavated rock
pile during the construction phase.'''^
Number of Average annual
Thermal load Area years active area
High 0.34 5 0.14
Intermediate 0.41 5 0.17
Low 0.17 5 0.066
a. To convert square kilometers to square miles, multiply by 0.3861.
b. Numbers are rounded to two significant figures.
c. The construction phase would last 5 years. Subsurface excavation
and rock pile activities would continue during the operation and
monitoring phase (see Section G.1.5).
Table G-10 lists the fugitive dust release rate from disposal and storage of the excavated rock pile by
thermal load scenario. Table G-11 lists the air quality impacts from fugitive dust as pollutant
concentration and percent of regulatory limit.
Table G-10. Fugitive dust released from the excavated rock pile
during the construction phase (PMip).*
Emission Emission rate
Thermal load Period (kilograms)'' (grams per second)*^
High
Annual
19,000 per year
0.61
24-hour
53 per day
0.61"
Intermediate
Annual
23,000 per year
0.74
24-hour
64 per day
0.74"
Low
Annual
9,400 per year
0.30
24-hour
26 per day
0.30"
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert grams per second to pounds per hour, multiply by 7.9366.
d. Based on a continuous release.
Fugitive dust emissions from the excavated rock pile during the construction phase would produce small
offsite PMio concentrations. Both the annual and 24-hour average concentrations of PMio would be less
than 1 percent of the regulatory standards. The low thermal load scenario would have the smallest
concentrations due to the smaller area of active disturbance, which is directly related to the taller pile with
a resultant smaller surface-area-to-volume ratio.
Table G-11 also lists potential air quality impacts for releases of cristobalite. The methods used were the
same as those described in Section G. 1.4.2 for the construction phase, where cristobalite was assumed to
be 28 percent of the fugitive dust released, based on its percentage in parent rock. The land withdrawal
area boundary cristobalite concentration would be small, about 0.25 percent or less of the benchmark
level discussed in Section G. 1 .
G-10
Air Quality
Table G-11. Fugitive dust (PMio) and cristobalite air quality impacts
(micrograms per cubic meter) from the excavated rock pile during the
construction phase.
Percent of
Maximum
Regulatory
regulatory
Thermal load
Period
concentration"
limit"
limit"
PM,o
High
Annual
0.074
50
0.15
24-hour
0.62
150
0.41
Intermediate
Annual
0.090
50
0.18
24-hour
0.76
150
0.51
Low
Annual
0.036
50
0.071
24-hour
0.30
150
0.19
Cristobalite
High
Annual
0.021
10=
0.21
Intermediate
Annual
0.025
10=
0.25
Low
Annual
0.010
10=
0.010
a. Numbers are rounded to two significant figures.
b. Sources: 40 CFR 50.4 through 50.1 1 and Nevada Adminisuative Code
445B.39L
c. This value is a benchmark; there are no regulatory limits for cristobalite other
than worker exjxjsure limits. See Section G. 1 .
For Modules 1 and 2, the volume of rock excavated dtiring the construction phase would be nearly 1.8
million cubic meters (2.3 million cubic yards) for all three thermal load scenarios (TRW 1999b, pages 6-7
and 6-53). This represents an increase of about 16 percent over the Proposed Action for the high thermal
load scenario, and a slight decrease of about 5 percent for the intermediate and low thermal load
scenarios. The estimated air quality impacts would change proportionately from Proposed Action
impacts, increasing 16 percent for the high thermal load scenario and decreasing by 5 percent for the
intermediate and low thermal load scenarios.
G.1 .4.4 Fugitive Dust from Concrete Batch Facility
The concrete batch facility for the fabrication and curing of tunnel inverts and tunnel liners would emit
dust. This facility would run 3 hours a day and would produce 1 15 cubic meters (150 cubic yards) of
concrete per hour of operation (TRW 1999b, pages 4-4 and 4-5). It would operate 250 days per year.
Table G-12 lists emission factor estimates for the concrete batch facility (EPA 1995b, pages 11.12-1 to
11.12-5). About 0.76 cubic meter (1 cubic yard) of typical concrete weighs 1,800 kilograms
(4,000 pounds) (EPA 1995b, page 11.12-3). The size of the aggregate storage pile for the concrete batch
facility would be 800 square meters (0.2 acre) (TRW 1999b, pages 4-4 and 4-5).
Table G-12. Dust release rates for the concrete batch facility (kilograms
per 1,000 kilograms of concrete)."'''
Source/activity
Emission rate
Sand and aggregate transfer to elevated bin
Cement unloading to elevated storage silo
Weight hopper loading
Mixer loading
Wind erosion from aggregate storage
0.014
0.13
0.01
0.02
3.9 kilograms per hectare" per day
a. Source: EPA (1995b, page 11.12-3).
b. To convert kilograms to pounds, multiply by 2.2046.
c. 3.9 kilograms per hectare = about 21 pounds per acre.
G-11
Air Quality
Table G-13 lists the dust release rates of the concrete batch facility. The releases would be the same for
all thermal load scenarios. Table G-14 lists estimated potential air quality impacts as the estimated
pollutant concentration and percent of regulatory limit.
Table G-13. Dust release rates for the concrete batch facility
during the operation and monitoring phase (PMip)."
Emission rate
Period Emission (kilograms)'' (grams per second)*^
Annual 36,000 per year 1 . 1
24-hour 140 per day 13^
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert grams per second to pounds per hour, multiply by 7.9366.
d. Based on a 3-hour release.
Table G-14. Particulate matter (PMio) air quality impacts
(micrograms per cubic meter) from the concrete batch facility
during the construction phase.
Maximum Regulatory Percent of
Period concentration' limit regulatory limit'
Annual 0.14 50 0.27
24-hour 2^2 150 U
a. Numbers are rounded to two significant figures.
Dust emissions from the concrete batch facility during the operation and monitoring phase would produce
small offsite PMio concentrations. The annual and 24-hour averaged concentrations of PMio would be
less than 1 percent and about 1.5 percent of the regulatory standards, respectively.
For Modules 1 and 2, the air quality impacts from the concrete batch facility during the construction
phase would be the same as for the Proposed Action.
G. 1.4.5 Exhaust Emissions from Construction Equipment
Diesel- and gasoline-powered equipment would emit all four criteria pollutants during the construction
phase. EPA (1991, pages II-7-1 to 11-7-7) provided pollutant emission rate estimates for heavy-duty
equipment. This analysis assumed construction equipment would emit the average of the EPA reference
emission rates. Table G-15 lists the emission rates for this equipment.
Table G-15. Pollutant emission rates (kilograms" per
1,000 liters'* of fuel) for construction equipment.*^
Estimated emission
Pollutant Diesel Gasoline
Carbon monoxide 15 450
Nitrogen dioxide 39 13
PMio 3.5 0.86
Sulfur dioxide 3/7 0.63
a. To convert kilograms to pounds, multiply by 2.2046.
b. To convert liters to gallons, multiply by 0.26418.
c. Source: Averageof rates fi'om EPA (1991, pages II-7-1 to
II-7-7).
Table G-16 lists the estimated average amount of fuel per year for the construction of the North and South
Portal Operations Areas. The fuel for the South Portal Operations Area would include fuel consumed
during maintenance of the excavated rock pile.
G-12
Air Quality
Table G-16. Amount of fuel consumed per year during the
construction phase
(liters).'"
South Portal
Operations Area''
North Portal
Operations Area**
Thermal load
Diesel
Gasoline
Diesel
High
Intermediate
Low
360,000
360,000
560,000
20,000
20,000
20,000
640,000
640,000
640,000
a. To convert liters to gallons, multiply by 0.26418.
b. Numbers are rounded to two significant figures.
c. Source: Based on total fuel use from TRW (1999b, page 6-3).
d. Source: Basedon total fuel use from TRW (1999a, Table 6.1, page 71).
Table G-17 lists pollutant releases from construction equipment for each thermal load scenario. The
emission rate for the annual concentration was calculated from the total fuel consumed, assuming the
same amount of fuel would be consumed each year.
Table G-17. Pollutant release rates from surface equipment during the construction phase.
Mass of pollutant per
Emission rate'
averaging period (kilograms)''
(grams per second)''
Pollutant
Period
South
North
South
North
High and intermediate thermal load
Nitrogen dioxide
Annual
14,000
25,000
0.46
0.80
Sulfur dioxide
Annual
1,400
2,400
0.043
0.076
24-hour
5.4
9.6
0.019
0.33
3-hour
2.0
3.6
0.019
0.33
Carbon monoxide
8-hour
57
39
2.0
1.3
1 -horn-
7.2
4.8
2.0
1.3
PM,o
Annual
1,300
2,200
0.040
0.071
24-hour
5.1
8.9
0.18
0.31
Low thermal load
Nitrogen dioxide
Annual
22,000
25,000
0.71
0.80
Sulfur dioxide
Annual
2,100
2,400
0.067
0.076
24-hour
8.4
9.6
0.29
0.33
3-hour
3.2
3.6
0.29
0.33
Carbon monoxide
8-hour
69
39
2.4
1.3
1-hour
8.7
4.8
2.4
1.3
PM,o
Annual
2,000
2,200
0.062
0.071
24-hour
7.9
8.9
0.27
0.31
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. Based on an 8-hour release for averaging pwriods 24 hours or less.
d. To convert grams per second to pounds per hour, multiply by 7.9366.
Table G-18 lists the impacts on air quality from construction equipment emission by thermal load
scenario as the pollutant concentration in air and the percent of the regulatory limit. Emissions from
surface equipment during the construction phase would produce small offsite (outside the land withdrawal
area) criteria pollutant concentrations. All concentrations would be less than 1 percent of the regulatory
standards.
For Modules 1 and 2, the same analysis method was used as that for the Proposed Action, but the amount
of fuel used in the South Portal Operations Area would vary from the Proposed Action. Diesel fuel use
would be about 7.4 times larger for the high and intermediate thermal load scenarios and about 4.8 times
larger for the low thermal load scenario. Gasoline use would be two times larger for all thermal load
scenarios (TRW 1999b, page 6-45). There would be no change in the amount of fuel used during the
G-13
Air Quality
Annual
0.13
100
0.13
Annual
0.013
80
0.016
24-hour
0.096
365
0.026
3-hour
0.77
1,300
0.059
8-hour
1.8
10,000
0.018
1-hour
11
40,000
0.028
Annual
0.012
50
0.024
24-hour
0.090
150
0.060
Annual
0.16
100
0.16
Annual
0.016
80
0.020
24-hour
0.12
365
0.032
3-hour
0.93
1,300
0.071
8-hour
2.1
10,000
0.020
1-hour
12
40,000
0.031
Annual
0.014
50
0.029
24-hour
0.11
150
0.072
Table G-18. Air quality impacts from construction equipment during the construction phase
(micrograms per cubic meter)."
Maximum Regulatory Percent of
Pollutant Period concentration limit'' regulatory limit
High and intermediate thermal load
Nitrogen dioxide
Sulfur dioxide
Carbon monoxide
PM,o
Low thermal load
Nitrogen dioxide
Sulfur dioxide
Carbon monoxide
PM.o
a. Numbers are rounded to two significant figures.
b. Source: 40 CFR 50.4 through 50. 1 1 and Nevada Administrative Code 445B.39 1 .
construction of the North Portal. These increases in fuel use would lead to estimated air quaUty impacts
that would be about 3.5 times larger for the high and intermediate thermal load scenarios and about 2.5
times larger for the low thermal load scenario except for carbon monoxide. Carbon monoxide air quality
impacts, which are more heavily weighted towards gasoline, would be about 2.5, 2.5 and 2.0 times larger
than the Proposed Action for the high, intermediate, and low thermal load scenarios, respectively.
G.1.4.6 Exhaust from Boiler
A proposed boiler in the South Portal Operations Area would emit the four criteria pollutants. The boiler
would use diesel fuel and provide steam and hot water for the heating, ventilation, and air conditioning
system. The analysis assumed that this boiler would be the same size as the boiler that would operate in
the North Portal Operations Area during the operation and monitoring phase (TRW 1999a, Table 6-2,
page 75) but not during construction. Table G-19 lists the annual emission rates of the boiler in the South
Portal Operations Area. To estimate the short-term (24 hours or less) emission rate, the analysis assumed
the boiler would run 250 days (6,000 hours) per year. Given the annual boiler emissions, this was a
conservative assumption because continuous operation 365 days (8,760 hours) per year would result in
lower daily emissions. This assumption considered periods when the boiler would not be operating. The
actual period of boiler operation is not known. In addition, specific information on the boiler stack height
and exhaust air temperature (which would affect plume rise) has not been developed. The analysis
assumed that releases would be from ground level, which overestimates actual concentrations. Table
G-20 lists releases of criteria pollutants by the boiler. Table G-21 lists estimated potential air quality
impacts as pollutant concentrations in air and percent of regulatory limit.
Table G-19. Annual pollutant release rates (kilograms per year)" for the South
Portal Operations Area boiler.'''^ __^
Pollutant Annual emission rate
Nitrogen dioxide 58,000
Sulfur dioxide 20,000
Carbon monoxide 15,000
PMin 5,600
a. To convert kilograms to tons, multiply by 0.0011023.
b. Source: TRW (1999a, Table 6-2, page 75).
c. Numbers are rounded to two significant figures.
G-14
Air Quality
Table G-20. Pollutant release rates from the boiler during the construction
phase.'
Mass of pollutant
(kilograms)''
per
Emission rate*^
Pollutant
Period
averagmg time
(grams per second)**
Nitrogen dioxide
Annual
58,000
1.83
Sulfur dioxide
Annual
20,000
0.63
24-hour
80
0.92
3-hour
10
0.92
Carbon monoxide
8-hour
20
0.67
1-hour
2.5
0.67
PM,o
Annual
5,600
0.18
24-hour
22
0.25
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. Based on an 8-hour release for averaging periods of 24 hours or less.
d. To convert grams per second to pounds f)er hour, multiply by 7.9366.
Table G-21. Air quality impacts from boiler pollutant releases from the South
Portal Operations Area during the construction phase (micrograms per cubic
meter of pollutant).
Maximum
Percent of
Pollutant
Period
concentration"
Regulatory limit*"
regulatory limit'
Nitrogen dioxide
Annual
0.22
100
0.22
Sulfur dioxide
Annual
0.076
80
0.095
24-hour
0.94
365
0.26
3-hour
5.5
1,300
0.43
Carbon
8-hour
2.0
10,000
0.020
monoxide
1-hour
12
40,000
0.031
PM,o
Annual
0.022
50
0.044
24-hour
0.27
150
0.18
a. Numbers are rounded to two significant figures.
b. Sources: 40 CFR 50.4 through 50.11 and Nevada Administrative Code 445B.391.
Emissions from the boiler during the construction phase would produce small offsite (outside the land
withdrawal area) criteria pollutant concentrations. All concentrations would be less than 1 percent of the
regulatory standards.
For Modules 1 and 2, the air quality impacts from the boiler during the construction phase would be the
same as those for the Proposed Action.
G.1.5 OPERATION AND MONITORING PHASE
This section describes the method DOE used to estimate air quality impacts during the operation and
monitoring phase (2010 to 2110). Activities during this phase would include the continued development
of the subsurface facilities, which would last 22 years for all thermal load scenarios. Emplacement
activities in the surface and subsurface facilities would continue concurrently with development
operations for 24 years; 76 years of monitoring and maintenance would begin after the end of
emplacement operations. The duration of the monitoring and maintenance period has not been finalized,
but could be as long as 276 years for a 300-year operation and monitoring phase. For purposes of
analysis, workers would use the following schedule for activities during the operation and monitoring
phase: three 8-hour shifts a day, 5 days a week, 50 weeks a year; the maintenance of the excavated rock
pile would occur in one 8-hour shift a day, 5 days a week, 50 weeks a year.
G-15
Air Quality
For Modules 1 and 2, the continued development of the subsurface facilities would last 36 years for all
thermal load scenarios. Emplacement activities in the surface and subsurface facilities would continue
concurrently with development operations for 38 years. The duration of the monitoring and maintenance
period has not been finalized, but could be as long as 262 years for a 300-year operation and monitoring
phase.
The analysis estimated air quality impacts by calculating pollutant concentrations from various operation
and monitoring activities. Emission rates were developed for each activity that would result in pollutant
releases. The emission rates were multiplied by the unit release concentrations (see Section G.1.3) to
calculate the pollutant concentration for comparison to the various regulatory limits.
The principal emission sources of particulates would be dust emissions from concrete batch facility
operations and fugitive dust emissions from excavation and storage on the excavated rock pile. Fuel
combustion from maintenance of the excavated rock pile and emissions from the North Portal and Soutl
Portal boilers would be principal sources of nitrogen dioxide, sulfur dioxide, and carbon monoxide. Th
following sections describe these sources in more detail.
G.1.5.1 Fugitive Dust from Concrete Batch Facility
The concrete batch facility for the fabrication and curing of tunnel inverts and liners would emit dust.
The analysis assumed that the dust emissions from the concrete batch facility would be the same as those
during the construction phase. Thus, the dust release rate and potential air quality impacts would be the :
same as those listed in Tables G-13 and G-14.
G.1 .5.2 Fugitive Dust from Subsurface Excavation
The excavation of rock from the repository would generate fugitive dust in the drifts. Some of the dust
would reach the external atmosphere through the repository ventilation system. Fugitive dust emission
rates from excavation during operations would be the same as those during the construction phase. Thus,
the fugitive dust release rate and potential air quality impacts for excavation of rock would be the same as
those listed in Tables G-7 and G-8. Air quality impacts from cristobalite released during excavation of
the repository would be the same as those listed in Table G-8.
G.1 .5.3 Fugitive Dust from Excavated Rocl( Pile
The disposal and storage of excavated rock on the excavated rock pile would release fugitive dust. The
analysis used the same method to estimate fugitive dust releases from the excavated rock pile during
operations that it used for the construction phase (See Section G. 1.4.3). Table G-22 lists the areas of the
active portion of the excavated rock pile by thermal load scenario. The total land area used for storage
and the active portion of the excavated rock pile was based on the amount of rock that would be stored
during operations (TRW 1999b, page 6-17). Sections G.1. 4.1 and G. 1.4.3 compare the excavated rock
pile areas for the three thermal load scenarios.
Table G-22. Estimated active excavated rock pile area (square kilometers)^ during
subsurface excavation activities during the operation and monitoring phase.''
Years of repository Annual average
Thermal load Storage area development active area
High 0.63 22 0.058
Intermediate 0.76 22 0.069
Low LO 22 0.095
a. To convert square kilometers to acres, multiply by 247. 1 .
b. Numbers are rounded to two significant figures.
G-16
M
Air Quality
While the land area used for storage of excavated rock during the operation and monitoring phase would
be nearly twice as large as that used during the construction phase for the high and intermediate thermal
load scenarios, the active area per year would be about half of that for construction due to the larger
number of years over which storage would occur (22 years compared to 5 years). The land area used
during the operation and monitoring phase for the low thermal load scenario would be nearly 10 times
that used during the construction phase. The annual active area would be larger during the operation and
monitoring phase than during the construction phase, but only about twice as large because of the longer
period over which storage would take place (22 years compared to 5 years). Table G-23 lists fugitive dust
releases from the excavated rock pile; Table G-24 lists potential air quality impacts as the pollutant
concentration and percent of the regulatory limit.
Table G-23. Fugitive dust release rate from the excavated rock pile during the
operation and monitoring phase (PMip).^
Emissions
Emission rate'^
Thermal load
Period
(kilograms)''
(grams
> per second)
High
Annual
8,200 per year
0.26
24-hour
22 per day
0.26
Intermediate
Annual
9,800 per year
0.31
24-hour
27 per day
0.31
Low
Annual
13,000 per year
0.42
24-hour
37 per day
0.42
a Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. Based on a continuous release.
d. To convert grams per second to pounds per hour, multiply by 7.9366.
Table G-24. Fugitive dust (PMio) and cristobalite air quality impacts from the
excavated rock pile during the operation and monitoring phase (micrograms per
cubic meter).
Percent of
Maximum
Regulatory
regulatory
Thermal load
Period
concentration"
limit''
limit"
PM,o
High
Annual
0.031
50
0.062
24-hour
0.27
150
0.18
Intermediate
Annual
0.038
50
0.075
24-hour
0.32
150
0.21
Low
Annual
0.051
50
0.10
24-hour
0.43
150
0.29
Cristobalite
High
Annual
0.0087
10'
0.087
Intermediate
Annual
0.011
10'
0.11
Low
Annual
0.014
10'
0.14
a. Numbers are rounded to two significant figures.
b. Source: 40 CFR 50.4 through 50. 1 1 and Nevada Administrative Code 445B.39 1 .
c. This value is a benchmark; there is no regulatory limit for cristobalite. See Section G.l.
Fugitive dust emissions from the excavated rock pile during the operation and monitoring phase would
produce very small offsite (outside the land withdrawal area) PMio concentrations. Both annual and
24-hour average concentrations of PMio would be less than 1 percent of the regulatory standards for all
three thermal load scenarios.
Table G-24 also lists potential air quality impacts for releases of cristobalite. The methods used were the
same as those described in Section G. 1.4.2 for the construction phase, where cristobalite was assumed to
be 28 percent of the fugitive dust released, based on its percentage in parent rock. The site boundary
G-17
Air Quality
cristobalite concentration would be small, about 0. 1 percent of the benchmark level discussed in Section
G.l.
The Module 1 and 2 analysis used the same technique as for the Proposed Action, but the estimated active
excavated rock pile area would be about 1 .4, 1 .2, and 1 . 1 times larger than the Proposed Action for the
high, intermediate, and low thermal load scenarios, respectively, based on the volumes of rock added
annually to the pile (TRW 1999b, page 6-56). The estimated air quality impacts from the excavated rock
pile would also be 1.4, 1.2, and 1.1 times larger than the Proposed Action for the high, intermediate, and
low thermal load scenarios, respectively.
G.1.5.4 Exhaust from Excavated Rock Pile Maintenance Equipment
Surface equipment would emit the four criteria pollutants during excavated rock pile maintenance. The
analysis used the same method to determine air quality impacts for surface equipment during operations
that it used for the construction phase (see Section G. 1.4.5). Table G-15 lists the pollutant release rates of
the equipment. Table G-25 lists the average amount of fuel consumed each year during the operation and
monitoring phase at the South Portal Operations Area.
Table G-25. Annual amount of fuel (liters)^ consumed
during the operation and monitoring phase.'''^
I
Thermal load Diesel Gasoline
High 350,000 4,500
Intermediate 350,000 4,500
Low 2,800,000 9,000
a. To convert liters to gallons, multiply by 0.26418.
b. Source: Based on total fuel use from TRW (1999b, pages 6-14
and 6-21).
c. Numbers are rounded to two significant figures.
I
Table G-26 lists pollutant release rates for surface equipment during operations activities of the operation
and monitoring phase. Monitoring activity emissions would be much smaller. Table G-27 lists potential
air quality impacts.
Table G-26. Pollutant release rates from surface equipment during the operation and monitoring phase.'
Mass of pollutant per Emission rate*^
Pollutant Period averaging time (kilograms)'' (grams per second)
High and intermediate thermal load
Nitrogen dioxide Annual 14,000 0.44
Sulfur dioxide Annual 1,300 0.041
24-hour 5.2 0.18
3-hour 4.9 0.18
Carbon monoxide 8-hour 29 1.0
1-hour 3.6 1.0
PMio Annual 1,200 0.039
24-hour 4.9 0.17
Low thermal load
Nitrogen dioxide Annual 1 10,000 3.5
Sulfur dioxide Annual 10,000 0.33
24-hour 42 1.4
3-hour 16 1.4
Carbon monoxide 8-hour 180 6.4
1-hour 23 6.4
PM,o Annual 9,700 0.31
24-hour 39 L4
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. Based on an 8-hour release for averaging periods of 24 hours or less.
d. To convert grams per second to pounds per hour, multiply by 7.9366.
G-18
Air Quality
Table G-27. Air quality impacts from surface equipment during the operation and monitoring phase
(micrograms per cubic meter of pollutant).
Maximum
Percent of
Pollutant
Period
concentration^
Regulatory limit*"
regulatory limit'
High and intermediate thermal load
Nitrogen dioxide
Annual
0.052
100
0.052
Sulfur dioxide
Annual
0.0049
80
0.0063
24-hour
0.034
365
0.0094
3-hour
0.27
1,300
0.021
Carbon monoxide
8-hour
0.58
10,000
0.0056
1-hour
3.3
40,000
0.0084
PM.o
Annual
0.0046
50
0.0092
24-hour
0.032
150
0.021
Low thermal load
Nitrogen dioxide
Annual
0.42
100
0.42
Sulfur dioxide
Annual
0.040
80
0.051
24-hour
0.28
365
0.076
3-hour
2.2
1,300
0.17
Carbon monoxide
8-hour
3.7
10,000
0.036
1-hour
21
40,000
0.053
PM.o
Annual
0.037
50
0.074
24-hour
0.26
150
0.17
a. Numbers are rounded to two significant figures.
b. Sources: 40 CFR 50.4 through 50.11 and Nevada Administrative Code 445B.391.
Emissions from surface equipment during operation and monitoring would produce very small
concentrations of offsite (outside the land withdrawal area) criteria pollutants. All estimated
concentrations would be less than 1 percent of the regulatory standards.
The Module 1 and 2 analysis used the same technique as for the Proposed Action, but the amount of fuel
used during the operation and monitoring phase would increase. Annual diesel fuel use during
development would increase by 1.6, 3.0, and 2.0 times the Proposed Action; annual gasoline use would
increase by 1.2, 1.8, and 1.5 times the Proposed Action for the high, intermediate, and low thermal load
scenarios, respectively, based on total fuel use (TRW 1999b, page 6-53). Annual diesel fuel use during
emplacement would increase only by about 1 percent over the Proposed Action for all thermal load
scenarios (TRW 1999b, page 6-61). Estimated air quality impacts for surface equipment during the
operation and monitoring phase under Module 1 and 2 would increase by about 1.6, 3.0, and 2.0 times the
Proposed Action for the high, intermediate, and low thermal load scenarios.
G.1.5.5 Exhaust from Boiler
Boilers in the North and South Portal Operations Areas would emit the four criteria pollutants. The
annual emission rates of the boiler in the North Portal Operations Area would be the same as those listed
in Table G-19 (the boilers were assumed to be the same size). There would be small variations in the
North Portal boiler emissions for the transportation and waste packaging options because of different
operational requirements. The emissions listed in Table G-19 are for the combination of legal-weight
truck transport and uncanistered waste scenario, which would require the largest boiler because a larger
Waste Handling Building would be required (TRW 1999a, pages 66 to 75). Other options would require
a slightly smaller boiler (TRW 1999a, Table 6-2, page 75) and the release rate of pollutants would be
about 15 percent smaller. The size of the boiler would not depend on the thermal load scenario. The
analysis assumed the boiler would run 250 days (6,000 hours) per year. Given an annual emission rate,
this was a conservative assumption because continuous operation 365 days (8,760 hours) per year would
result in lower daily emissions. This assumption considered periods when the boiler would not be
operating. The actual period of boiler operation is not known. Rates from the North Portal boiler for
G-19
Air Quality
evaluating pollutant releases during the operation and monitoring phase would be the same as those listed
in Table G-20 for the South Portal boiler.
Table G-28 lists estimated potential air quality impacts as pollutant concentrations in air and percent of
regulatory limit. These impacts would be due to emissions from the boilers in the North and South Portal
Operations Areas. Although total emissions during the operation and monitoring phase would be double
those during the construction phase (when only the South Portal boiler would operate), air quality impacts
would not double because of different atmospheric dispersion factors from the two operations areas to the
location of the hypothetically maximally exposed individual. Emissions from the two boilers during the
operation and monitoring phase would produce small offsite criteria pollutant concentrations. All
concentrations would be less than 1 percent of the regulatory standards.
Table G-28. Air quality impacts from boiler pollutant releases from both North and
South Portal Operations Areas (micrograms per cubic meter of pollutant).
Maximum
Regulatory
Percent of
Pollutant
Period
concentration*
limit"
regulatory limit'
Nitrogen dioxide
Annual
0.40
100
0.40
Sulfur dioxide
Annual
0.14
80
0.18
24-hour
1.8
365
0.49
3-hour
11
1,300
0.85
Carbon monoxide
8-hour
3.7
10,000
0.037
1-hour
24
40,000
0.061
PM,o
Annual
0.039
50
0.078
24-hour
0.51
150
0.34
a. Numbers are rounded to two significant figures.
b. Sources: 40 CFR 50.4 through 50. 1 1 and Nevada Administrative Code 445B.391 .
For Module 1 or 2, the estimated air quality impacts from boilers during the operation and monitoring
phase would be the same as those for the Proposed Action.
G.1.6 CLOSURE PHASE
This section describes the method used to estimate air quality impacts during the closure phase at the
proposed repository. The closure phase would last 6, 6, or 15 years for the high, intermediate, or low
thermal load scenario, respectively. For Modules 1 and 2, the closure phase would last 13, 17, and
27 years for the high, intermediate, and low thermal load scenarios, respectively. The work schedule
would be one 8-hour shift per day, 5 days a week, 50 weeks a year.
The analysis estimated air quality impacts by calculating pollutant concentrations from various closure j
activities. Emission rates were developed for each activity that would result in releases of pollutants.
These pollutant emission rates were then multiplied by the unit release concentration (see Section G.1.3)
to calculate the pollutant concentration for comparison to the various regulatory limits.
The sources of particulates would be emissions from the backfill plant and the concrete batch facility and
fugitive dust from closure activities on the surface and the reclamation of material from the excavated
rock pile for backfill. The principal source of nitrogen dioxide, sulfur dioxide, and carbon monoxide
during closure would be fuel combustion. The following sections describe these sources in more detail.
G.1.6.1 Dust from Backfill Plant
The Closure Backfill Preparation Plant would process (separate, crush, screen, and wash) rock from the
excavated rock pile for use as backfill for the underground access openings (TRW 1999b, pages 4-77 and
4-78). The facility would have the capacity to handle 91 metric tons (100 tons) an hour (TRW 1999b,
G-20
Air Quality
pages 4-77 and 4-78). For purposes of analysis, the backfill plant would run 6 hours a shift, 2 shifts a
day, 5 days a week, 50 weeks a year.
The plant was assumed to have emissions similar to a crushed-stone processing plant. Table G-29 lists
the emission rates for various activities associated with a crushed stone processing plant (EPA 1995b,
pages 11.19.2-1 to 11.19.2-8). Table G-30 lists estimated pollutant release rates for the backfill plant.
Table G-31 lists potential air quality impacts as pollutant concentrations in air and percent of regulatory
limit.
Table G-29. Emission rates from a crushed stone processing plant.*"**
Emission rate (kilogram'^ per 1,000
Source/activity kilograms of material processed) •
Dump to conveyor or truck 0.00005
Screening 0.0076
Crusher 0.0012
Fine screening 0.036
a. Source: EPA (1995b, pages 11.19.2-1 to 11.19.2-8).
b. Numbers are rounded to two significant figures.
c. To convert kilograms to pounds, multiply by 2.2046.
Table G-30. Dust release rates from the backfill plant (PMip).'
Emission Emission rate
Period (kilograms)'' (grams per second)*^
Annual 12,000 per year 0.39
24-hour 49 per day U^
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert grams per second to pounds per hour, multiply by 7.9366.
d. Based on a 12-hour release period.
Table G-31. Particulate matter (PMio) air quality impacts from
backfill plant (micrograms per cubic meter).
Maximum Regulatory Percent of regulatory
Period concentration' limit'' limit'
Annual 0.047 50 0.093
24-hour U 150 OTl
a. Numbers are rounded to two significant figures.
b. Sources: 40 CFR 50.4 through 50.1 1 and Nevada Administrative Code
445B.391.
Dust emissions from the backfill plant would produce small PMio concentrations. Both annual and
24-hour average concentrations of PMio would be less than 1 percent of the regulatory standards for all
thermal load scenarios.
For Modules 1 and 2, the estimated air quality impacts for the backfill plant would be the same as those
for the P*roposed Action.
G.1.6.2 Fugitive Dust from Concrete Batch Facility
A concrete batch facility for the fabrication of seals would be similar to the facility that would operate
during the construction and operation and monitoring phases (see Sections G. 1.4.4 and G. 1.5.1). The
only difference would be that it would run only ten 3-hour shifts a year per concrete seal (TRW 1999b,
page 4-78). The analysis assumed that two seals per year would be produced. Table G-12 lists activities
associated with the concrete batch facility and their emissions. Table G-32 lists emissions from the
concrete batch facility during closure. Table G-33 lists potential air quality impacts as pollutant
concentration in air and percent of regulatory limit.
G-21
Air Quality
Table G-32. Dust release rates from the concrete batch facility
during the closure phase (PMip)."
Mass of pollutant Emission rate
Period (kilograms)'' (grams per second)'^
Annual 2,800 per year 0.090
24-hour 140 per day 13^
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert grams per second to pounds per hour, multiply by 7.9366.
d. Based on a 3-hour release period.
Table G-33. Particulate matter (PMio) air quality impacts from the
concrete batch facility during the closure phase (micrograms per
cubic meter).
Maximum Regulatory Percent of
Period concentration" limit'' regulatory limit'
Annual 0.011 50 0.022
24-hour 2,2 150 L5
a. Numbers are rounded to two significant figures.
b. Source: 40 CFR 50.4 through 50. 1 1 and Nevada Administrative Code
445B.391.
Dust emissions from the concrete batch facility during closure would produce small offsite (outside the
land withdrawal area) PMio concentrations. The annual and 24-hour average concentrations of PMio
would be less than 1 percent and around 1 .5 percent, respectively, of the regulatory standards.
For Modules 1 and 2, the estimated air quality impacts from the concrete batch facility during the closure
phase would be the same as those for the Proposed Action.
G.1 .6.3 Fugitive Dust from Closure Activities
Closure activities such as smoothing and reshaping the excavated rock pile and demolishing buildings
would produce the same fugitive dust releases as construction activities because they would disturb nearly
the same amount of land. Thus, the pollutant release and air quality impacts from fugitive dust emissions
from surface closure activities would be the same as those listed in Tables G-5 and G-6, respectively.
G.1.6.4 Fugitive Dust from Excavated Rock Pile
During backfill operations, fugitive dust would occur from the removal of excavated rock from the
storage pile. The analysis used the same method to estimate fugitive dust emission from the excavated
rock pile during the closure phase that it used for the construction phase (Section G. 1.4.3). Table G-34
lists the total area of the excavated rock pile disturbed and the active portion, based on the amount of
material to be removed from the pile (TRW 1999b, page 6-39). The analysis assumed that the rock used
Table G-34. Active excavated rock pile area (square kilometers)" during the
closure phase.''
Total area disturbed Number of Active area
Thermal load for backfill operation years of closure (per year)
High 0.21 6 0.069
Intermediate 0.27 6 0.091
Low 026 15 0.035
a. To convert square kilometers to acres, multiply by 247.1.
b. Numbers are rounded to two significant figures.
G-22
Air Quality
in backfill would be from a limited area of the excavated rock pile, rather than from all over the pile.
Table G-35 lists fugitive dust releases from the excavated rock pile. Table G-36 lists potential air quality
impacts from the pile as pollutant air concentration and percent of regulatory limit.
Table G-35. Fugitive dust release rates from the excavated rock pile during the
closure phase (PMip)."
Emission
Emission rate*^
Thermal load
Period
(kilograms)''
(grams per second)''
High
Annual
9.800 per year
0.31
24-hour
27 per day
0.31
Intermediate
Annual
13,000 per year
0.41
24-hour
35 per day
0.41
Low
Annual
5,000 per year
0.16
24-hour
14 per day
0.16
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. Based on a continuous release.
d. To convert grams per second to pounds per hour, multiply by 7.9366.
Table G-36. Fugitive dust (PMio) and cristobalite air quality impacts from the
excavated rock pile during the closure phase (micrograms per cubic meter).
Maximum
Regulatory
Percent of
Thermal load
Period
concentration"
limit"
regulatory limit"
PM,o
High
Annual
0.037
50
0.074
24-hour
0.32
150
0.21
Intermediate
Annual
0.049
50
0.098
24-hour
0.42
150
0.28
Low
Annual
0.019
50
0.038
24-hour
0.16
150
0.11
Cristobalite
High
Annual
0.010
W
0.10
Intermediate
Annual
0.014
l(f
0.14
Low
Annual
0.0053
10-^
0.053
a. Numbers are rounded to two significant figures.
b. Source: 40 CFR 50.4 through 50. 11 and Nevada Administrative Code 445B.39 1.
c. This value is a benchmark; there is no regulatory limit for cristobalite. See Section G.l.
Fugitive dust emissions from the excavated rock pile during closure would produce small offsite PMio
concentrations. Both the annual and 24-hour average concentrations of PMio would be less than 1 percent
of the regulatory standards for all three thermal load scenarios.
Table G-36 also lists potential air quality impacts for releases of cristobalite. The methods used were the
same as those described in Section G.I. 4.2 for the construction phase, where cristobalite was assumed to
be 28 percent of the fugitive dust released, based on its percentage in parent rock. The land withdrawal
area boundary cristobalite concentration would be small, about O.I percent of the benchmark level
discussed in Section G.I.
For Modules 1 and 2, the same technique was used, but the estimated active excavated rock pile area
would be about 20 percent larger, 4 percent smaller, and 6 percent larger than the Proposed Action for the
high, intermediate, and low thermal load scenarios, respectively, based on the volume of rock added to the
pile (TRW 1999b, page 6-79). The estimated air quality impacts from the excavated rock pile would also
be about 20 percent larger, 4 percent smaller, and 6 percent larger than the Proposed Action for the high,
intermediate, and low thermal load scenarios, respectively.
G-23
Air Quality
G. 1.6.5 Exhaust Emissions from Surface Equipment
The consumption of diesel fuel and gasoline by surface equipment would emit the four criteria pollutants
during closure. The analysis used the same method to determine pollutant release rates during closure
that it used for the construction phase (see Section G. 1.4.5). Table G-15 lists the estimated pollutant
release rates of the equipment that would consume the fuel. Table G-37 lists by thermal load scenario the
average amount of fuel consumed per year. The length of the closure phase would be 6, 6, or 15 years for
the high, intermediate, or low thermal load scenario, respectively. Closure of the North Portal Operations
Area would last 6 years (TRW 1999a, page 79).
Table G-37. Annual amount of fuel consumed (liters)" during the closure phase.**
Thermal load
South Portal diesel''
North Portal diesel
High
Intermediate
Low
250,000
620,000
510,000
340,000
340,000
340,000
a. To convert liters to gallons, multiply by 0.26418.
b. Numbers are rounded to two significant figures.
c. Source: Based on total fuel consumed from TRW (1999b, page 6-37).
d. Source: Based on total fuel consumed from TRW (1998, page 87).
Table G-38 lists pollutant releases from surface diesel consumption. Table G-39 lists potential air quality
impacts as pollutant concentration in air and percent of regulatory limit. Concentrations would be less
than 1 percent of the regulatory limit for all thermal load scenarios.
Table G-38. Pollutant release rates from surface equipment during the closure phase.^
Mass of pollutant per averaging
period (kilograms)''
Emission rate'
(grams per second)''
Pollutant
Period
South
North
South
North
High thermal load
Nitrogen dioxide
Annual''
9,800
13,000
0.31
0.42
Sulfur dioxide
Annual
930
1,300
0.030
0.040
24-hour''
3.7
5.1
0.13
0.18
3-hourf
1.4
1.9
0.13
0.18
Carbon monoxide
8-hour^
15
21
0.52
0.71
l-hour"
1.9
2.6
0.52
0.71
PM,o
Annual
870
1,200
0.028
0.038
24-hour
3.5
4.7
0.12
0.16
Intermediate thermal load
Nitrogen dioxide
Annual
24,000
13,000
0.77
0.42
Sulfur dioxide
Annual
2,300
1,300
0.073
0.040
24-hour
9.2
5.1
0.32
0.18
3-hour
3.5
1.9
0.32
0.18
Carbon monoxide
8-hour
37
21
1.3
0.71
1-hour
4.7
2.6
1.3
0.71
PM.o
Annual
2,100
1,200
0.068
0.038
24-hour
8.6
4.7
0.30
0.16
Low thermal load
Nitrogen dioxide
Annual
20,000
13,000
0.63
0.42
Sulfur dioxide
Annual
1,900
1,300
0.060
0.040
24-hour
7.6
5.1
0.26
0.18
3-hour
2.8
1.9
0.26
0.18
Carbon monoxide
8-hour
31
21
1.1
0.71
1-hour
3.8
2.6
1.1
0.71
PM,„
Annual
1,800
1,200
0.056
0.038
24-hour
7.1
4.7
0.24
0.16
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. Based on an 8-hour release period for averaging periods of 24 hours or less.
d. To convert grams per second to pounds per hour, multiply by 7.9366.
G-24
Air Quality
Table G-39. Air quality impacts (micrograms per cubic meter) from surface construction equipment
during the closure phase.
u •
Maximum
Percent of
Pollutant
Period
concentration^
Regulatory limit''
regulatory limit^
High thermal load
Nitrogen dioxide
Annual
0.080
100
0.080
Sulfur dioxide
Annual
0.0076
80
0.0095
24-hour
0.057
365
0.016
3-hour
0.45
1,300
0.035
Carbon monoxide
8-hour
0.67
10,000
0.0065
1-hour
4.1
40,000
0.010
PM.o
Annual
0.0071
50
0.014
24-hour
0.053
150
0.035
Intermediate thermal load
Nitrogen dioxide
Annual
0.13
100
0.13
Sulfur dioxide
Annual
0.013
80
0.016
24-hour
0.093
365
0.025
3 -hour
0.74
1,300
0.057
Carbon monoxide
8-hour
1.1
10,000
0.011
1-hour
6.6
40,000
0.017
PMio
Annual
0.012
50
0.024
24-hour
0.087
150
0.058
Low thermal load
Nitrogen dioxide
Annual
0.12
100
0.12
Sulfur dioxide
Annual
0.011
80
0.015
24-hour
0.082
365
0.022
3-hour
0.66
1,300
0.050
Carbon monoxide
8-hour
0.98
10,000
0.0095
1-hour
5.9
40,000
0.015
PMio
Annual
0.010
50
0.020
24-hour
0.076
150
0.051
a. Numbers are rounded to two
significant figures.
b. Sources: 40 CFR 50.4 throui
'h 50.1 1 and Nevada Administrative Code 445B.391.
For Modules 1 and 2, the same technique was used, but the amount of fuel used during the closure phase
would increase. The annual diesel fuel use during closure would be 1.9, 0.81, and 1.2 times that of the
Proposed Action for the high, intermediate, and low thermal load scenarios, respectively, based on total
fuel use (TRW 1999b, page 6-77). The annual diesel fuel use for closure of the North Portal facility
would be the same as that for the Proposed Action for all thermal load scenarios. Estimated air quality
impacts for surface equipment during the operation and monitoring phase under Modules 1 and 2 would
increase by about 1.4, 0.87, and 1.1 times the Proposed Action for the high, intermediate, and low thermal
load scenarios, respectively.
G.1.7 RETRIEVAL SCENARIO
This section describes the method used to estimate air quality impacts during possible retrieval at the
proposed repository. The retrieval contingency includes the construction of a retrieval storage facility and
storage pad, and retrieval of the waste. Retrieval would last 1 1 years (TRW 1999b, page 6-32), while
construction of the retrieval storage facility and storage pads would last 10 years (TRW 1999a, page
1-20). DOE would construct the storage facility before beginning retrieval activities. Storage pads would
be constructed in modules concurrently with retrieval activities. The analysis considered concurrent air
quality impacts of retrieval and construction. The retrieval scenario work schedule would be one 8-hour
shift a day, 5 days a week, 50 weeks a year.
G-25
Air Quality
The analysis estimated air quality impacts by calculating pollutant concentrations from various activities
associated with retrieval. Emission rates were developed for each activity that would result in releases of
pollutants. These rates were multiplied by the unit release concentration (see Section G.I.3) to calculate
pollutant concentrations for comparison to the various regulatory limits.
The principal sources of particulates would be fugitive dust emissions from construction activities
associated with the waste retrieval facility. The principal source of nitrogen dioxide, sulfur dioxide, and
carbon monoxide would be fuel combustion during the construction of the waste retrieval facility and
during retrieval of the waste. The following sections describe these sources in more detail.
G.1.7.1 Fugitive Dust from Construction of Retrieval Storage Facility
Construction activities such as earth moving and truck traffic would produce fugitive dust during the
construction of the retrieval storage facility and storage pad. The analysis used the same method to
estimate fugitive dust releases during retrieval as that for construction (see Section G. 1.4.1). The amount
of land disturbed to build the retrieval storage facility and storage pad would be 1 square kilometer
(250 acres) (TRW 1999a, Table 1-2, page 1-22). In addition, a 1.8-kilometer (1.1 -mile) rail line (TRW
1999a, page 1-16) would also be constructed. Assuming the rail line is 0.06 kilometer (0.04 mile) wide,
the rail line would require an additional 0. II square kilometer (27 acres) of land to be disturbed.
Table G-40 lists fugitive dust release rates from construction of the retrieval facility and storage pad.
Table G-4I lists air quality impacts as pollutant concentration in air and percent of regulatory limit.
Fugitive dust emissions from construction of the retrieval facility and storage pad would produce small
offsite (outside the land withdrawal area) PMio concentrations. Annual and 24-hour average
concentrations of PMio would be less than I percent for facility construction and about 2 percent for
storage pad construction of the regulatory standards for all three thermal load scenarios.
Table G-40. Fugitive dust release rates from surface construction of
retrieval storage facility and storage pad (PMip)."
Pollutant emission Emission rate
Period (kilograms)'' (grams per second)*^
Annual 25,000 per year 0.80
24-hour 100 per day 3^5^
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert grams per second to pounds per hour, multiply by 7.9366.
d. Based on an 8-hour release period.
Table G-41. Fugitive dust (PMio) air quality impacts from surface
construction of the retrieval storage facility and storage pad (micrograms
per cubic meter).
Maximum Regulatory Percent of
Period concentration^ limit'' regulatory limit"
Annual 0.096 50 0.19
24-hour 067 150 0.44
a. Numbers are rounded to two significant figures.
b. Sources: 40 CFR 50.4 through 50. 1 1 and Nevada Administrative Code 445B.391.
G.1.7.2 Exhaust from Construction Equipment
Surface equipment would emit the four criteria pollutants during retrieval and during the construction of
the retrieval storage facility and storage pad. The analysis used the same method to estimate pollutant
release rates from fuel consumed by construction equipment during retrieval that was used for the
construction phase (see Section G. 1.4.5). During retrieval, fuel would be consumed at the South Portal
G-26
Air Quality
Operations Area; during the construction of the retrieval facility and storage pad, fuel would be consumed
at the North Portal Operations Area. Table G-15 lists the pollutant release rates of the equipment that
would consume the diesel fuel. The maximum amount of fuel used annually would be about 1.46 million
liters (390,000 gallons) for surface construction (TRW 1999a, Table 1-2, page 1-22), about 1.7 million
liters (460,0(X) gallons) for surface retrieval operations (TRW 1999a, Table 1-3, page 1-24), and about
27,000 liters (7,200 gallons) for subsurface retrieval operations (TRW 1999b, page 6-33). Total
maximum annual usage would be about 1.9 million liters (5(X),000 gallons).
Table G-42 lists pollutant release rates for surface equipment during retrieval. Table G-43 lists the
potential air quality impacts. Emissions from surface equipment during retrieval would produce small
offsite criteria pollutant concentrations. All concentrations would be less than 1 percent of the regulatory
standards.
Table G-42. Pollutant release rates from surface equipment during the retrieval scenario.^
Mass of pollutant per Emission rate'^
Pollutant Period averaging time (kilograms)'' (grams per second)''
Nitrogen dioxide Annual 75,000 2.4
Sulfur dioxide Annual 7,100 0.22
24-hour 28 0.98
3-hour 11 0.98
Carbon monoxide 8-hour 110 4.0
1-hour 14 4.0
PMio Annual 6,600 0.21
24-hour 26 092
a. Numbers are rounded to two significant figures.
b. To convert kilograms to pounds, multiply by 2.2046.
c. Based on an 8-hour release period for averaging periods of 24 hour or less.
d. To convert grams per second to pounds per hour, multiply by 7.9366.
Table G-43. Air quality impacts from surface equipment during the retrieval scenario (micrograms per
cubic meter of pollutant).
Maximum Percent of
Pollutant Period concentration' Regulatory limit'' regulatory limit'
Nitrogen dioxide
Sulfur dioxide
Carbon monoxide
PM.o
Annual
0.23
100
0.24
Annual
0.022
80
0.028
24-hour
0.18
365
0.049
3-hour
1.4
1,300
0.11
8-hour
2.1
10,000
0.020
1-hour
13
40,000
0.033
Annual
0.021
50
0.042
24-hour
0.17
150
0.11
a. Numbers are rounded to two significant figures.
b. Sources: 40 CFR 50.4 through 50. 1 1 and Nevada Administrative Code 445B.391 .
G.2 Radiological Air Quality
This section describes the methods DOE used to analyze potential radiological impacts to air quality at
the proposed Yucca Mountain Repository during the construction, operation and monitoring, and closure
phases, and a possible retrieval scenario. The results are presented in Chapter 4, Section 4.1.2. It
discusses the radioactive noble gas krypton-85, which would be released from surface facilities during the
handling of spent nuclear fuel, and naturally occurring radon-222 and its radioactive decay products,
which would be released from the rock to the subsurface facility and then to the ventilation air. The
excavated rock pile would not be a notable additional source of radon-222, because the rock would not
have enhanced concentrations of uranium or radium (the sources of radon-222) in comparison to surface
G-27
Air Quality
rock. Somewhat higher concentrations of radon-222 could be present at the rock pile itself but, in
general, concentrations of radon-222 released from the excavated rock pile would not differ greatly from
naturally occurring surface concentrations of radon.
G.2.1 LOCATIONS OF HYPOTHETICALLY EXPOSED INDIVIDUALS AND POPULATIONS
Members of the public and noninvolved workers could be exposed to atmospheric releases of
radionuclides from repository activities. Doses to the maximally exposed individual and population
within 80 kilometers (50 miles) were evaluated for the public. The dose to the maximally exposed
noninvolved worker and the noninvolved worker populations at the repository and at the Nevada Test Site
were also evaluated.
Public
The location of the maximally exposed individual member of the public would be about 20 kilometers
(12 miles) south of the repository at the land withdrawal area boundary. This was determined to be the
location of unrestricted public access that would have the highest annual average concentration of
airborne radionuclides (see Section G.2.2). The locations calculated for nonradiological air quality
impacts (Section G.1.2) would be somewhat different because the analysis estimated exposure to
nonradiological pollutants for acute (short-term) exposures (1 to 24 hours) and for annual (continuous)
exposures.
Table G-44 lists the estimated population of about 28,000 within 80 kilometers (50 miles) of the
repository. This is the predicted population for 2000, based on projected changes in the region, including
the towns of Beatty, Pahrump, Indian Springs, and the surrounding rural areas. The population in the
vicinity of Pahrump was included in Table G-44 and evaluated for air quality impacts, even though the
Table G-44. Projected year 2000 population distribution within 80 kilometers (50 miles) of repository
site.^-"'^
Distance (kilometers)
Direction
8
16
24
32
40
48
56
64
72
80
Totals
S
0
0
16
238
430
123
0
10
0
0
817
SSW
0
0
0
315
38
0
0
7
0
0
360
SW
0
0
0
0
0
0
868
0
0
0
868
wsw
0
0
0
0
0
0
0
0
87
0
87
w
0
0
0
638
17
0
0
0
0
0
655
WNW
0
0
0
936
0
0
0
0
0
20
956
NW
0
0
0
28
2
0
0
0
33
0
63
NNW
0
0
0
0
0
0
0
0
0
0
0
N
0
0
0
0
0
0
0
0
0
0
0
NNE
0
0
0
0
0
0
0
0
0
0
0
ME
0
0
0
0
0
0
0
0
0
0
0
ENE
0
0
0
0
0
0
0
0
0
0
0
E
0
0
0
0
0
0
0
0
0
0
0
ESE
0
0
0
0
0
0
0
0
1,055
0
1.055
SE
0
0
0
0
3
0
13
0
0
206
222
SSE
0
0
0
0
23
172
6
17
6,117
16,399"
22,734
Grand Total
27,817
a. Source: 2000 population projected based on population data in TRW (1998, page 3-7).
b. To convert kilometers to miles, multiply by 0.62137.
c. There is a 4-kilometer (about 2.5-mile)-radius area around the North Portal, from which the analysis determined the
80-kilometer (50-mile) area.
d. Includes the Pahrump vicinity pwpulation, which extends beyond the 80-kilometer region.
G-28
Air Quality
population extends beyond the SO-kilometer region. The analysis calculated both annual population dose
and cumulative dose for the project phases over more than 100 years of construction, operation and
monitoring, and closure.
Noninvolved (Surface) Workers
The analysis assumed noninvolved workers on the surface would be at the site 2,000 hours a year (8 hours
a day, 5 days a week, 50 weeks a year), or about 23 percent of the total number of hours in a year (8,760).
All siuface workers, regardless of work responsibility, were considered to be noninvolved workers for
evaluation of exposure to radon-222 and radon decay products released from the subsurface facilities. For
releases of noble gases (principally krypton-85) from spent fuel handling activities, potentially exposed
noninvolved workers would be all surface workers except those in the Waste Handling and Waste
Treatment Buildings. The noble gases would be released from the stack of the Waste Handling Building
and workers in these facilities would not be exposed.
The maximally exposed noninvolved worker location would be in the South Portal Operations Area,
where air from repository development activities would be exhausted. The analysis assumed that this
worker would be in the office building about 1(X) meters (330 feet) northeast of the South Portal. This
worker would be exposed to the annual average concentration of radon during the construction phase as
radon concentrations increased with the increasing level of subsurface development. However, during
operational activities, the radon level would remain approximately constant at the baseline concentration
because the development area of the repository, ventilated and exhausted through the South Portal, would
remain relatively constant. There would be no South Portal ventilation during monitoring activities and
the closure phase, but the maximally exposed noninvolved worker would still be in the South Portal
Operations Area.
The population and distribution of repository workers required to staff the North Portal Operations Area
surface facilities would depend on the commercial spent nuclear fuel packaging scenario. As shown in
Table G-45, the uncanistered packaging scenario would have the highest labor requirements for all project
Table G-45. Noninvolved (surface) worker population distribution for Yucca Mountain activities
Packaging scenario
Worker location
Uncanistered
Disposable canister Dual
-purpose canister
Construction
North Portal
656
457
485
South Portal
70
70
70
Operation and monitoring
Emplacement and development
781"
630"
636"
North Portal
1,277
962
982
South Portal
70
70
70
Monitoring and maintenance
North Portal - decommissioning
1,354
982
1,023
North Portal - monitoring and maintenance
35
35
35
South Portal
6
6
6
Closure
North Portal
363
256
275
South Portal
6
6
6
Retrieval
North Portal - construction
780
780
780
North Portal - operations
108
108
108
South Portal
70
70
70
a. Sources: North Portal: TRW (1999a, pages 74, 75, and 79 to 81); South Portal: TRW (1999b. page 4-85).
b. Total workers exposed to krypton-85 releases from surface facilities. Does not include Waste Handling Building or Waste
Treatment Building workers; does include 70 workers at the South Portal.
G-29
Air Quality
phases and activities in comparison to the disposable canister and dual-purpose canister scenarios. The
number of North Portal workers would not vary for different thermal load scenarios. The estimated
population of workers in the South Portal Operations Area was based on the number of full-time
equivalents. This includes many workers who would be on the surface for only a portion of a day, as they
prepared for underground work in the surface operations area. The number of South Portal workers was
also assumed to remain constant for all thermal load scenarios.
Also evaluated as a potentially exposed noninvolved worker population were DOE workers at the Nevada
Test Site. The analysis used a Nevada Test Site worker population of 6,576 workers (DOE 1996,
Volume I, Appendix A, page A-69). For purposes of analysis, all these workers were assumed to be
about 50 kilometers (30 miles) east-southeast of the repository at Mercury, Nevada.
G.2.2 METEOROLOGICAL DATA AND ATMOSPHERIC DISPERSION FACTORS
The basis for the atmospheric dispersion factors used in the dose calculations was a joint frequency
distribution file for 1993 to 1997. These data were based on site-specific meteorological measurements
made at air quality and meteorology monitoring Site 1, combined for 1993 to 1997 (TRW 1999c, page
U). Site 1 is about 1 kilometer (0.6 mile) south of the proposed North Portal surface facility location.
Similar topographic exposure would lead to similar prevailing northerly and southerly winds at both
locations. DOE used these data because an analysis of the data collected at all the sites showed Site 1 to
be most representative of the surface facilities (TRW 1999c, page 7). The joint frequency data are
somewhat different from the more detailed meteorological data used for the nonradiological air quality
analysis. The dose calculations required only annual average data because they compare doses to annual
limits, whereas criteria pollutant limits have 1-, 3-, 8-, or 24-hoiu- averaging periods and the calculation of
short-term criteria pollutant concentrations required hourly meteorological data. The nonradiological
analysis also calculated concentrations only at the land withdrawal area boundary, not at onsite locations
where workers would be.
Depending on the project phase and level of activity, subsurface ventilation air could be exhausted from
any or all of three locations: the South Portal, emplacement (exhaust) shaft 1 or emplacement (exhaust)
shaft 2. Both of these exhaust shafts would be on the ridge above the repository. Table G-46 lists the
distribution of exhaust ventilation air among the three subsurface release points for project phases and
activities. These distributions were used to calculate annual average atmospheric dispersion factors for
radon releases from the subsurface.
The GENII software system (Napier et al. 1997, all) was used to calculate annual average atmospheric
dispersion factors for radon released from the subsurface exhaust points and for noble gases released fron
the Waste Handling Building stack. The releases from the South Portal would be at ground level, while
releases from the two emplacement shafts (ES-1 and ES-2) on the ridge above the repository were
modeled as 60-meter (200-foot) releases. Noble gas releases from the Waste Handling Building would 1
from a 60-meter (200-foot) stack, also modeled as an elevated release. The population distribution data in
Tables G-44 and G-45 were used to calculate population-weighted dispersion factors for public and
noninvolved worker populations, which were then used to calculate collective doses. Table G-47 lists the]
individual and population-weighted atmospheric dispersion factors for the radon and krypton-85 release
points at the site. These values do not incorporate the release distribution data in Table G-46. The radon
dispersion factors would vary slightly among some combinations of project phase and thermal load
scenarios because of the slight differences in release point contributions noted in Table G-46. Krypton-85
dispersion factors would not be affected.
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Air Quality
Table G-46. Distribution (percent) of repository subsurface exhaust ventilation air.^
Project phase and activity
Thermal load
scenario
South Portal
Emplacement
(exhaust) shaft
1
Emplacement
(exhaust) shaft
2
I Proposed Action
Construction
Operation and monitoring
Development and emplacement
Monitoring and maintenance
Closure
Retrieval scenario
Inventory Modules 1 and 2
Construction
Operation and monitoring
Development and emplacement
Monitoring and maintenance
Closure
All
100
High
47
53
Intermediate
47
53
Low
55
42
All
100
Same exhaust distribution as monitoring and maintenance
Same exhaust distribution as monitoring and maintenance
All
100
High
46
54
Intermediate
39
61
Low
42
40
High
100
Intermediate
100
Low
50
18
50
Same exhaust distribution as monitoring and maintenance
a. Source: Rasmussen (1998, all); TRW (1999b, pages 4-33 to 4-48).
G.2.3 RADIOLOGICAL SOURCE TERMS
There would be two distinctly different types and sources of radionuclides released to the air from
activities at the repository. Naturally occurring radon-222 and its radioactive decay products would be
released from the subsurface facility during all phases as the repository ventilation system removed
airborne particulates from development operations and exhausted air heated by the emplaced materials.
Radioactive noble gases would be released from commercial spent nuclear fuel during handling and
transfer operations in the surface facilities during the operation and monitoring phase. Section G.2.3. 1
discusses the releases of radon-222 and radon decay products. Section G.2.3.2 discusses the releases of
radioactive noble gases from commercial spent nuclear fuel.
G.2.3.1 Release of Radon-222 and Radon Decay Products from the Subsurface Facility
In the subsurface facility the noble gas radon-222 would diffuse continually from the rock into the air of
the repository drifts. Radioactive decay of the radon in the air of the drift would produce radon decay
products, which would begin to come into equilibrium (having the same activity) with the radon-222
because their radioactive half-lives are much shorter than the 3.8-day half-life of radon-222. Key
radionuclide members of the radon-222 decay chain are polonium-218 (sometimes known as radium A)
and polonium-214 (radium C), with half-lives of 3.05 minutes and 164 microseconds, respectively.
Exhaust ventilation would carry the radon-222 and the radon decay products from the repository.
The estimates of radon-222 and radon decay product releases were based on concentration observations
made in the Exploratory Studies Facility subsurface areas during site characterization. Because the
repository would encompass the subsurface areas of the Exploratory Studies Facility, the analysis
assumed that these observations would be a reasonable baseline. Concentrations at the 7,350-meter
(4.6-mile) measuring station in the South Ramp ranged from 0.65 to 163 picocuries per liter with the
ventilation system operating (TRW 1999c, electronic file attachment 7350EBF.XLS). The measured
50th-percentile concentration was 24 picocuries per liter, with 5th- and 95th-percentile concentrations of
1.7 and 124 picocuries per liter, respectively. Because the distribution of these concentration data was
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Air Quality
Table G-47. Atmospheric dispersion factors for potentially exposed individuals and populations from
releases at the repository site."
Release Receptor Dispersion
type*^ type Receptor location factor''
Release location
Radon releases'
Public
South Portal
South Portal
Emplacement shafts 1, 2^
Emplacement shafts 1 , 2^
Noninvolved workers
South Portal
South Portal
South Portal
South Portal
Emplacement shaft 1
Emplacement shaft 1
Emplacement shaft 2
Emplacement shaft 2
Emplacement shafts 1, 2^
Krypton-85 releases
Public
Waste Handling Bldg. stack
Waste Handling Bldg. stack
Noninvolved workers
Waste Handling Bldg. stack
Waste Handling Bldg. stack
Waste Handling Bldg. stack
Waste Handling Bldg. stack
Waste Handling Bldg. stack
Waste Handling Bldg. stack
G
G
E
E
G
G
G
G
E
E
E
E
E
E
E
E
E
E
E
E
E
individual 20 km*^ south
population 80 km radius
individual 20 km south
population 80 km radius
individual 100 meters'" northeast
population South Portal Operations Area
individual North Portal 2.8 km north-northeast*
individual Nevada Test Site, 50 km east-southeast'
individual North Portal 4.2 km southeast
individual South Portal 6.3 km south-southeast
individual North Portal, 4.5 km east-southeast
individual South Portal, 5.3 km southeast
individual Nevada Test Site, 50 km east-southeast
individual 20 km south
population 80 km radius
individual North Portal, 0.4 km north-northwest
individual South Portal, 2.8 km south-southwest
population Uncanistered packaging scenario
population Disposable canister packaging scenario
population Dual-purpose canister packaging scenario
individual Nevada Test Site, 50 km east-southeast'
2.2x10'
1.2x10"^
6.0x10"^
3.0x10-
6.2x10"'
3.2x10"^
1.9x10"^
6.9x10'
9.0x10'
2.0x10"*
4.9x10"'
6.7x10"'
2.7x10"'
6.0x10''
3.0x10"'
1.5x10"*
5.4x10"*
2.4x10"*
1.9x10""
1.9x10""
2.7x10"'"
a. Numbers are rounded to two significant figures.
b. Source: Radon releases: TRW (1999b, pages 4-33 to 4-48); krypton-85 releases: TRW (1999a, page 41).
c. G = ground level; E = elevated.
d. Dispersion factor units are seconds per cubic meter for individuals, and person-seconds per cubic meter for populations.
e. Radon includes radon-222 and its radioactive decay products.
f. To convert kilometers to miles, multiply by 0.62137.
g. Difference in dispersion between the two emplacement shafts is small for this application,
h. To convert meters to feet, multiply by 3.2808.
i. The population dose was calculated at this point by multiplying the individual dispersion factor times population size.
highly skewed, the analysis assumed that the 50th-percentile value was most representative of the entire
concentration range.
Exhaust ventilation flowrates in the South Ramp when the radon concentration measurements were made
measured from about 100 to 125 cubic meters per second (214,000 to 265,000 cubic feet per minute)
(TRW 1999c, electronic file attachment DECRPT.XLS). A value of 1 10 cubic meters per second
(230,000 cubic feet per minute) was used as a representative South Ramp flowrate. This information,
combined with an Exploratory Studies Facility excavated volume of 360,0(X) cubic meters (470,000 cubic
yards) (TRW 1999b, page 4-27), yielded a calculated repository air exchange rate of about 1 per 3,300
seconds (about one exchange per hour) and a baseline for radon-222 releases. The exchange rate is the
excavated volume (in cubic meters) divided by the ventilation flowrate (in cubic meters per second). The
analysis assumed these conditions would be representative for the Exploratory Studies Facility through
the beginning of the construction phase. The estimated release of radon-222 and radon decay products for
this configuration would be about 80 curies per year.
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Air Quality
Table G-48 lists the key input parameters, namely the beginning and ending excavated repository
volumes, repository average ventilation rates, and repository average air exchange rates, for each of the
phases and thermal load scenarios of the Proposed Action. The analysis assumed that increases in
excavated repository volume and ventilation flowrate would occur linearly. In addition. Table G-48 lists
the estimated releases of radon-222 and radon decay products annually and by phase.
Table G-48. Estimated radon-222 releases for repository
activities for the
Proposed Action
inventory."
Average
Repository '
volume
ventilation
Annual
(millions of cubic
rate (cubic
average radon''
Total radon""
pprioH and meters]
!"•
meters per
second)
Average air
exchange rate
release
release
thermal load Beginning
Ending
(curies)
(curies)
Construction (5 years)
High 0.36
1.9
205
6,200
300
1,500
Intermediate 0.36
2.2
205
7,200
340
1,700
Low 0.36
2.2
205
7,200
340
1,700
Operations (24 years)
High 1.9
4.7
570
6,700
880
21,000
Intermediate 2.2
5.7
570
7,900
1,000
25,000
Low 2.2
14
680
13,000
1,900
46,000
Monitoring (76 years)
High 4.7
4.7
190
24,000
1,100
83,000
Intermediate 5.7
5.7
190
29,000
1,300
99,000
Low 14
14
490
28,000
3,200
240,000
Total Operation and Monitoring Phase (100 years)
High
1,000
100,000
Intermediate
1,200
120,000
Low
2,900
290,000
Closure phase (6, 6, and 15 years)
High 4.7
4.7
190
24,000
1,100
6,600
Intermediate 5.7
5.7
190
29,000
1,300
7,900
Low 14
14
490
28,000
3,200
48,000
Total, all phases (111, 111. 120 years)
High
110,000
Intermediate
130,000
Low
340,000
Retrieval scenario (14 years)
High 4.7
4.7
190
24,000
1,100
14,000
a. Numbers are rounded to two significant figures; totals might not (
squal sums of values due to rounding.
b. Source: TRW (1999b, pages 4-27, 6-6, and 6-16).
c. To convert cubic meters to cubic yards, multiply by 1.3079.
d. Includes radon-222 and radon decay products.
Construction Phase
During the 5 years of construction, 1.5 million cubic meters (1.96 million cubic yards) of rock would be
removed for the high thermal load scenario and 1.9 million cubic meters (2.4 million cubic yards) for the
intermediate and low thermal load scenarios (TRW 1999b, page 6-6). During the same period, the
ventilation flow would increase from 1 10 cubic meters per second (230,(X)0 cubic feet per minute) to 270
cubic meters per second (570,000 cubic feet per minute) (TRW 1999b, pages 4-33 to 4-38). Releases of
radon-222 would be low but would vary within 15 percent among all three thermal load scenarios,
because they would have the same ventilation flow rates but different repository volumes.
Operation and IVIonitoring Phase
Operation Activities. Development activities would last 22 years during operation and monitoring.
During this period about 2.9 million, 3.4 million, and 11.8 million cubic meters (3.8 million, 4.5 million,
and 15.4 million cubic yards) of rock would be removed for the high, intermediate, and thermal load
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Air Quality
scenarios, respectively (TRW 1999b, page 6-16). The repository excavation would be complete during
the last two years of the operation activity period, as emplacement activities continued. The flowrate for
the repository during emplacement and development activities of the high and intermediate thermal load
scenarios would be the maximum development side flowrate [270 cubic meters per second (570,000 cubic
feet per minute)], and the maximum emplacement side flowrate [300 cubic meters per second (640,000
cubic feet per minute)] (TRW 1999b, pages 4-33 to 4-38). The flowrate during the low thermal load
scenario would vary from 570 to 740 cubic meters per second (1.2 million to 1.6 million cubic feet per
minute), depending on the stage of emplacement activities.
The estimation of radon releases for the high and intermediate thermal load scenarios was based on
development and emplacement activities taking place only in the upper (primary) block. However, for
the low thermal load scenario development and emplacement would be incremental, beginning in the
upper block, moving on to the lower block, and finally to the Area 5 block (TRW 1999b, page 3-3).
When emplacement in a block was complete, that block would enter an interim period of monitoring and
maintenance as activities continued in the other blocks. The analysis assumed that the upper block would
be in this interim status for 10 years and the lower block for 5 years.
The high and intermediate thermal load scenarios would have the lowest radon releases because they
would use only the upper (primary) block. The low thermal load scenario would have a higher radon
release because of the greater repository volume, which would require three blocks, and the added
contribution from exhaust ventilation during the interim monitoring and maintenance of the upper and
lower blocks.
Monitoring Activities. No excavation would take place during monitoring, and the exhaust flowrate
would remain constant. The much greater repository volume for the low thermal load scenario, which
would require larger exhaust flowrates, would result in larger releases of radon-222 and radon decay
products to the atmosphere through the exhaust ventilation.
Monitoring and maintenance activities would last from 26 to 276 years. Total releases of radon over
26 years would be approximately 29,000, 34,000, and 84,000 curies for the high, intermediate, and low
thermal load scenarios, respectively. Total releases of radon over 276 years would be approximately
300,000, 360,000, and 890,000 curies for the high, intermediate, and low thermal load scenarios,
respectively. The estimated annual radon release and concentration would be the same as those listed for
monitoring in Table G-48.
For 100 years of operation and monitoring, the low thermal load scenario would involve approximately
2.5 times more radon release than the high or intermediate thermal load scenario. About 70 to 75 percent|
of the radon would be released during the monitoring and maintenance period for all three thermal load
scenarios, not including the interim monitoring and maintenance for the low thermal load scenario.
Closure Phase
Annual releases of radon-222 and radon decay products during the closure phase would be the same as fo
the monitoring period. Differences in the lengths of the closure phases for the three thermal load
scenarios would lead to differences in the total amount of radon released. Differences among the thermal|
load scenarios would be for the same reasons as for the monitoring period, namely the larger repository
volume and exhaust ventilation flowrate of the low thermal load scenario.
Retrieval
Only the high thermal load scenario was evaluated for a postulated retrieval scenario. Annual releases of
radon-222 and radon decay products would be the same as for the monitoring activities and closure
phases. Releases were estimated for 13 years, including 2 years of retrieval-related construction activities
plus 1 1 years of retrieval operations.
G-34
Air Quality
Inventory Modules 1 and 2
Releases of radon-222 and radon decay products for Inventory Modules 1 and 2 were estimated using the
same methods as for the Proposed Action. The major differences would be the larger repository volumes
and higher ventilation flowrates, which would result in larger releases of radon. In addition, 38 years
would be required to complete operations (which includes 36 years of development), 62 years would be
required for monitoring, and the closure phase would be longer. Table G-49 lists the estimates of radon
release and key parameter values. Releases of radon would be higher for the inventory modules than for
the Proposed Action in all cases.
Table G-49. Estimated radon-222 releases for repository activities for Inventory Modules 1 or 2.
Repository volume
Average
Average
Annual
Total
(millions of cubic
ventilation rate
air
average
radon
meters
)"■'
_ (cubic meters
per second)
exchange
rate(s)
radon release
(curies)
release
Thermal load
Beginning
Ending
(curies)
Construction (5 years)
High
0.36
2.1
205
6,900
330
1,600
Intermediate
0.36
2.1
205
6,900
330
1,600
Low
0.36
2.1
205
6,900
330
1,600
Operations (38 years)
High
2.1
8.7
590
9,500
1,300
49,000
Intermediate
2.1
9.0
690
8,200
1,300
51,000
Low
2.1
24
800
16,000
3,100
120,000
Monitoring (62 years)
High
8.7
8.7
300
29,000
2,000
125,000
Intermediate
9.0
9.0
490
18,000
2,100
130,000
Low
24
24
890
27,000
5,500
340,000
Total operation and monitoring phase (100 years)
High
1,700
170,000
Intermediate
1,800
180,000
Low
4,600
460,000
Closure (13, 17, and 27 years)
High
8.7
8.7
300
29,000
2,000
26,000
Intermediate
9.0
9.0
490
18,000
2,100
35,000
Low
24
24
890
27,000
5,500
150,000
Totals (118. 122, and 132 years)
High
200,000
Intermediate
220,000
Low
610,000
a. Numbers are rounded to two significant figures; totals might not equal sums of values due to rounding.
b. Source: TRW (1999b, pages 4-27, 6-47, and 6-55).
c. To convert cubic meters to cubic yards, multiply by 1 .3079.
G.2.3.2 Release of Radioactive Noble Gases from the Surface Facility
The unloading and handling of commercial spent nuclear fuel would produce the only routine emissions
of manmade radioactive materials from repository facilities. No releases would occur as a result of
emplacement activities. Shipping casks containing uncanistered spent nuclear fuel in dual-purpose
canisters would be opened in the transfer pool of the Waste Handling Building at the North Portal
Operations Area. Shipping casks containing spent nuclear fuel in disposable canisters would be opened in
a dry transfer cell. During spent fuel handling and transfer, radionuclides could be released from a small
percentage of fuel elements with pinhole leaks in the fuel cladding; only noble gases would escape the
pool and enter the ventilation system of the Waste Handling Building (TRW 1999a, page 17). The largest
release of radionuclides firom surface facilities would be krypton-85, with about 2,600 curies released
annually from the uncanistered and dual-purpose canister packaging options. Krypton-85 would also be
the major dose contributor from the airborne pathway. Releases of other noble gas radionuclides would
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Air Quality
be very small, with estimated annual releases of about 0.0000010 curie of krypton-81, 0.000033 curie of
radon-219, 0.014 curie of radon-220, 0.0000046 curie of radon-222, and small quantities of xenon-127
(TRW 1999a, page 75). The same annual releases would occur for both the Proposed Action and for the
inventory modules. Table G-50 lists estimated annual average releases of krypton-85 from fuel handling
by packaging option. All spent nuclear fuel and DOE high-level radioactive waste in disposable canisters
would be transferred from shipping casks to disposal containers inside shielded rooms (hot cells) in the
Waste Handling Building. Because all DOE material would be in disposable canisters under all
packaging scenarios, no radionuclide releases from these materials would occur.
Table G-50. Krypton-85 releases (curies) from surface facility handling activities for commercial
spent nuclear fuel during the operation and monitoring phase. °
Packaging option
Annual release
Proposed Action
(24 years)
Inventory Module 1 or 2
(38 years)
Uncanistered
Disposable canister
Dual-purpose canister
2,600
90
2,600
61,000
2,200
62,000
97,000
3,500
98,000
a. Numbers are rounded to two significant figures.
b. Source: TRW (1999a, page 75).
Releases from the surface facility would be the same for the three thermal load scenarios. These releases
were based on the following assumptions for commercial spent nuclear fuel (TRW 1999a, pages 18
and 19):
• Pressurized-water reactor bumup of about 40 gigawatt-days per metric ton of uranium with
3.6-percent enrichment and an average of 26 years decay
• Boiling-water reactor bumup of 32 gigawatt-days per metric ton of uranium with 3.0-percent
enrichment and an average of 27 years decay
• A failure rate of 0.25 percent for fuel assemblies in the canisters, allowing gaseous radionuclides
(isotopes of krypton, radon, and xenon) to escape
• Radionuclides other than noble gases (such as cobalt-60, cesium-137, and strontium-90) would not
escape the transfer pool if released from fuel assemblies
G.2.4 DOSE CALCULATION METHODOLOGY
The previous three sections provided information on the location and distribution of potentially affected
individuals and populations (Section G.2.1), atmospheric dispersion (Section G.2.2), and the type and
quantity of radionuclides released to air (Section G.2.3) in the Yucca Mountain region. The analysis used
these three types of information to estimate the radionuclide concentration in air (in picocuries of
radionuclide per liter of air) at a specific location or for an area where there would be a potentially
exposed population. The estimation of the radiation dose to exposed individuals or populations from
concentrations of radionuclides in air used this information and published or derived dose factors. This
section describes the concentration-to-dose conversion factors that the analysis used to estimate radiation
dose to members of the public and noninvolved workers from releases of radionuclides at the repository.
G.2.4.1 Dose to the Public
The analysis estimated doses to members of the public using screening dose factors from the National
Council on Radiation Protection and Measurements (NCRP 1996, Volume I, pages 113 and 125). The
analysis considered all exposure pathways, including inhalation, ingestion, and direct external radiation
from radionuclides in the air and on the ground. For noble gases such as krypton-85, only direct external
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Air Quality
exposure from the radionuclides in the air would be a contributing pathway. For radon-222, the short-
lived decay products would account for essentially all of the dose. The screening dose factors indicate
that direct external radiation from radionuclides deposited on the ground would account for about 40
percent of the dose; ingestion of these decay products in foodstuffs and inadvertently consumed soil
would account for about 60 percent, based on the published screening dose factors. Inhalation and
external irradiation from radionuclides in the air would be minor exposure pathways. The analysis
calculated the estimated dose from a specific radionuclide by multiplying the radionuclide-specific dose
factor by the estimated air concentration at the exposure location. The results are reported in Chapter 4,
Section 4.1.2. Table G-51 lists the screening dose factors for krypton-85 and radon-222 for members of
the public. Results are presented in Chapter 4, Section 4.1.2.
Table G-51. Factors for estimating dose to the public and
noninvolved workers per concentration of radionuclide in air
(millirem per picocurie per liter per hour) for krypton-85 and radon-
222."'''
Radionuclide Public*^ Noninvolved worker
Krypton-85 0.0000013 0.0000013
Radon-222 025^ 0.029'
a. Numbers are rounded to two significant figures.
b. Dose factors for radon-222 include dose contribution from decay products.
c. Source: NCRP (1996, page 61); assumed an exposure time of 8,000 hours
per year.
d. Includes all exposure pathways.
e. Source: ICRP (1994, pages 5 and 24); 100 percent equilibrium between
radon and decay products; inhalation pathway only.
G.2.4.2 Dose to Noninvolved Workers
The analysis used a National Council on Radiation Protection and Measurements screening dose factor to
calculate doses to noninvolved workers from krypton-85 because the exposure pathway is simple (air
submersion only) and is the same as for members of the public. Table G-5 1 also lists this factor.
However, the analysis did not use a National Council on Radiation Protection and Measurements
screening dose factor to estimate the dose to noninvolved workers from radon-222 and its decay products.
The parameters and exposure scenarios used to derive the National Council on Radiation Protection and
Measurements screening dose factors for radon-222 and its decay products would not be appropriate for
the potential exposure scenario for noninvolved workers at the Yucca Mountain site. Dose to
noninvolved workers on the surface would be due mainly to inhalation of the radon decay products, and
not from the other exposure pathways noted above for the public. Therefore, the analysis developed a
Yucca Mountain repository-specific exposure scenario using site-specific parameters where appropriate.
The dose conversion factor is from Publication 65 of the International Commission on Radiological
Protection (ICRP 1994, page 24). This dose factor, which is 0.5 rem per working level month for
inhalation of radon decay products by workers, corresponds to 0.029 millirem per picocurie per liter per
hour, with radon decay products in 100 percent equilibrium (equilibrium factor of 1.0) with the radon-
222 parent (ICRP 1994, page 5).
In estimating dose from radon and radon decay products released firom the subsurface facility, the analysis
assumed the maximally exposed noninvolved worker would be in an office about 1(X) meters (330 feet)
northeast of the South Portal. For the construction phase and development activities, the noninvolved
worker exposure analysis used the distribution of radon concentration measurements made at the
7,350-meter (4.6-mile) station in the South Ramp of the Exploratory Studies Facility. These were the best
available data for estimating releases of radon from the facility (TRW 1999c, page 12). There would be
no releases from the South Portal during the other project phases. Measured concentrations ranged from
0.65 to 163 picocuries per liter, with a median value of 24 picocuries per liter, as noted in Section G.2.3.1.
In addition, the analysis considered the distribution of the measured values of the equilibrium fraction
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Air Quality
between radon-222 and the decay products. This value ranged from 0.0022 to 0.44, with a median of 0. 14
(TRW 1999c, electronic file attachment RNFBF.XLS). The annual average atmospheric dispersion factor
from the South Portal to the office building would be approximately 6.2 x 10'^ seconds per cubic meter
for both the construction phase and development activities (Table G-47), although differences in exhaust
flowrate (205 and 269 cubic meters per second, respectively, would result in minor differences in
dispersion. The analysis assumed the maximally noninvolved worker would be exposed from 1,600 to
2,000 hours per year.
The estimated median dose to a maximally exposed noninvolved worker during the construction phase
would be approximately 5 (4.7 to 5.4) millirem per year. The dose from the Proposed Action
intermediate and low thermal load scenarios would be somewhat higher than that from the high thermal
load scenario because of the larger average repository volume for these two scenarios during the
construction phase (Table G-48). The estimated 5th-percentile dose would be about 0.2 millirem per year
for both cases and the 95th-percentile dose would be 42 and 48 millirem per year, respectively. The dose
during development activities would be the same for all three thermal load scenarios, with a median dose
of about 3.4 millirem per year. The estimated 5th-percentile dose would be about 0.2 millirem per year
and the 95th-percentile dose about 29 millirem per year. These estimates were made using a Monte Carlo
uncertainty analysis. There would be a small contribution from external radiation, but the analysis did not
consider it because it would be indistinguishable from normal external background radiation. The
estimated dose from Module 1 or 2 would be about the same as those for the intermediate and low
thermal load scenarios.
During the construction phase the maximally exposed noninvolved worker would receive a somewhat
larger potential dose because of a larger average repository volume, which would be exhausted through
the South Portal, and additional radon release. During operations the ventilation systems for the
subsurface development and emplacement areas would be separate. The analysis assumed that the
volume during Exploratory Studies Facility operations would represent the volume of the development
side exhausted through the South Portal. This volume is somewhat smaller than the estimated average
construction phase repository volume.
REFERENCES
Cowherd, Muleski, and Kinsey
1988
DOE 1995
DOE 1996
DOE 1997a
Cowherd, C, G. E. Muleski, and J. S. Kinsey, 1988, Control of Open
Fugitive Dust Sources, Final Report, pp. 4.1 to 5.41, EPA-450/3-88-
008, Midwest Research Institute, Kansas City, Missouri. [243438]
DOE (U.S. Department of Energy), 1995, Department of Energy
Programmatic Spent Nuclear Fuel Management and Idaho National
Engineering Laboratory Environmental Restoration and Waste
Management Programs: Final Environmental Impact Statement,
DOE/EIS-0203-F, Office of Environmental Management, Idaho
Operations Office, Idaho Falls, Idaho. [102617]
DOE (U.S. Department of Energy), 1996, Final Environmental Impact
Statement for the Nevada Test Site and Off-Site Locations in the State
of Nevada, DOE/EIS-0243-F, Nevada Operations Office, Las Vegas,
Nevada. [239895]
DOE (U.S. Department of Energy), 1997a, Waste Isolation Pilot Plant
Disposal Phase Final Supplemental Environmental Impact Statement,
DOE/EIS-0026-S-2, Carlsbad Area Office, Carlsbad, New Mexico.
[238195]
G-38
Air Quality
DOE 1997b
DOE 1997c
DOE 1998
EPA 1987
EPA 1988
EPA 1991
EPA 1995a
EPA 1995b
EPA 1996
Fransioli 1999
ICRP 1994
DOE (U.S. Department of Energy), 1997b, Final Waste Management
Programmatic Environmental Impact Statement for Managing
Treatment, Storage, and Disposal of Radioactive and Hazardous
Waste, DOE/E1S-0200-F, Office of Environmental Management,
Washington, D.C. [232988]
DOE (U.S. Department of Energy), 1997c, Yucca Mountain Site
Characterization Project - Map for Contaminated Areas, map, YMP-
97-022.0, Office of Civilian Radioactive Waste Management, Yucca
Mountain Project Office, Las Vegas, Nevada. [MOL. 199905 1 1 .0292]
DOE (U.S. Department of Energy), 1998, Air Quality Control Design
Analysis, BCADOOOOO-017 17-0200-00008, Revision 00, Office of
Civilian Radioactive Waste Management, Washington, D.C.
[MOL. 19980729.0044]
EPA (U.S. Environmental Protection Agency), 1987, On-Site
Meteorological Program Guidance for Regulatory Modeling
Applications, Wordperfect® reissue of the June 1987 EPA document,
EPA-450/4-87-013, Office of Air Quality Planning and Standards,
Office of Air and Radiation, Research Triangle Park, North Carolina.
[210292]
EPA (U.S. Environmental Protection Agency), 1988, Gap Filling
PMio Emission Factors for Selected Open Area Dust Sources, EPA-
450/4-88-003, Midwest Research Institute, Kansas City, Missouri.
[243553]
EPA (U.S. Environmental Protection Agency), 1991, Compilation of
Air Pollutant Emission Factors, Volume II: Mobile Sources, AP-42,
Supplement A, Washington, D.C. [243439]
EPA (U.S. Environmental Protection Agency), 1995a, User's Guide
for Industrial Source Complex (ISC3) Dispersion Models, EPA-454/B-
95-003a, Emissions, Monitoring, and Analysis Division, Office of Air
Quality Planning and Standards, Research Triangle Park, North
Carolina. [243563]
EPA (U.S. Environmental Protection Agency), 1995b, Compilation of
Air Pollutant Emission Factors, Fifth Edition, AP-42, Volume I:
Stationary Point and Area Sources, Research Triangle Park, North
Carolina. [226367]
EPA (U.S. Environmental Protection Agency), 1996, Ambient Levels
and Noncancer Health Effects of Inhaled Crystalline arui Amorphous
Silica: Health Issue Assessment, EPA/600/R-95/1 15, National Center
for Environmental Assessment, Office of Research and Development,
Washington, D.C. [243562]
Fransioli, P., 1999, 'Telephone Log for Number of Days with
Precipitation Greater Than 0. 1 Inches," internal personal
communication with C. Fosmire, February 4, TRW Environmental
Safety Systems Inc., Las Vegas, Nevada. [MOL. 199905 11.0282]
ICRP (International Commission on Radiological Protection), 1994,
Protection Against Radon-222 at Home and at Work, Publication 65,
Pergamon Press, Oxford, Great Britain. [236754]
G-39
Air Quality
lessen 1998
Napier et al. 1997
NCRP 1996
Rasmussen 1998
Seinfeld 1986
Smith 1999
TRW 1998
TRW 1999a
TRW 1999b
TRW 1999c
lessen, J., 1998, "Additional Land Disturbance at Yucca Mountain
from Repository Construction (Base Case and Extended Inventory),"
internal memorandum, July 23, Jason Technologies Corporation, Las
Vegas, Nevada. [MOL.19990602.0181]
Napier, B. A., D. L. Strenge, R. A. Peloquin, J. V. Ramsdell, and P. D.
Rittmann, 1997, RSICC Computer Code Collection, GENII 1.485,
Environmental Radiation Dosimetry Software System, CCC-601, PNL-
6584, Radiation Safety Information Computational Center, Oak Ridge
National Laboratory, Hanford, Washington. [206898]
NCRP (National Council on Radiation Protection and Measurements),
1996, Screening Models for Releases of Radionuclides to Atmosphere,
Surface Water, and Ground, Recommendations of the National
Council on Radiation Protection and Measurements, Report No. 123,
Bethesda, Maryland. [225158, Volume 1; 234986, Volume 2]
Rasmussen, D. G., 1998, "More Questions on Repository Ventilation,"
electronic communication to T. Dcenberry (Dade Moeller &
Associates), September 23, TRW Environmental Safety Systems Inc.,
Las Vegas, Nevada. [MOL. 1 99905 1 1 .0300]
Seinfeld, J. H., 1986, Atmospheric Chemistry arul Physics of Air
Pollution, pp. 26-31, John Wiley and Sons, Inc., New York, New
York. [243754]
Smith, A., 1999, 'Telephone Log for Disturbed Area of Muck Pile in a
Given Year," personal communication with C. Fosmire (PNNL),
February 4, Argonne National Laboratory, Argonne, Illinois.
[MOL. 199905 11.0283]
TRW (TRW Environmental Safety Systems Inc.), 1998, Yucca
Mountain Site Characterization Project: Summary of Socioeconomic
Data Analyses Conducted in Support of the Radiological Monitoring
Program, April 1997 to April 1998, Las Vegas, Nevada.
[MOL. 19980803.0064]
TRW (TRW Environmental Safety Systems Inc.), 1999a, Repository
Surface Design Engineering Files Report, BCBOOOOOO-017 17-5705-
00009, Revision 03, Las Vegas, Nevada. [MOL. 19990615.0238]
TRW (TRW Environmental Safety Systems Inc.), 1999b, Engineering
File - Subsurface Repository, BCAOOOOOO-01717-5705-00005,
Revision 02 with DCNl, Las Vegas, Nevada. [MOL. 19990622.0202,
document; MOL. 19990621.0157, DCNl]
TRW (TRW Environmental Safety Systems Inc.), 1999c,
Environmental Baseline File for Meteorology and Air Quality,
BOOOOOOOO-017 17-5705-00126, Revision 00, Las Vegas, Nevada.
[MOL.19990302.0186]
G-40
Appendix H
Potential Repository Accident
Scenarios: Analytical Methods
and Results
Potential Repository Accident Scenarios: Analytical Methods and Results
Section
H.1
H.2
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.1
H.2.2
H.3
H.4
H.5
References
1
1.1
1.2
1.3
1.4
1.5
2
3
4
4.1
4.1.1
4.1.2
4.1.3
4.1.4
4.2
4.3
4.4
4.5
5
TABLE OF CONTENTS
Page
General Methodology H-1
Potential Repository Accident Scenarios H-2
Radiological Accident Scenarios H-2
Internal Events - Waste Handling Building H-2
Cask/Carrier Transport and Handling Area H-5
Canister Transfer System H-5
Assembly Transfer System H-5
Disposal Container Handling System H-6
Waste Emplacement and Subsurface Facility Systems '. H-6
Internal Events - Waste Treatment Building H-7
External Events H-8
Source Terms for Repository Accident Scenarios H-16
Commercial Spent Nuclear Fuel Drop Accident Scenario Source Term H-17
Crud H-17
Fuel Rod Gap H-19
Fuel Pellet H-20
Conclusions H-21
Transporter Runaway and Derailment Accident Source Term H-21
DOE Spent Nuclear Fuel Drop Accident Source Term H-22
Seismic Accident Scenario Source Term H-22
Low-Level Waste Drum Failure Source Term H-25
Assessment of Accident Scenario Consequences H-27
Nonradiological Accident Scenarios H-28
Accident Scenarios During Retrieval H-3I
Accident Scenarios During Monitoring and Closure H-32
Accident Scenarios for Inventory Modules 1 and 2 H-32
H-32
Table
H-1
H-2
H-3
H-4
H-5
H-6
H-7
H-8
LIST OF TABLES
Page
Bounding internal accident scenarios for the Waste Handling Building and
emplacement operations H-4
External events evaluated as potential accident initiators H-9
Typical commercial spent nuclear fuel characteristics H-17
Inventory used for typical reactor fuel H-18
Source term parameters for commercial spent nuclear fuel drop accident scenarios H-21
Source term used for N-Reactor Mark FV fuel drop accident scenario analysis H-23
Radiological consequences of repository operations accidents for median
meteorological conditions H-29
Radiological consequences of repository operations accidents for unfavorable
meteorological conditions H-30
LIST OF FIGURES
Figure
H-1 Integrated seismic hazard results: summary hazard curves for peak horizontal
acceleration
Page
.H-15
H-iii
Potential Repository Accident Scenarios: Analytical Methods and Results
APPENDIX H. POTENTIAL REPOSITORY ACCIDENT SCENARIOS:
ANALYTICAL METHODS AND RESULTS
This appendix describes the methods and detailed results of the analysis the U.S. Department of Energy
(DOE) performed for the Yucca Mountain Repository Environmental Impact Statement (EIS) to assess
radiological impacts from potential accident scenarios at the proposed repository. The methods apply to
repository accidents that could occur during preclosure only, including operation and monitoring,
retrieval, and closure. In addition, this appendix describes the details of calculations for specific accidents
that the analysis determined to be credible. Appendix J describes the analytical methods and results for
accidents that could occur at the 72 commercial and 5 DOE sites and during transportation to the
proposed repository.
The accident scenarios in this analysis, and the estimated impacts, are based on current information from
the repository design (TRW 1999a, all). The results are based on assumptions and analyses that were
selected to ensure that the impacts from accident scenarios are not likely to be underestimated. DOE has
not developed the final design and operational details for the repository, and these details could result in
lower impacts. The Department is currently engaged in preliminary efforts to identify accidents and
evaluate their impacts as required to support the License Application for the repository that it will send to
the Nuclear Regulatory Commission, and to show that the repository would comply with appropriate
limits on radiation exposure to workers and the public from accidental releases of radionuclides. The
final design could include additional systems and operational requirements to reduce the probability of
accidents and to mitigate the release of radionuclides to ensure compliance with these safety
requirements. The results from the accident analysis to meet licensing requirements would be more
specific and comprehensive than those discussed in this appendix and would reflect final repository
design and operational details.
H.1 General Methodology
Because of the large amount of radioactive material to be handled at the proposed repository (see
Appendix A), the focus of the analysis was on accident scenarios that could cause the release of
radioactive material to the environment. The methodology employed to estimate the impact of accidents
involving radioactive material included (I) evaluation of previous accident analyses performed for the
repository, (2) identification of bounding accidents (reasonably foreseeable accidents with the maximum
consequences) from the previous analyses, (3) identification of other credible accidents the previous
analyses did not evaluate, (4) analyses of the selected accidents to determine the amount of radioactive
material an accident could release to the environment, and (5) estimation of the consequences of the
release of radioactive material in terms of health effects to workers and the public.
The analysis approach involved identifying bounding accidents (that is, accidents with maximum
consequences) for each operational phase of the proposed repository. The analysis evaluated the impacts
for these accidents, assuming the accident occurred without regard to the estimated probability. Thus, the
analysis provides the impacts that could occur for the worst credible accidents. The results do not
represent risk estimates because the impacts do not include a consideration of accident probability, which
in most cases is very low. The risk from all repository accidents would be likely to be far less than the
low risk, which DOE estimated by assuming that all of the bounding (maximum consequence) accidents
would occur.
Accident frequency estimates were derived to establish the credibility of accident sequences and were not
used to establish risk. Estimates of accident frequency are very uncertain due to the preliminary nature of
the currently available repository design information and would be more fully evaluated in the safety
H-1
Potential Repository Accident Scenarios: Analytical Methods and Results
analysis required to support a License Application for the repository. Based on the available design
information, the accident analysis approach was used to ensure that impacts from accidents are not likely
to be underestimated (whether they are low-probability with high-consequence accidents or high-
probability with low-consequence accidents).
For accidents not involving radioactive materials, the analysis determined that application of accident
statistics from other DOE operations provided a reasonable estimate of nonradiological accident impacts
(see Section H.2.2).
H.2 Potential Repository Accident Scenarios
The proposed Yucca Mountain Repository has been the subject of intense evaluations for a number of
years. Some of these evaluations included in-depth considerations of preclosure accidents that could
occur during repository operations. The EIS used these previous evaluations, to the extent they are
applicable and valid, to aid in the identification of initiating events, develop sequences, and estimate
consequences. The EIS groups accidents as radiological accidents (Section H.2.1) that involve the
unplanned release of radioactive material, and nonradiological accidents that involve toxic and hazardous
materials (Section H.2.2).
H.2.1 RADIOLOGICAL ACCIDENT SCENARIOS
Previous analyses that considered impacts of radiological accidents during preclosure included
evaluations by Sandia National Laboratories and others (Jackson et al. 1984, all; SNL 1987, all; Ma et al.
1992, all; BMI 1984, all), and include more recent evaluations (DOE 1996a,b, all; DOE 1997a,b all;
Kappes 1998, all; TRW 1997a, all). These evaluations were reviewed to assist in this assessment of
radiological impacts from accidents during repository operations. In addition, EISs that included accident
evaluations involving spent nuclear fuel and high-level radioactive waste were reviewed and used as
applicable (USN 1996, all; DOE 1995, all).
Radiological accidents involve an initiating event that can lead to a release of radioactive material to the
environment. The analysis considered accidents separately for two types of initiating events: (1) internal
initiating events that would originate in the repository and involve equipment failures or human errors, or
a combination of both, and (2) external initiating events that would originate outside the facility and affect
the ability of the facility to maintain confinement of radioactive or hazardous material. The analysis
examined a spectrum of accidents, from high-probability/low-consequence accidents to low-probability/
higher-consequence accidents.
H.2.1. 1 Internal Events - Waste Handling Building
The most recent and comprehensive repository accident scenario analysis for internal events in the Waste
Handling Building is presented in Kappes (1998, all). This analysis considered the other important
applicable accidents that previous analyses identified. It performed an in-depth evaluation of all
operations planned for the repository and identified bounding accidents (those with the highest estimated
risk) for each operation. More than 150 accidents were selected for analysis in eight operational
categories. The accidents were identified based on multiple sources, including the Preliminary MGDS
Hazards Analysis (DOE 1996b, all), current facility design drawings, and discussions with design
personnel. These 150 accidents were reduced to 16 bounding accidents by retaining accidents that would
produce the highest doses for groups of similar events (Kappes 1998, page 35). DOE used event trees and
fault tree evaluation to estimate frequencies for the accidents. A review of these evaluafions indicated
that they were valid for use in the EIS with a few exceptions (noted below).
H-2
Potential Repository Accident Scenarios: Analytical Methods and Results
RISK
Risk is defined as the possibility of suffering
liarm. It considers both the frequency (or
probability) and consequences of an
accident. In the scientific community, risk is
usually defined and computed as the product
of the frequency of an accident and the
consequences that result. This is the
definition of risk used in this analysis.
Rather than develop a single, overall
expression of the risks associated with
proposed actions, DOE usually finds it more
informative in its EIS accident scenario
analyses to consider a spectrum of accidents
from low-probability, relatively high-
consequence accidents to high-probability,
low-consequence accidents. Nevertheless,
risk is a valuable concept to apply in
evaluating the spectrum of accident
scenarios to ensure that accidents that are
expected to dominate risk have been
adequately considered.
The evaluation used to identify internal accidents
did not evaluate criticality events quantitatively
(Kappes 1998, page 34). Continuing evaluations
are under way to assess the probability and
consequences of a criticality event. The risk from
criticality events, however, would be unlikely to
exceed the risk from the bounding events
considered below. This preliminary conclusion is
based on several factors:
• The probability of a criticality event would be
very low. This is based on the Nuclear
Regulatory Commission design requirement
(10 CFR Part 60) that specifies that two
independent low-probability events must occur
for criticality to be possible and that this
requirement will be part of the licensing basis
for the repository. On the basis of this
requirement, the event is unlikely to be credible
(Jackson et al. 1984, page 18). Further, a
criticality event would require the assembly of
fuel with sufficient fissionable material to
sustain a criticality. Since the commercial
spent-nuclear fuel to be handled at the
repository is spent (that is, it has been used to
produce power), the remaining fissionable material is limited. For the pressurized-water reactor fiiel,
the amount of fuel that contains sufficient fissionable material to achieve criticality is only a small
percent spent nuclear fuel (DOE 1998a, page C-46). This material would have to be assembled in
sufficient quantity to achieve criticality, and the moderator (water) would somehow have to be added
to the assembled material. A quantitative estimate of criticality frequency is planned as part of the
license application (Kappes 1998, page 34).
• The criticality event that could occur despite the preventive measures described above would be
unlikely to compromise the confinement function of the ventilation and filtration system of the Waste
Handling Building. These features would inhibit the release of particulate radionuclides. By contrast,
the seismic event scenario (discussed in Section H.2. 1.3) assumes failure of these mitigating features.
• Criticality could occur only if the material was moderated with water and had sufficient fissionable
material in a configuration that could allow criticality. The water surrounding the material would act
to inhibit the release of particulate material (DOE 1994, Volume 1, Appendix D, page F-85) and,
thus, would limit the source term.
• During the monitoring and closure phase of operations, water needs to enter a waste package that
contains fuel with sufficient fissionable material to go critical. Water would have to flood a drift and
leak into a defective waste package to cause a criticality. Such an event is considered not credible
due to the lack of sufficient water sources, detection and remediation of water in-leakage, and
high-quality leak proof waste packages.
Considering these factors, the criticality event does not appear to be a large potential contributor to risk.
H-3
Potential Repository Accident Scenarios: Analytical Methods and Results
Table H-1 lists the bounding accident scenarios identified in Kappes (1998, page 40). For each accident
scenario, the table lists (1) the location of the accident, (2) the material at risk, or the amount of
radioactive material involved in the accident, and (3) if the analysis assumed that filtration (high-
efficiency particulate air filters) would be available to mitigate radioactive material releases. Filtration
would be provided in most areas of the Waste Handling Building (TRW 1999b, page 41) and in the
subsurface emplacement facilities (TRW 1999a, page 4-61). The Frequency column in Table H-1 lists
the estimated annual frequency of the event (Kappes 1998, all). The last column indicates if the EIS
analysis retained, eliminated, or adjusted details of the accident scenario.
Table H-1. Bounding internal accident scenarios for the Waste Handling Building and emplacement
operations.
Location" Number
Accident
Material at risk'^ Filters Frequency Disposition
A
A
A
A
A
A
A
B
B
B
1
2
3
4
5
6
7
10
c
11
c
12
c
13
c
14
D
15
D
D
E
E
16
17
18
19
6.9-meter drop of shipping cask
6.9-meter drop of shipping cask
7.1-meter drop of shipping cask
7.1-meter drop of shipping cask
4.1 -meter drop of shipping cask
4.1-meter drop of shipping cask
4.1 -meter drop of shipping cask
8.6-meter drop of canister
6.3-meter drop of multicanister
overpack
6.3-meter drop of multicanister
overpack
5-meter drop of transfer basket
5-meter drop of transfer basket
7.6-meter drop of transfer basket
7.6-meter drop of transfer basket
6-meter vertical drop of disposal
container
6-meter vertical drop of disposal
container
2.5-meter horizontal drop of
disposal container
Rockfall on waste package
Transporter runaway and derailment
61 BWR assemblies
61 BWR assemblies
26 PWR assemblies
26 PWR assemblies
61 BWR assemblies
61 BWR assemblies
26 PWR assemblies
DOE high-level waste
N-Reactor fuel
N-Reactor fuel
8 PWR assemblies
8 PWR assemblies
16 BWR assemblies
16 BWR assemblies
21 PWR assemblies
21 PWR assemblies
21 PWR assemblies
44 BWR assemblies
21 PWR assemblies
No
4.5x10"
Retained
Yes
d
Eliminated
No
6.1x10"
Retained
Yes
—
Eliminated
No
1.4x10-'
Retained
Yes
—
Eliminated
No
1.9x10'
Retained
Yes
4.2x10'
Eliminated'
Yes
4.5x10''
Retained
No
2.2x10-^
Added^
Yes
1.1x10"^
Retained
No
2.8x10"'
Added^
Yes
7.4x10"'
Retained
No
1.9x10"'
Added^
Yes
1.8x10"'
Retained
No 8.6x10"' Added^
Yes 3.2x10"'' Eliminated^
No 4.2x10"* Eliminated"
Yes 1.2x10"' Retained
a. Lx)cation designators; A = Cask/Carrier Transport and Handling Area, B = Canister Transfer System, C = Assembly
Transfer System, D = Disposal Container Handling System, E = Waste Emplacement and Subsurface Facility.
b. To convert meters to feet, multiply by 3.2808.
c. BWR = boiling-water reactor; PWR = pressurized-water reactor.
d. Eliminated from evaluation because current design does not include a filter system for this area (Kappes 1998, page 40).
e. Eliminated on the basis that it would not be a risk contributor because the N-Reactor multicanister overpack drop (accident
scenario BIO) has an estimated frequency more than 10 times higher, and the N-Reactor fuel has a higher radionuclide
inventory (Appendix A).
{. These accident scenarios, involving loss of filtration, were added because they would exceed the level of credibility
recommended by DOE (frequency greater than 1x10"' per year) (DOE 1993, page 28). The corresponding U.S. Nuclear
Regulatory Commission limit (used in Kappes 1998, page 4) is 1 x 10"' per year. The Commission considers accidents with
frequencies less than 1 x 10 * per year to be beyond design basis events.
g. Eliminated because it would not contribute to risk in comparison to accident scenario 15 at location D„ a higher drop event
that would produce larger consequences with a higher frequency.
h. Eliminated on the basis of low frequency, below the credible level of I x 10"'.
i. Frequency adjusted to account for the filtration system in the current design.
The following paragraphs contain details of the postulated accident scenarios in each location.
H-4
Potential Repository Accident Scenarios: Analytical Methods and Results
H.2.1 .1 .1 Cask/Carrier Transport and Handling Area
These accidents (Table H-1, location A, accidents 1 through 7) would involve mishaps that could occur
during the process of handling the transportation casks at the repository. The transportation casks would
be designed to withstand impacts from collisions and drops, and this capability is augmented by impact
limiters, which would be required during transportation. After cask arrival at the repository, the limiters
would be removed to facilitate handling of the casks. The casks would then become more vulnerable to
damage from physical impact. The analysis assumed that damage to the casks would occur if they were
dropped from heights greater than the design basis of 2 meters (6.6 feet) (Kappes 1998, page 13) without
the impact limiters. The various heights of the drops in the "Accident" column in Table H-1 correspond
to the maximum height to which the casks could be lifted during the various operations the analysis
assumed crane failure would occur. The material-at-risk column lists the contents of the casks when the
accident occurred. The largest casks are designed to hold either 61 boiling-water reactor or
26 pressurized-water reactor fuel assemblies.
Accident scenarios from Kappes (1998) that assume a filtration system is available (accidents A2, A4,
and A6) were eliminated from consideration in the EIS because the current design concept of the
Cask/Carrier Transport and Handling Area does not include such a filtration system; they were considered
in Kappes (1998, page 40) for information only.
H.2.1 .1 .2 Canister Transfer System
The Canister Transfer System would handle canisters that arrived at the repository and were suitable for
direct transfer to the disposal container. The bounding accident scenarios for these operations would be
canister drops of DOE high-level radioactive waste and N-Reactor fuel (accidents 8 and 9 at location B in
Table H-1). The analysis eliminated the DOE high-level radioactive waste canister drop because it would
not be a risk contributor in comparison to the N-Reactor fuel drop. The N-Reactor multicanister overpack
drop would have a frequency more than 10 times greater than that for the high-level radioactive waste
canister drop, and the N-Reactor radionuclide inventory would be greater (see Appendix A). The EIS
analysis added an additional accident scenario, which would be a drop of the N-Reactor fuel canister with
loss of the filtration system. The analysis estimated the filtration system failure probabilities by using the
fault tree analysis technique, and the results differ somewhat from the failures identified in Section
H.2.1. 1.3 due to design variations dependant on location in the surface facilities of the repository. DOE
computed this accident scenario probability by combining the accident drop probability of 0.00045 with
the filter system failure of 4.8 x 10"'* from Kappes (1998, page 4) for an accident sequence frequency of
2.2 X 10'^ per year. [Kappes (1998, page 4) did not consider accident sequences with frequencies less
than 1 X 10"^.] This sequence frequency is based on failure of the heating, ventilating, and air
conditioning system such that it would not provide filtration for 24 hours following the accident,
consistent with Kappes (1998, page VIII- 1).
H.2.1 .1 .3 Assembly Transfer System
The Assembly Transfer System would handle bare, intact commercial spent nuclear fuel assemblies from
pressurized- and boiling-water reactors. The assemblies would be unloaded from the transportation cask
in the cask unloading pool. Next, they would be moved to the assembly staging pool where they would
be placed in baskets that contained either four pressurized-water reactor assemblies or eight boiling-water
assemblies. The baskets would be moved from the pool and transferred to the assembly drying station
from which they would be loaded, after drying, in the disposal containers. The bounding accident
scenarios found during a review of this operation (Kappes 1998, page 40) were drops of a suspended
basket loaded with fuel assemblies on another loaded basket in the drying vessel (accident scenarios 1 1
and 13 at location C from Table H-1). DOE added two accident scenarios to the EIS analysis that
H-5
Potential Repository Accident Scenarios: Analytical Methods and Results
included failure of the high-efficiency particulate air filtration system (accident scenarios 12 and 14 at
location C from Table H-1). DOE computed the frequency of these accidents by combining the accident
drop frequency with the filter failure probability of 0.000025, which corresponds to the failure probability
of the heating, ventilation, and air conditioning system in the assembly transfer area (Kappes 1998,
page 11). Thus, the frequency of a drop accident and subsequent failure of the heating, ventilation, and
air conditioning system during the 24 hours (the period assumed that the filtration system would need to
operate to remove the particulate material effectively) would be:
• For boiling-water reactor assembly drop: 0.01 1 x 0.000025 = 0.00000028
• For pressurized-water reactor assembly drop: 0.0074 x 0.000025 = 0.00000019
H.2.1 .1 .4 Disposal Container Handling System
The Disposal Container Handling System would prepare empty disposal containers for the loading of
nuclear materials, transfer disposal containers to and from the assembly and canister transfer systems,
weld the inner and outer lids of the disposal containers, and load disposal containers on the waste
emplacement transporter. After the disposal container had been loaded and sealed, it would become a
waste package. Disposal containers would be lifted and moved several times during the process of
preparing them for loading on the waste emplacement transporter. DOE examined the details of these
operations and identified numerous accident scenarios that could occur (Kappes 1998, Attachment V).
The bounding accident scenarios from this examination would be the disposal container drop accident
scenarios listed as accident scenarios 15 and 17 at Location D in Table H-1. However, the analysis
eliminated accident scenario 17 because it would be a minor contributor to risk in comparison to accident
scenario 15. Accident scenario 15, which would have a higher probability (by about a factor of 6), would
produce a higher radionuclide release due to the increased drop height (by a factor of more than 2). Thus,
the overall risk contribution from accident scenario 17 would be less than 10 percent of the risk from
accident scenario 15. For the EIS, DOE added another accident scenario (16) to account for the
possibility of loss of filtration. The analysis assumed that the heating, ventilation, and air conditioning
filtration system would fail with a probability of 0.00048 (Kappes 1998, page 4).
H.2.1 .1 .5 Waste Emplacement and Subsurface Facility Systems
The waste emplacement system would transport the loaded and sealed waste package from the Waste
Handling Building to the subsurface emplacement area. This system would operate on the surface
between the North Portal and the Waste Handling Building, and in the underground ramps, main drifts
(tunnels), and emplacement drifts. It would use a reusable railcar for waste package transportation. The
railcar would be moved into the waste emplacement area by an electric locomotive, and the waste
package would be placed in the emplacement drift. The bounding accident scenarios identified (Kappes
1998, page 40) for this operation would be accident scenarios 18 and 19 at location E, as listed in Table
H-1. However, DOE eliminated accident scenario 18 (rockfall on waste package) because the estimated
frequency of a radioactive release from such an event is not credible (estimated frequency of 4.2 x 10"
per year) (Kappes 1998, page VI-5).
An accident scenario involving a failure of the ventilation system in conjunction with a transporter
runaway and collision (accident scenario F19 from Table H-1) would not be credible, so the sequence was
not analyzed. The original transporter runaway and derailment accident scenario assumed the
involvement of 44 boiling-water reactor assemblies (Kappes 1998, page 40). The EIS analysis assumed
the involvement of 21 pressurized-water reactor assemblies because (1) they would represent a slightly
higher impact potential due to the greater radionuclide inventory than that in the smaller 44 boiling-water
reactor assemblies and would, therefore, bound the equivalent accident involving such assemblies, and
H-6
Potential Repository Accident Scenarios: Analytical Methods and Results
(2) an accident scenario involving pressurized-water reactor fuel would be more likely because DOE
expects to emplace about twice as much of this type of fuel in the proposed repository (Appendix A).
Section H.2. 1.4 describes the source terms (amount and type of radionuclide release) for these accident
scenarios, and Section H.2.1.5 assesses the estimated consequences from the accident scenarios.
H.2.1 .2 internal Events - Waste Treatment Building
An additional source of radionuclides could be involved in accidents in the Waste Treatment Building.
This building, which would be connected to the northeast end of the Waste Handling Building, would
house the Site-Generated Radiological Waste Handling System (TRW 1999b, page 37). This system
would collect site-generated low-level radioactive solid and liquid wastes and prepare them for disposal.
The radioactivity of the waste streams would be low enough that no special features would be required to
meet Nuclear Regulatory Commission radiation safety requirements (shielding and criticality)
(TRW 1999b, page 38).
The liquid waste stream to the Waste Treatment Building would consist of aqueous solutions that could
contain radionuclides resulting from decontamination and washdown activities in the Waste Handling
Building. The liquid waste would be evaporated, mixed with cement (grouted), and placed in 0.21 -cubic-
meter (55-gallon) drums for shipment off the site (TRW 1999b, page 53). The evaporation process would
reduce the volume of the liquid waste stream by 90 percent (DOE 1997c, Summary).
The solid waste would consist of noncompactible and compactible materials and spent ion-exchange
resins. These materials ultimately would be encapsulated in concrete in 0.21-cubic meter (55-gallon)
drums after appropriate processing (TRW 1999b, page 55).
Water in the Assembly Staging Pools of the Waste Handling Building would pass through ion exchange
columns to remove radionuclides and other contaminants. These columns would accumulate
radionuclides on the resin in the columns. When the resin is spent (unable to remove radionuclides
effectively from the water), the water flow would be diverted to another set of columns, and the spent
resin would be removed and dewatered for disposal as low-level waste or low-level mixed waste. These
columns could have external radiation dose rates associated with them because of the activation and
fission product radionuclides accumulated on the resins. They would be handled remotely or
semiremotely. During the removal of the resin and preparation for offsite shipment in the Waste
Treatment Building, an accident scenario involving a resin spill could occur. However, because the
radionuclides would have been chemically bound to the resin in the column, an airborne radionuclide
release would be unlikely. Containment and filter systems in the Waste Treatment Building would
prevent exposure to the public or noninvolved workers. Some slight exposure of involved workers could
occur during the event or during recovery operations afterward. DOE made no further analysis of this
event.
Because there is no detailed design of the Waste Treatment Building at present and operational details are
not yet available, DOE used the recent Waste Management Programmatic EIS (DOE 1997c, all) and
supporting documentation (Mueller et al. 1996, all) to aid in identifying potential accident scenarios and
evaluating radionuclide source terms. For radiological impacts, the analysis focused on accident
scenarios with the potential for airborne releases to the atmosphere. The liquid stream can be eliminated
because it has a very low potential for airborne release; the radionuclides would be dissolved and energy
sources would not be available to disperse large amounts of the liquid into droplets small enough to
remain airborne. Many low-level waste treatment operations, including evaporation, solidifying
(grouting), packaging, and compaction can be excluded because they would lack sufficient mechanistic
stresses and energies to create large airborne releases, and nuclear criticalities would not be credible for
H-7
Potential Repository Accident Scenarios: Analytical Methods and Results
low-level waste (Mueller et al. 1996, page 13). Drum-handling accidents are expected to dominate the
risk of exposure to workers (Mueller et al. 1996, page 93).
The estimated frequency of an accident involving drum failure is about 0.0001 failure per drum operation
(Mueller et al. 1996, page 39). The total number of drums containing grouted aqueous waste would be
2,280 per year (DOE 1997c, page 30). The analysis assumed that each drum would be handled twice,
once from the Waste Treatment Building to the loading area, and once to load the drum for offsite
transportation. Therefore, the frequency of a drum failure involving grouted aqueous waste would be:
Frequency = 2,280 aqueous (grouted) low-level waste drums per year
X 2 handling operations per drum
X 0.0001 failure per handling operation
= 0.46 aqueous (grouted) low-level waste drum failures per year.
The number of solid-waste grouted drums produced would be 2,930 per year (DOE 1997c, page 35).
Assuming two handling operations and the same failure rate yields a frequency of drum failure of.
Frequency = 2,930 solid low-level waste drums per year
X 2 handling operations per drum
X 0.0(X)1 failure per handling operation
= 0.59 solid low-level waste drum failures per year.
Failure of these drums would result in a release of radioactive material, which later sections evaluate
further.
H.2.1.3 External Events
External events are either external to the repository (earthquakes, high winds, etc.) or are natural
processes that occur over a long period of time (corrosion, erosion, etc.). DOE performed an evaluation
to identify which of these events could initiate accidents at the repository with potential for release of
radioactive material.
Because some external events evaluated as potential accident-initiating events would affect both the
Waste Treatment and Waste Handling Buildings simultaneously [the buildings are physically connected
(TRW 1999b, page 38)], this section considers potential accidents involving external event initiators, as
appropriate, for the combined buildings.
Table H-2 lists generic external events developed as potential accident initiators for consideration at the
proposed repository and indicates how each potential event could relate to repository operations based oil
an initial evaluation process. The list, from DOE (1996b, page 15), was developed by an extensive
review of relevant sources and known or predicted geologic, seismologic, hydrologic, and other
characteristics. The list includes external events from natural phenomena as well as man-caused events, j
The center column in Table H-2 (relation to repository) represents the results of a preliminary evaluation]
to determine the applicability of the event to the repository operations, and is based in part on evaluatior
previously reported in DOE (1996b, all). Events were excluded for the following reasons:
• Not applicable because of site location (condition does not exist at the site)
• Not applicable because of site characteristics (potential initiator does not exist in the vicinity of the
site)
H-8
I
Potential Repository Accident Scenarios: Analytical Methods and Results
Table H-2. External events evaluated as potential accident initiators/
Event
Relation to repository
Comment
Aircraft crash
Avalanche
Coastal erosion
Dam failure
Debris avalanche
Dissolution
Epeirogenic displacement (tilting of
the Earth's crust)
Erosion
Extreme wind
Extreme weather
Fire (range)
Flooding
Denudation
Fungus, bacteria, algae
Glacial erosion
High lake level
High tide
High river stage
Hurricane
Inadvertent future intrusion
Industrial activity
Intentional future intrusion
Lightning
lx)ss of offsite or onsite power
Low lake level
Meteorite impact
Military activity
Orogenic diastrophism
Pipeline rupture
Rainstorm
Sandstorm
Sedimentation
Seiche
Seismic activity, uplift
Seismic activity, earthquake
Seismic activity, surface fault
Seismic activity, subsurface fault
Static fracture
Stream erosion
Subsidence
Tornado
Tsunami
Undetected past intrusions
Undetected geologic features
Undetected geologic processes
Volcanic eruption
Volcanism, magmatic
Volcanism, ash flow
Volcanism, ash fall
Waves (aquatic)
A
C
B
C
A
A
D (earthquake)
D (flooding)
A
A
A
A
E
E
B
C
B
C
B
E
A
E
A
A
C
A
A
D (earthquake)
C
D (flooding)
A
B
B
D (earthquake)
A
D (earthquake)
D (earthquake)
D (earthquake)
B
D (earthquake)
A
B
E
D (earthquake, volcanism
ash fall)
D (erosion, earthquake,
volcanism ash fall)
D (volcanism ash fall)
D (volcanism ash fall)
D (volcanism ash fall)
A
B
Caused by excessive rainfall
Chemical weathering of rock
Large-scale surface uplifting and subsidence
Includes extreme episodes of fog, frost, hail, ice cover, etc.
Wearing away of ground surface by weathering
A potential waste package long-term corrosion process not
relevant during the repository operational period'
To be addressed in postclosure Performance Assessment
Movement of Earth's crust by tectonic processes
Surface water waves in lakes, bays, or harbors
Rock breakup caused by stress
Sinking of Earth's surface
Sea wave caused by ocean floor disturbance
a. Source: DOE (1996b, page 15).
b. A = retained for further evaluation; B = not applicable because of site location; C = not applicable because of site
characteristics (threat of event does not exist in the vicinity of the site); D = included in another event as noted; E = does not
represent an accident-initiating event for proposed repository operations.
c. Source: TRW (1999a, all).
H-9
Potential Repository Accident Scenarios: Analytical Methods and Results
• Included in another event
• Does not represent an accident-initiating event for proposed repository operations
The second column of Table H-2 identifies the events excluded for these reasons. The preliminary
evaluation retained the events identified in Table H-2 with "A" for further detailed evaluation. The
results of this evaluation are as follows:
1 . Aircraft Crash. The EIS analysis evaluated the frequency of aircraft crashes on the proposed
repository to determine if such events could be credible and, therefore, candidates for consequence
analysis. This frequency determination used analytical methods recommended for aircraft crashes
into hazardous facilities (DOE 1996c, all).
An earlier analysis assumed that the only reasonable aircraft crash threat would be from military
aircraft operations originating from Nellis Air Force Base (Kimura, Sanzo, and Sharirli 1998, page 8),
primarily because commercial and general aviation aircraft are restricted from flying over the Nevada
Test Site. DOE considered this assumption valid and adopted it for the EIS analysis.
The formula used in the crash frequency analysis, taken from Kimura, Sanzo, and Sharirli (1998,
pages 9 to 12) based on DOE (1996c, all), was:
F = (N, ^ A,) X Aeff X A. X (4 ^ 71) X (Reff + Re)
where:
F = the frequency per year of aircraft crashes on the repository
N, = total number of aircraft overflights per year
A, = total area of the overflight region
Aeff = effective area of the repository (target area)
X = crash rate of the aircraft per mile of flight
Reff = effective radius of the repository (target area)
Re = radius of the crash area potentially affected by a distressed aircraft
The parameters in this formula were quantified as follows:
Nt The estimated total number of flights in the flight corridor in the vicinity of the repository would
be 13,000 per year, with the repository located on the western edge of the corridor, which extends
49 kilometers (30 miles) to the east. Most flights would not be observed from the repository.
However, this value was used in a recent crash assessment for a Nevada Test Site facility beneath
the same airspace as the repository (Kimura, Sanzo, and Sharirli 1998, page 7). Future Nellis
operations could result in increased overflights. The only known planned change in future
activities involve the bed-down of F-22 fighter aircraft. This planned activity involves 17 aircraft
that will be at Nellis by 2010. The additional aircraft would increase flight activities by only 2 to
3 percent over current activities (Myers 1997, page 2).
A, The total area of the overflight area would be about 3,400 square kilometers (1,300 square miles)
(Kimura, Sanzo, and Sharirli 1998, page 18).
H-10
Potential Repository Accident Scenarios: Analytical Methods and Results
Aeff The analysis estimated the repository target area by assuming that the roof of the Waste Handling
Building would be the only vulnerable location at the repository with the potential for a large
radionuclide release as a result of an aircraft impact. This is because the Waste Handling
Building would be the only facility that would handle bare spent nuclear fuel assemblies. The
shipping casks and the waste packages loaded with spent nuclear fuel or high-level radioactive
waste would not be vulnerable to air crash impacts because both would have steel walls thick
enough to prevent aircraft penetration. The Waste Treatment Building would not contain large
amounts of radioactive material, so radionuclide releases from accidents involving this building
would not produce large impacts (see Section H.2.1.4 for details). Further, the walls of the Waste
Handling Building around areas for the handling of canisters and fuel assemblies would be
1.5 meters (5 feet) thick to a level of 9 meters (30 feet), and then 1 meter (3.3 feet) thick to the
intersection with the roof (TRW 1999b, pages 31 to 37). The aircraft crash would not penetrate
these walls because the concrete penetration capability for aircraft is limited to about 0.76 meter
(2.5 feet) (see Appendix K for details). Therefore, the only likely vulnerable target area at the
repository would be the roof of the Waste Handling Building, which would consist of concrete 20
to 25 centimeters (8 to 10 inches thick) (TRW 1999b, pages 31 to 37). The overall footprint of
the Waste Handling Building would be about 163 meters by 165 meters (535 feet by 540 feet),
which would produce a target area of approximately 27,0(X3 square meters (290,000 square feet).
X The crash rate for the small military aircraft involved in the overflights [primarily F-15s, F-16s,
and A-lOs (USAF 1999, pages 1-34 to 1-35)] would be 1.14 x 10"^ per kilometer (1.84 x 10"* per
mile) (Kimura, Sanzo, and Sharirli 1998, page 7). Large military aircraft fly over the area to
some extent, but have a lower crash rate [1.17 x 10' per kilometer (1.9 x 10"' per mile) (Kimura,
Sanzo, and Sharirli 1998, page 7)]. Thus, the use of the small aircraft crash rate bounds the large
aircraft crash rate.
Reff The effective radius of the repository is the equivalent radius of the repository target effective
area (Aeff), or R^ff is equal to the square root of the quotient 27,000 square meters divided by pi,
which is about 93 meters (310 feet).
Re The radius of the crash area potentially affected by a distressed military aircraft represents the
distance an aircraft could travel after engine failure (glide distance). This distance is the glide
ratio of the aircraft times the elevation of the flight above the ground. The aircraft are required to
fly a minimum of 4,300 meters (14,(X)0 feet) above mean sea level while in the airspace over the
repository (Kimura, Sanzo, and Sharirli 1998, page 5). The actual altitude flown varies from
4,600 to 7,000 meters (15,000 to 23,000 feet) (Tullman 1997, page 4). For this analysis, a mean
altitude of 5,800 meters (19,0{X) feet) was assumed. Because the Waste Handling Building would
be at about 1,100 meters (3,680 feet) (TRW 1998a, page 1-6), the mean flight elevation for
aircraft above the repository ground level would be about 4,700 meters (10,000 feet). The glide
ratio for the aircraft involved in the overflights (F-15, F-16, and A-10) is 8 (Thompson 1998, all).
Therefore, Re would be 4,700 meters multiplied by 8, which is 38,000 meters or 38 kilometers
(23 miles).
Substituting these values into the frequency equation yields:
F = (13,000 ^ 3,400) X 0.027 X 1.14x10'* X (4^71) X (38 -I- 0.093)
= 5.6 X 10"* crash per year.
Thus, aircraft crashes on the vulnerable area of the repository are not credible because the probability
would be below 1 x 10"^ per year, which is the credible limit specified by DOE (1993, page 28).
H-11
Potential Repository Accident Scenarios: Analytical Methods and Results
2. Debris Avalanche. This event, which can result from persistent rainfall, would involve the sudden
and rapid movement of soil and rock down a steep slope. The nearest avalanche potential to the
proposed location for the Waste Handling Building is Exile Hill (the location of the North Portal
entrance). The base of Exile Hill is about 90 meters (300 feet) from the location of the Waste
Handling Building. Since Exile Hill is only about 30 meters (100 feet) high (TRW 1997a, page 5.09),
it would be unlikely that avalanche debris would reach the Waste Handling Building. Furthermore,
the design for the Waste Handling Building includes concrete walls about 1.5 meters (5 feet) thick
(TRW 1999b, page 38) that would provide considerable resistance to an impact or buildup of
avalanche debris.
3. Dissolution. Chemical weathering could cause mineral and rock material to pass into solution.
This process, called dissolution, has been identified as potentially applicable to Yucca Mountain
(DOE 1996b, page 18). However, this is a very slow process, which would not represent an accident-
initiating event during the preclosure period being considered in this appendix.
4. Extreme Wind. Extreme wind conditions could cause transporter derailment (TRW 1997b,
page 72), the consequences of which would be bounded by a transporter runaway accident scenario.
The runaway transporter accident scenario is discussed further in Section H.2.1.4.
5. Extreme Weather. This potential initiating event includes various weather-related phenomena
including fog, frost, hail, drought, extreme temperatures, rapid thaws, ice cover, snow, etc. None of
these events would have the potential to cause damage to the Waste Handling Building that would
exceed the projected damage from the earthquake event discussed in this section. In addition, none of
these events would compromise the integrity of waste packages exposed on the surface during
transport operations. Thus, the earthquake event and other waste package damage accident scenarios
considered in this appendix would bound all extreme weather events. It would also be expected that
operations would be curtailed if extreme weather conditions were predicted.
6. Fire. There would be two potential external fire sources at the repository site — diesel fuel oil storage
tank fires and range fires. Diesel fuel oil storage tanks would be some distance [more than 90 meters
(300 feet)] from the Waste Handling Building and Waste Treatment Building (TRW 1999b,
Attachment IV Figure 4). Therefore, a fire at those locations would be highly unlikely to result in any
meaningful radiological consequences. Range fires could occur in the vicinity of the site, but would
be unlikely to be important accident contributors due to the clearing of land around the repository
facilities. Furthermore, the potential for early fire detection and, if necessary, active fire protection
measures and curtailment of operations (TRW 1999b, Section 4.2) would minimize the potential for
fire-initiated radiological accidents. DOE is performing detailed evaluations of fire-initiating events
(Kappes 1998, page III-2), and will incorporate the results in the Final EIS as appropriate.
7. Flooding. Flash floods could occur in the vicinity of the repository (DOE 1996b, page 21).
However, an earlier assessment (Kappes 1998, page 32) screened out severe weather events as
potential accident-initiating events primarily by assuming that operational rules will preclude
transport and emplacement operations whenever there are local forecasts of severe weather. A
quantitative analysis of flood events (Jackson et al. 1984, page 34) concluded that the only radioactive
material that extreme flooding would disperse to the environment would be decontamination sludge
from the waste treatment complex. The doses resulting from such dispersion would be limited to
workers, and would be very small (Jackson et al. 1984, page 53). A more recent study reached a
similar conclusion (Ma et al. 1992, page 3-11).
8. Industrial Activity. This activity would involve both drift (tunnel) development activities at the
repository and offsite activities that could impose hazards on the repository.
H-12
Potential Repository Accident Scenarios: Analytical Methods and Results
a. Emplacement Drift Development Activities - Drift development would continue during waste
package emplacement activities. However, physical barriers in the main drifts would isolate
development activities fi^om emplacement activities (TRW 1999a, page 4-52). Thus, events that
could occur during drift development activities would be unlikely to affect the integrity of waste
packages.
b. External Industrial Activities - The analysis examined anticipated activities in the vicinity of the
proposed repository to determine if accident-initiating events could occur. Two such activities—
the Kistler Aerospace activities and the Wahmonie rocket launch facility — could initiate
accidents at the repository from rocket impacts. The Wahmonie activities, which involved rocket
launches from a location several miles east of the repository site, have ended (Wade 1998, all), so
this facility no longer poses a risk to the repository. The planned Kistler Aerospace activities
would involve launching rockets from the Nevada Test Site to place satellites in orbit (DOE
1996d, Volume 1, page A-42). However, at present there is insufficient information to assess if
this activity would pose a threat to the repository. As details become available, the Final EIS will
evaluate the potential for this activity to become an external accident-initiating event. (Aircraft
activity was discussed in item 1 above.)
9. Lightning. This event has been identified as a potential design-basis event (DOE 1997b, pages 86
and 87). Therefore, the analysis assumed that the designs of appropriate repository structures and
transport vehicles would include protection against lightning strikes. The lightning strike of principal
concern would be the strike of a transporter train during operations between the Waste Handling
Building and the North Portal (DOE 1997b, page 86). The estinnated frequency of such an event
would be 1.9 X 10"^ per year (Kappes 1998, page 33). DOE expects to provide lightning protection
for the transporter (TRW 1998b, Volume 1, page 18) such that a lightning strike that resulted in
enough damage to cause a release would be well below the credibility level of 1 x 10 per year (DOE
1993, page 28).
10. Loss of Off site Power. A preliminary evaluation (DOE 1997b, page 84) concluded that a
radionuclide release from an accident sequence initiated by a loss of offsite power would be unlikely.
Loss of offsite power events could result in a failure of the ventilation system and of the overhead
crane system. However, there would be emergency power for safety systems at the site (TRW 1999b,
page 45). Loss of offsite power was included as a contributor to the frequency of crane failure
(Kappes 1998, page III-6), as listed in the event frequencies in Table H-1.
1 1 . l\/leteorite Impact. This event would not be credible based on a strike frequency of 2 x 10 per
year for a damaging meteorite [based on data in Solomon, Erdmann, and Okrent (1975, page 68)].
This estimate accounts for the actual area of the Waste Handling Building roof given previously in
item 1.
12. Military Activity. Two different military activities would have the potential to affect repository
operations. One is the possibility of an aircraft crash from overflights from Nellis Air Force Base. The
analysis determined that this event would not be credible, as described above in this section. The
second potential activity is the resumption of underground nuclear weapons testing, which the United
States has suspended. The only impact such testing could impose on the repository would be ground
motion associated with the energy released from the detonation of the weapon. The impact of such
motion was the subject of a recent study that concluded that ground motions at Yucca Mountain from
nuclear tests would not control seismic design criteria for the potential repository (Walck 1996,
page i).
H-13
Potential Repository Accident Scenarios: Analytical Methods and Results
13. Sandstorm. Severe sandstorms could cause transporter derailments and sand buildup on structures.
However, such events would be unlikely to initiate accidents with the potential for radiological
release. Ma et al. (1992, page 3-11) reached a similar conclusion. Furthermore, it is assumed that
DOE probably would curtail operations if local forecasts indicated the expected onset of high winds
with potential to generate sandstorms (Kappes 1998, page 32). For these reasons, the analysis
eliminated this event from further consideration.
14. Seismic Activity, Earthquake (including subsidence, surface faults, uplift, subsurface fault,
and static fracture). DOE has selected the beyond-design-basis earthquake for detailed analysis.
The seismic design basis for the repository specifies that structures (including the Waste Handling
Building), systems, and components important to safety should be able to withstand the horizontal
motion from an earthquake with a return frequency of once in 10,000 years (annual probability of
occurrence of 0.0001) (Kappes 1998, page VII-3). A recent comprehensive evaluation of the seismic
hazards associated with the site of the proposed repository (USGS 1998, all) concluded that a 0.0001-
per-year earthquake would produce peak horizontal accelerations at the site of about 0.53g (mean
value). Structures, systems, and components are typically designed with large margins over the
seismic design basis to account for uncertainties in material properties, energy absorption, damping,
and other factors. For nuclear powerplant structures, the methods for seismic design provide a factor
of safety of 2.5 to 6 (Kennedy and Ravindra 1984, page R-53). In the absence of detailed design
information, the analysis conservatively assumed that the Waste Handling Building would collapse at
an acceleration level twice that associated with the design-basis earthquake, or 1.1^. Figure H-1
shows that this acceleration level would be likely to occur with a frequency of about 2 x 10^ per year
(mean value).
The Waste Treatment Building is not considered a safety-related structure. Accordingly, the seismic
design basis for this building is to withstand an earthquake event with a return frequency of 1,000
years (annual exceedance probability of 1 x 10"^ per year) (TRW 1999b, page 14). Consistent with
the assumption for the Waste Handling Building, it is assumed that the Waste Treatment Building
would collapse during an earthquake that produced twice the design level acceleration. From Figure
H-1, the design-basis acceleration for a 1 x 10"^ per year event is 0.18^. Thus, the building collapse is
assumed to occur at an acceleration level of 0.36, which has an estimated return frequency of about
2 X 10'" per year. The analysis retains these events as accident initiators, and evaluates the
consequences in subsequent sections. The effects of other seismic-related phenomena included under
this event (subsidence, surface faults, uplift, etc.) would be unlikely to produce greater consequences
than those associated with the acceleration produced by the seismic event selected for analysis
(complete collapse of the Waste Handling and Waste Treatment Buildings).
15. Tornado. The probability of a tornado striking the repository is estimated to be 3 x 10"' (one chance
in 10 million) based on an assessment of tornado strike probability for any point on the Nevada Test
Site (DOE 1996d, page 4-146), which is adjacent to the proposed repository. This is slightly above
the credibility level of 1 x 10'^ for accidents, as defined by DOE (DOE 1993, page 28). However,
most tornadoes in the western United States have relatively modest wind speeds. For example, the
probability of a tornado with wind speeds greater than 100 miles per hour is 0.1 or less (Ramsdell and
Andrews 1986, page 41). Thus, winds strong enough to damage the Waste Handling Building are
considered to be not credible.
Tornadoes can generate missiles that could penetrate structures at the repository, but radioactive
material would be protected either by shipping casks, the Waste Handling Building with thick
concrete walls, or the waste package. Therefore, tornado-driven missiles would not be a great hazard.
H-14
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Potential Repository Accident Scenarios: Analytical Methods and Results
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0.001
0.0001
0.00001
0.000001
85th-percentile
Mean
Median
15th-percentile
0.0000001
_L
J_
J_
_L
_L
_L
0.2 0.4 0.6 0.8 1 1.2 1.4 1.6
Peak horizontal acceleration {g)
1.8
Source: USGS (1998. Figure 7-A).
Figure H-1. Integrated seismic hazard results: summary hazard curves for peak horizontal acceleration.
H-15
Potential Repository Accident Scenarios: Analytical Methods and Results
16. Volcanism, Ash Fall. The potential for volcanic activity at the proposed repository site has been
studied extensively. A recent assessment (Geomatrix and TRW 1996, page 4-46) estimates that the
mean annual frequency of a volcano event that would intersect the repository footprint would be
1.5 X 10"^ per year (with 5 percent and 95 percent bounds of 5 x 10'° and 5 x 10'^ per year), which is
below the frequency of a credible event (DOE 1993, page 28). This result is consistent with a
previous study of volcano activity at the site (DOE 1998b, all). Impacts from a regional volcanic
eruption would be more likely; such an event could produce ash fall on the repository, and would be
similar to the sandstorm event discussed above. Ash fall could produce a very heavy loading on the
roof of the Waste Handling Building. Studies have concluded, however, that the worst case event
would be an ash fall of 3 centimeters (1.2 inches) and analyses to date indicate that repository
structures would not be affected by a 3-centimeter ash fall (DOE 1998b, Volume 1, pages 2-9).
17. Sabotage. The analysis separately considered sabotage (not listed in Table H-2) as a potential
initiating event. This event would be unlikely to contribute to impacts from the repository. The
repository would not represent an attractive target to potential saboteurs due to its remote location and
the low population density in the area. Furthermore, security measures DOE would use to protect the
waste material from intrusion and sabotage (TRW 1999b, pages 58 to 60) would make such attempts
unlikely to succeed. At all times the waste material would be either in robust shipping or disposal
containers or inside the Waste Handling Building, which would have thick concrete walls. On the
basis of these considerations, DOE concluded that sabotage events would be unlikely at the
repository. DOE expects that both the likelihood and consequences of sabotage events would be
greater during transportation of the material to the repository (DOE 1997d, page 14). Appendix J
presents the impacts of sabotage events during transportation.
Based on the external event assessment, DOE concluded that the only external event with a credible
potential to release radionuclides of concern would be a large seismic event. This conclusion is supported
by previous studies that screened out all external event accident initiators except seismic events (Ma et al.
1992, page 3-11; Jackson et al. 1984, pages 12 and 13). DOE is continuing to evaluate a few external
events (Kappes 1998, page 33), and will examine the results of these evaluations to confirm the Draft EIS
conclusions. If revisions are necessary, they will be provided in the Final EIS.
H.2.1.4 Source Terms for Repository Accident Scenarios
Following the definition of the accident scenarios as provided in previous sections, the analysis then
estimated a source term for each accident scenario retained for analysis. The source term specification
needed to include several factors, including the quantity of radionuclides released, the elevation of the
release, the chemical and physical forms of the released radionuclides, and the energy (if any) of the
plume that would carry the radionuclides to the environment. These factors would be influenced by the
state of the material involved in the accident and the extent and type of damage estimated for the accident
sequence. The estimate of the source term also considered mitigation measures, either active (for
example, filtration systems) or passive (for example, local deposition of radionuclides or containment),
that would reduce the amount of radioactive material released to the environment.
The analysis developed the source term for each accident scenario retained for evaluation. These include
the accident scenarios retained from the internal events as listed in Table H-1 and the seismic event
retained from the external event evaluation. Because many of the internal event-initiated accidents would
involve drops of commercial spent nuclear fuel, the analysis considered the source term for these
accidents as a group. Accordingly, source terms were developed for the following accident scenarios:
commercial spent nuclear fuel drops, transporter runaway and derailment, DOE spent nuclear fuel drop,
seismic event, and low-level waste drum failure.
H-16
Potential Repository Accident Scenarios: Analytical Methods and Results
H.2.1 .4.1 Commercial Spent Nuclear Fuel Drop Accident Scenario Source Term
Commercial sp)ent nuclear fuel contains more than 100 radioactive isotopes (SNL 1987, Appendix A).
Not all of these isotopes, however, would be important in terms of a potential to cause adverse health
effects (radiotoxicity) if released, and many would have decayed by the time the material arrived at the
repository. Based on the characteristics of the radioactivity associated with an isotope (including type and
energy of radioactive emissions, amount produced during the fissioning process, half-life, physical and
chemical form, and biological impact if inhaled or ingested by a human), particular isotopes could be
meaningful contributors to health effects if released. To determine the important radionuclides for an
accident scenario consequence analysis, DOE consulted several sources. The Nuclear Regulatory
Commission has identified a minimum of eight radionuclides in commercial spent nuclear fuel that "must
be analyzed for potential accident release" (NRC 1997, page 7-6). Repository accident scenario
evaluations (SNL 1987, pages 5-3 and 5-4) identified 14 isotopes (five of which were also on the Nuclear
Regulatory Commission list) that contribute to "99 percent of the total dose consequence." A more recent
analysis (DOE 1996a, pages 6 to 9) lists 24 radionuclides (10 of which were not included in either of the
other two lists) that are important for consequence analysis (99.9-percent cumulative dose for at least one
organ). The DOE analysis also included carbon-14. Appendix A contains a list of 53 radionuclides,
which includes the important isotopes discussed above. DOE used this longer list in the development of
the source term for the accident scenario analyses.
Commercial spent nuclear fuel includes two primary types — boiling-water reactor and pressurized-water
reactor spent fuel. For these commercial fuels, the radionuclide inventory depends on bumup (power
history of the fuel) and cooling time (time since removal from the reactor). The EIS accident scenario
analysis used "typical" fuels for each type. These typical fuels are representative of the majority of the
fuel DOE would receive at the repository (see Appendix A). Table H-3 lists the characteristics of typical
commercial spent nuclear fuel types.
Table H-3. Typical commercial spent nuclear
A recent sensitivity study examined the fuel characteristics.^
consequences from accident scenarios involving
bounding fuel types and illustrated the adequacy of
selecting typical fuel types for this accident scenario
analysis. Table H^ lists the radionuclide inventory
selected for estimating the accident scenario
consequences for the fuel types selected (typical
boiling-water reactor and pressurized-water reactor).
Commercial spent nuclear fuel damaged in the
accidents evaluated in this EIS could release
radionuclides from three different sources. These sources, and a best estimate of the release potential, are
as follows:
H.2.1 .4.1 .1 Crud. During reactor operation, crud (corrosion material) builds up on the outside of the
fuel rod cladding and becomes radioactive from neutron activation. Five years after discharge from the
reactor (the minimum age of any commercial spent nuclear fuel for acceptance at the repository), the
dominant radioactive constituent in the crud is cobalt-60, which accounts for 98 percent of the activity
(Sandoval et al. 1991, page 15). Cobalt-60 concentration measurements have been made on several
boiling-water and pressurized-water reactor fuel rods; the results indicate that the maximum activity
density is 0.000(X)94 curie per square centimeter for pressurized-water reactors and 0.000477 curie per
square centimeter for boiling-water reactors (Sandoval et al. 1991, pages 14 and 15). The maximum
values are about twice the average value over the length of the fuel rod (Sandoval et al. 1991, page 14).
Accordingly, the values used in these source term determinations were 0.00005 for pressurized-water
Cooling time
Burnup
Fuel type*"
(years)
(GWd/MTHM)'
PWR typical
25.9
39.56
BWR typical
27.2
32.2
a. Source: Appendix A.
b. PWR = pressurized-water reactor;
BWR = boiling-
water reactor.
c. GWd/MTHM
= gigawatt-days per metric ton of heavy
metal.
H-17
Potential Repository Accident Scenarios: Analytical Methods and Results
Table H-4. Inventory used for typical reactor fuel (curies
per assembly).'''^
Pressurized-
Boiling-water
Isotope
water reactor
reactor
Hydrogen-3
9.8x10'
3.4x10'
Carbon- 14
6.4x10"'
3.0x10"'
Chlorine-36
5.4x10-^
2.2x10"'
CobaIt-60'^
1.4x10'
2.0x10'
NickeI-59
1.3
3.5x10"'
Nickel-63
1.8x10^
4.6x10'
Selenium-79
2.3x10'
7.9x10"^
Krypton-85
9.3x10^
2.9x10^
Strontium-90
2.1x10*
7.1x10'
Zirconium-93
1.2
4.8x10"'
Niobium-93m
8.2x10'
3.5x10"'
Niobium-94
5.8x10"'
1.9x10"^
Technetium-99
7.1
2.5
Rhodium- 102
1.2x10"^
2.8x10""
Ruthenium- 106
4.8x10"'
6.7x10""
Palladium- 107
6.3x10"^
2.4x10"^
Tin- 126
4.4x10"'
1.5x10"'
Iodine- 129
1.8x10'^
6.3x10"'
Cesium- 134
1.6x10'
3.4
Cesium-135
2.5x10"'
1.0x10"'
Cesium- 137
3.1x10"
l.lxio"
Samarium- 151
1.9x10^
6.6x10'
Lead-210
2.2x10"^
9.4x10"^
Radium-226
9.3x10"^
3.7x10"''
Radium-228
1.3x10"'°
4.7x10""
Actinium-227
7.8x10"^
3.1x10"^
Thorium-229
1.7x10"'
6.1x10"*
Thorium-230
1.5x10""
5.8x10"'
Thorium-232
1.9x10"'"
6.9x10"
Protactinium-231
1.6x10"'
6.0x10"^
Uranium-232
1.9x10"^
5.5x10"'
Uranium-233
3.3x10"'
1.1x10"'
Uranium-234
6.6x10"'
2.4x10"'
Uranium-235
8.4x10"'
3.0x10"'
Uranium-236
1.4x10"'
4.8x10"^
Uranium-238
1.5x10"'
6.2x10"^
Neptunium-237
2.3x10"'
7.3x10"^
Plutonium-238
1.7x10'
5.5x10^
Plutonium-239
1.8x10^
6.3x10'
Plutonium-240
2.7x10^
9.5x10'
Plutonium-241
2.0x10"
7.5x10'
Plutonium-242
9.9x10"'
4.0x10"'
Americium-241
1.7x10'
6.8x10^
Americium-242/242m
l.lxlO'
4.6
Americium-243
1.3x10'
4.9
Curium-242
8.7
3.8
Curium-243
8.3
3.1
Curium-244
7.0x10^
2.5x10^
Curium-245
1.8x10"'
6.3x10"^
Curium-246
3.8x10"^
1.3x10"^
Curium-247
1.3x10"'
4.3x10"*
Curium-248
3.9x10"'
1.2x10"'
Californium-252
3.1x10"*
6.0x10"'
a. Source: Appendix A, except cobalt-60.
b. Inventory numbers have been rounded to two significant figures.
c. Cobalt-60 inventory in crud, as calculated in this appendix.
H-18
Potential Repository Accident Scenarios: Analytical Methods and Results
reactors and 0.00025 for boiling-water reactors. Using the fuel rod dimensions and the number of rods
per fuel assembly from Appendix A, these concentrations produce the following total inventory of cobalt-
60 for a pressurized-water reactor fuel assembly at discharge:
Cobalt-60 inventory = fuel rod surface area per assembly x cobalt-60 concentration
(per assembly) = fuel rod diameter x n
X fuel rod length x number of fuel rods per assembly
X cobalt 60 concentration
For pressurized-water reactor assemblies, the corresponding values are (from Appendix A):
Pressurized-water = 0.95 centimeters x 3.14
reactor cobalt-60 x 366 centimeters x 264 rods
inventory x 0.00005 curie per square centimeter
(per assembly) = 14 curies per pressurized-water reactor fuel assembly
(at reactor discharge)
For boiling-water reactor assemblies, the corresponding values are (from Appendix A):
Boiling-water reactor = 1 .25 centimeters x 3. 14
cobalt-60 inventory x 366 centimeters x 55 rods
(per assembly) x 0.00025 curie per square centimeter
= 20 curies per boiling-water reactor fuel assembly
(at reactor discharge)
The analysis used these concentrations, decayed to appropriate levels (25.9 years for pressurized-water
reactor fuel and 27.2 years for boiling-water reactor fuel, from Table H-3), to obtain the final cobalt-60
inventory used in the source term determination.
The amount of crud that would be released from the surface of the fuel rod cladding is uncertain because
there are very few data for the accident conditions of interest, and the physical condition of the crud can
be highly variable (Sandoval et al. 1991, page 18). Two sources (NRC 1997, Table 7-1; NRC 1998,
Table 4-1) recommend a release fraction of 1.0 (100 percent of the cobalt-60) for accident conditions;
therefore, the EIS analysis assumed this value.
Following their release from the cladding, some crud particles would be retained by deposition on the
surrounding surfaces (the fuel assembly cladding, spacer grids and structural hardware). The estimated
fraction of released particles deposited on these surfaces would be 0.9 (SNL 1987, page 5-27), resulting in
an escape fraction of 0. 1 . In accidents involving casks or canisters, additional surfaces represented by
these components would offer surfaces for further plateout.
The inhalation radiation dose from cobalt-60 (or any radioactive particle) depends on the amount of
particulate material inhaled into and remaining in the lungs (called the respirable fraction). The analysis
assumed that the respirable fraction would be 0.05 (based on Wilmot 1981, page B-3). Therefore, the
analysis assumed that the total cobalt-60 respirable airborne release fraction would be 0.005 (the escape
fraction of 0.1 multiplied by the respirable fraction of 0.05) for accident scenarios involving commercial
spent nuclear fuel assemblies.
H.2.1 .4.1 .2 Fuel Rod Gap. The space between the fuel rod cladding and the fuel pellets (called the
gap) contains fission products released from the fuel pellets during reactor operation. The only
H-19
Potential Repository Accident Scenarios: Analytical Methods and Results
potentially important radionuclides in the gap are the gases tritium (hydrogen-3) and krypton-85, and the
volatile radionuclides strontium-90, cesium-134, cesium-137, ruthenium- 106, and iodine-129 (NRC 1997,
page 7-6). The Nuclear Regulatory Commission recommends fuel rod release fractions (the fraction of
the total fuel rod inventory) of 0.3 for tritium and krypton-85, 0.000023 for the strontium and cesium
components, 0.000015 for ruthenium- 106, and 0.1 for iodine under accident conditions that rupture the
cladding (NRC 1997, page 7-6). The release fraction for the gases (tritium and krypton), as expected,
would be rather high because most of the gas would be in the fuel rod gap and under pressure inside the
fuel rod. The analysis also considered the fraction of the rods damaged in a given accident scenario.
SNL (1987, page 6-19 et seq.) assumed that the fraction of damaged fuel pins in each assembly involved
in a collision or drop accident scenario would be 20 percent. Another assessment (Kappes 1998, page 18)
assumed that any drop of the fuel rods in a fuel assembly or basket of assemblies would result in failure of
10 percent of the fuel rods, regardless of the drop distance. Because neither value seems to have a strong
basis, the EIS analysis assumed the more conservative 20-percent figure. For the particulate species
released from the gap, the analysis applied a retention factor of 0.9 (escape factor of 0.1) to account for
local deposition of the particles on the fuel assembly structures, consistent with SNL (1987, page 5-27).
SNL (1987, page 5-28) also applies a similar factor to account for retention on the failed shipping cask
structures for accident scenarios involving cask failure. However, the EIS analysis judged that this factor
does not have a strong basis, especially because the actual mode of cask failure is unknown. For accident
scenarios that could rupture the cask, surfaces on the cask structure might not be in the path of the
released material and, therefore, would not be a potential deposition site. Furthermore, particulate
material, which would escape local deposition on the fuel assembly surfaces, probably would be less
susceptible to deposition on surfaces it encountered subsequently. Therefore, the analysis assumed no
retention factor for cask structures. The final consideration is the fraction of remaining airborne
particulates that would be respirable. No specific reference could be found to the volatile materials in the
gap. The analysis conservatively assumed, therefore, that the respirable fraction would be 1.0.
H.2.1 .4.1 .3 Fuel Pellet. During reactor operation, the fuel pellets undergo cracking from thermal and
mechanical stresses. This produces a small amount of pellet particulate material that contains
radionuclides. The analysis assumed that the radionuclides are distributed evenly in the fuel pellets so
that the fractional release of the pellet particulates is equivalent to the same fractional release of the total
inventory of the appropriate radionuclides in the fuel. If the fuel cladding failed during an accident, a
fraction of these particulates would be small enough (diameter less than 10 micrometers) for release to the
atmosphere and would be respirable (small enough to remain in the lungs if inhaled). Sandia National
Laboratories estimates this fraction to be 0.000001 (SNL 1987, page 5-26) based on experiments
performed at Oak Ridge National Laboratory. The EIS used this value to develop source terms for the
accident scenarios considered. Additional particulates could be produced by pulverization due to
mechanical stresses imposed on the fuel pellets from the accident conditions. This pulverization factor
has been evaluated in SNL (1987, page 5-17) and applied in Kappes (1998, page 1-3). Based on
experimental results involving bare fuel pellets, the analysis determined that the fraction likely to be
pulverized into respirable particles would be proportional to the drop height (which is directly
proportional to energy input) and would be:
2.0 X 10"^ X energy partition factor x unimpeded drop height (centimeters) (Kappes 1998, page 1-3).
The energy partition factor is the fraction of the impact energy that is available for pellet pulverization. A
large fraction of the impact energy is expended in deforming the fuel assembly structures and rupturing
the fuel rod cladding. It has been estimated (SNL 1987, page 5-25) that the energy partition factor is 0.2.
As indicated above, some of the dispersible pellet particulates released in the accident could deposit on
surfaces in the vicinity of the damaged fuel. Consistent with the particulate material considered above,
the estimated fraction that would not deposit locally and would remain airborne would be 0. 1 based on
H-20
Potential Repository Accident Scenarios: Analytical Methods and Results
■SNL (1987, page 5-26). Based on these considerations, the respirable airborne release fraction produced
from pulverization of the fuel pellets would be:
Respirable airborne release fraction
= 2x10" X drop height (centimeters)
X energy partition factor x fraction not deposited
X fuel rod damage fraction
= 2 X 10"' X drop height
X 0.2x0.1
xO.2
-10
= 8 X 10"'" X drop height
This result is reasonably consistent with the value of 8 x 10"' from SAIC (1998, page 3-9), which is
characterized as a bounding value for the respirable airborne release fraction for accident scenarios that
would impose mechanical stress on fuel pellets for a range of energy densities (drop heights). This value
would correspond to a drop from 1,000 centimeters (10 meters or 33 feet) based on the formulation
above.
H.2.1 .4.1.4 Conclusions. Table H-5 summarizes the source term parameters for commercial spent
nuclear fuel drop accident scenarios, as discussed above.
Table H-5. Source term parameters for commercial spent nuclear fuel drop accident scenarios.
Respirable
Damage
Fraction
not
Respirable
airborne release
Radionuclide"
Location
fraction
Release fraction
deposi
ted
fraction
fraction
Co-60
Clad surface
1.0
1.0
0.1
0.05
0.005
H-3, Kr-85,
Gap
0.2
0.3
1.0
1.0
0.06
C-14
1-129
Gap
0.2
0.1
1.0
1.0
0.02
Cs-i37,Sr-90
Gap
0.2
2.3x10"'
0.1
1.0
4.6x10"'
Ru-106
Gap
0.2
1.5x10"'
0.1
1.0
3.0x10"'
All solids
Gap (existing fuel fines)
0.2
1.0x10"*
0.1
1.0
2.0x10"*
All solids
Pellet
-pulverization
0.2
4.0xl0"*xh''
0.1
I.O
S-OxlO-'^xh"
a. Abbreviations: Co = cobalt; H :
strontium; Ru = ruthenium.
b. h = drop height in centimeters.
; hydrogen (H-3 = tritium); Kr = krypton; C = carbon; I = iodine; Cs = cesium; Sr =
H.2.1 .4.2 Transporter Runaway and Derailment Accident Source Term
This accident, as noted in Section H.2. 1.3, would involve the runaway and derailment of the waste
package transporter. It assumes the ejection of the waste package from the transporter during the event;
the waste package would be split open by impact on the access tunnel wall. The calculated maximum
impact speed would be 18 meters per second (38 miles per hour) (DOE 1997b, page 98). This analysis
assumed that the source term from the damage to the 21 pressurized-water reactor fuel assemblies in the
waste package is equivalent to a drop height that would produce the same impact velocity (equivalent to
the same energy input). The equivalent drop height was computed from basic equations for the motion of
a body falling under the influence of gravity:
and.
velocity = acceleration x time
distance = Vz x acceleration x time squared
H-21
Potential Repository Accident Scenarios: Analytical Methods and Results
where: velocity = velocity of the impact (18 meters per second)
time = time required for the fall
acceleration = acceleration due to gravity (9.8 meters per second squared)
By substitution,
distance = Vz x acceleration x (velocity h- acceleration)
= (velocity)' -^ (acceleration x 2)
= (18)- -(9.8x2)
= 16 meters
Thus, the calculation of the source term for this accident scenario assumed a drop height of 16 meters and
used the parameters in Table H-5 for the various nuclide groups.
H.2.1.4.3 DOE Spent Nuclear Fuel Drop Accident Source Term
Appendix A lists the various types of DOE spent nuclear fuel and high-level radioactive waste that the
Department would place in the proposed repository. A review of the inventory indicates that the spent
nuclear fuel from the Hanford Site (N-Reactor fuel) represents a large percentage of DOE spent nuclear
fuel. The N-Reactor fuel also has one of the highest radionuclide inventories of any of the DOE spent
fuels. Although a canister of naval spent nuclear fuel would have a higher radionuclide inventory than a
canister of N-Reactor fuel (Appendix A, Table A-18), the amount of radioactive material that would be
released from a naval canister during this hypothetical accident scenario would be less than the amount
released from an N-Reactor fuel canister due to the highly robust design of naval fuel (Appendix A,
Section A. 2.2. 5. 3) (USN 1996, all). Therefore, DOE selected N-Reactor spent nuclear fuel material as the j
bounding form to represent the source term for accidents that would involve DOE material. The analysis
derived the source term for accidents involving a drop of N-Reactor fuel from DOE (1995, page 5-88),
which lists the estimated source term for a drop of a cask containing 1,000 kilograms (2,2(X) pounds) of
N-Reactor fuel from a height of 4.6 meters (15 feet). For the repository accident scenario involving
N-Reactor fuel, a total of 4,800 kilograms (10,600 pounds) of fuel would be involved in a multi-canister
overpack drop (Appendix A) from a height of 6.3 meters (21 feet), as noted above. The analysis adjusted
the DOE (1995, page 5-88) source term upward by a factor of 4.8 to account for the increased amount of
material involved (4,800 kilograms as opposed to 1,000 kilograms), and by a factor of 1.37 to account for ' ;
the increased drop height (6.3/4.6) because the analysis assumed the source term would be proportional to
the energy input, which is proportional to the drop height. These two factors were applied to the DOE
(1995, page 5-88) source term and the result is listed in Table H-6. The behavior of N-Reactor fuel
during an accident is uncertain (Kappes 1998, page 15) and the Final EIS analysis might utilize a revised
source term estimate based on the results of further studies of this fuel. Furthermore, DOE has not
developed the requirements for receipt of the fuel at the repository. These requirements could influence
the source term, as could the corresponding requirements for processing the fuel prior to shipment.
H.2.1 .4.4 Seismic Accident Scenario Source Term
Waste Handling Building. In this event, as noted in Section H.2. 1.3, the Waste Handling Building
could collapse from a beyond-design-basis earthquake. Bare fuel assemblies being transferred during the
event would be likely to drop to the floor and concrete from the ceiling could fall on the fuel assemblies,
causing damage that could result in radioactive release, which would discharge to the atmosphere through ]
the damaged roof. In addition, other radioactive material stored or being handled in the Waste Handling
Building could be vulnerable to damage. To estimate the source term, the analysis evaluated the extent of ;
damage to the fuel rods and pellets for the assemblies being transferred and then examined the other
material that could be vulnerable.
H-22
Potential Repository Accident Scenarios: Analytical Methods and Results
Table H-6. Source term used for N-Reactor Mark FV fuel
drop accident
scenario analysis (curies)."
Total
Total
Total
Radionuclide
release
Radionuclide
release
Radionuclide
release
Tritium (Hj)
1.7x10"^
Tin- 119m
1.7x10"*
Europium- 154
8.3x10"'
Carbon- 14
2.6x10""
Tin- 121m
3.0x10'
Uranium-234
1.7x10"
Iron-55
1.3x10-'
Tin-126
5.6x10"'
Uranium-235
5.7x10"*
Nickel-59
1.4x10"'
Stibium- 125 (antimony)
2.4x10"^
Uranium-236
3.3x10"'
Nickel-63
1.7x10"'
Stibium- 126
7.9x10"*'
Uranium-238
1.4x10"
Cobalt-60
5.4x10"^
Stibium- 126m
5.6x10'
Neptunium-237
2.6x10"'
Selenium-79
2.9x10"'
Tellurium- 125m
6.7x10''
Plutonium-238
7.9x10"'
Krypton-85
2.4x10"^
Iodine- 129
2.3x10''
Plutonium-239
7.3x10"'
Strontium-90
3.6
Cesium- 134
2.3x10'^
Plutonium-240
5.9x10"'
Yttriuni-90
3.6
Cesium- 135
2.6x10'
Plutonium-241
4.3
Niobium-93m
7.2x10"'
Cesium- 137
4.9
Plutonium-242
4.9x10"'
Zirconium-93
1.3x10"
Cerium- 144
8.9x10"'
Americium-241
1.7x10"'
Technetium-99
9.7x10"
Praseodymium- 144
8.9x10"'
Americium-242
3.9x10"
Ruthenium- 106
8.0x10"
Praseodymium- 144m
1.1x10"*
Americium-242m
3.9x10"
Palladium- 107
6.7x10"*
Promethium-147
2.4x10'
Americium-243
5.4x10"'
Silver- 110m
1.3x10"^
Samarium- 151
4.6x10"'
Curium-242
3.2x10"
Cadmium- 113m
1.6x10"'
Europium- 152
4.9x10"
Curium-244
2.4x10"'
a. Source: DOE (1995, page 5-88), with adjustments as noted above.
The ceiling of the transfer cell, which would consist of concrete 20 to 25 centimeters (8 to 10 inches)
thick, would be about 15 meters (50 feet) high (TRW 1999b, Attachment IV, Figure 13). Typical
pressurized-water reactor fuel assemblies weigh 660 kilograms (1,500 pounds) each (see Appendix A).
The assemblies are about 21 centimeters (8.3 inches) wide by about 410 centimeters (160 inches) long,
for an effective cross-sectional area (horizontal) of 1 square meter (11 square feet) (SNL 1987, page 5-2).
The weight of a single fuel assembly is roughly equivalent to a 25-centimeter-thick concrete block with a
1 -square-meter cross-section [about 750 kilograms (1,700 pounds) based on a density of 2.85 grams per
cubic centimeter (180 pounds per cubic foot) (CRC 1997, page 15-28)]. Thus, as a first approximation,
the analysis assumed that the concrete blocks falling from the ceiling onto the fuel assemblies would
produce about the same energy as the fuel assemblies falling from the same height.
Some of the energy imparted to the fuel assemblies from the falling debris would be absorbed in
deforming the fuel assembly structures and, thus, would not be available to pulverize the fuel pellets. As
evaluated above for falling fuel assemblies, this energy absorption factor would result in an estimated
20 percent of the energy being imparted to the pellets and the rest absorbed by the structure (SNL 1987,
page 5-25). Finally, as noted above, the analysis used a 0.1 release factor (0.9 retention) to represent the
retention of the released fuel particles by deposition on the cladding and other fuel assembly structures
(SNL 1987, page 5-27). In addition, it assumed that additional retention would be associated with the
concrete and other rubble that would be on top, or in the vicinity, of the fuel assemblies. It assumed this
release factor would be 0.1 (0.9 retention) consistent with that used by SNL (1987, page 5-28) for
retention by deposition on the cask and canister materials that surround the fuel assemblies during
accident scenarios. It also assumed a fuel pellet pulverization factor of 8 x 10"'" x h, the same as that used
for fuel assembly drop accident scenarios. Thus, the overall pellet respirable airborne release fraction for
the fuel pellet particulates is:
Respirable airborne release fraction = 8 x 10"'° x drop height (centimeters) x rubble retention
= 8x 10"'° X 1,500x0.1
= 1.2 X 10"^
Other radioactive materials either stored or being handled in the Waste Handling Building could also be at
risk. For material in casks and canisters and waste packages, the analysis assumed that the damage
H-23
Potential Repository Accident Scenarios: Analytical Methods and Results
potential from falling debris would not be great enough to cause a large radionuclide release. This is
based on the fact that canisters and casks are quite robust and that, even if the containers were breached
by the energy of the impact, there would be very little energy remaining to cause fuel pellet pulverization.
There could be, however, bare fuel assemblies exposed in the dryers and in disposal containers awaiting
lid attachment. An estimated 375 bare pressurized-water reactor fuel assemblies could be exposed to
falling debris (Montague 1999, page 1). The location of this material would be as follows:
• Assembly transfer system dryers: 25 pressurized-water reactor assemblies
• Disposal canister handling system welding stations: 346 pressurized-water reactor assemblies
• Transfer operations: four pressurized-water reactor assemblies
Because the concrete roof heights over these areas would be roughly the same as the assembly transfer
system area in the Waste Handling Building [15 meters (50 feet)] where the analysis assumed the four
bare pressurized-water reactor assemblies would be involved, the analysis assumed the pellet
pulverization contribution to the source term to be equivalent to that for the fuel assemblies being
transferred. The overall source term, then, was determined by assuming 375 typical pressurized-water
reactor assemblies with the release fractions listed in Table H-5.
Boiling-water reactor fuel assemblies could be exposed at these areas, but the analysis evaluated only
pressurized-water reactor fuel assemblies because they would result in a slightly higher source term under
equivalent accident conditions and would be more likely to be involved because they would comprise a
larger amount of material (see Appendix A) to be received at the repository. Thus, the source term for the
seismic event would be 375 typical pressurized-water reactor fuel assemblies (Table H-4) with release
fractions based on Table H-5.
Waste Treatment Building. It is assumed that the radionuclide concentration for the dry compactable
waste in the Waste Treatment Building would be similar to that for power reactors (McFeely 1998,
page 2). This material would consist of paper, plastic, and cloth with a specific activity of 0.025 curie per
cubic meter (0.7 millicurie per cubic foot) (McFeely 1998, page 2). This activity would consist primarily
of cobalt isotopes (primarily cobalt-60) representing 67 percent of the total activity, and cesium, which
would contribute 28 percent of the total (McFeely 1999, all).
The Waste Treatment Building would operate a single shift per day, and would continuously process
waste such that no large accumulation would occur. Because Waste Handling Building operations would
be likely to involve three shifts per day (TRW 1999b, Section 6.2), the analysis assumed that three shifts
of solid waste would accumulate before the Waste Treatment Building began its single-shift operation.
The generation rate of solid compactible waste would be about 1,500 cubic meters (53,000 cubic feet) per
year (DOE 1997a, page 32) or about 0.17 cubic meter (5.8 cubic feet) per hour. Thus, three shifts (24
hours) of Waste Handling Building operation would produce about 4.0 cubic meters (140 cubic feet) of
solid compactible waste. The total radionuclide inventory in this waste would be:
Cobalt-60 = 4.0 cubic meters x 0.025 curie per cubic meters x 0.67 (cobalt-60 fraction)
= 0.07 curie
Cesium- 137 = 4.0 cubic meters x 0.025 curie per cubic meters x 0.28 (cesium- 137 fractions)
= 0.03 curie
H-24
Potential Repository Accident Scenarios: Analytical Methods and Results
The respirable airborne release fraction for a fire involving combustible low-level waste has been
conservatively estimated at 0.4 (Mueller et al. 1996, page D-21). Thus, the respirable airborne release
source term for the fire accident scenario would be:
Cobalt-60 = 0.07 curie x 0.4 = 0.028 curie
Cesium- 137 = 0.03 curie x 0.4 = 0.0 1 2 curie
The assumed release height for the accident scenario is 2 meters (6.6 feet). This is the minimum release
height for the consequences analysis and represents a ground-level release.
H.2.1.4.5 Low-Level Waste Drum Failure Source Term
As indicated in Section H.2.1.2, the most meaningful accident scenarios involving exposure to workers
would be those related to puncture or rupture of waste drums that contained low-level waste. Such events
could occur during handling operations and probably would involve the puncture of a drum by a forklift,
or the drop of the drum during stacking and loading operations.
Two types of waste drums would contain the processed waste. Concentrated liquid waste would be
mixed with cement and poured into 0.21 -cubic-meter (55-gallon) drums. Compacted and noncompacted
solid waste would also be placed in the same drums, which would, in turn, be placed in 0.32-cubic-meter
(85-gallon) drums with the space between the two drums grouted. The probability of a drum failure was
analyzed for these two drum types.
Following a drum failure, some fraction of the radionuclides in the waste would be released and workers
in the immediate vicinity could be exposed to the material. The amount released would depend on the
radionuclide concentration in the low-level waste material, the fraction of low-level waste released from
the drum on its failure, and the respirable airborne release fraction from the released waste.
For liquid waste, the concentration of radionuclides is expected to be (McFeely 1998, page 3):
Cobalt-60 = 0.001 curie per cubic meter
Cesium- 137 = 0.0015 curie per cubic meter
As noted in Section H.2.1.2, the evaporator would concentrate the liquid waste down to 10 percent of the
original generated so the concentration of radionuclides in the waste would be increased to:
Cobalt 60 = 0.01 curie per cubic meter
Cesium-137 = 0.015 curie per cubic meter
The grouting operation would dilute this concentration somewhat by adding cement, but this dilution
has been ignored for conservatism.
The total activity in a 0.21 -cubic meter (55-gallon) drum would become:
Cobalt-60 = 0.01 curie per cubic meter x 0.21 cubic meter
= 0.0021 curie per drum
Cesium-137 = 0.015 curie per cubic meter x 0.21 cubic meter
= 0.0032 curie per drum
H-25
Potential Repository Accident Scenarios: Analytical Methods and Results
For dry compacted waste, the total inventory in a 0.21 -cubic-meter (55-gallon) drum would be
Cobalt-60 = 0.21 cubic meter x 0.025 curie per cubic meter x 0.67 (cobalt-60 fraction)
= 0.0035 curie
Cesium-137 = 0.21 cubic meter x 0.025 curie per cubic meter x 0.28 (cesuim-137 fraction)
= 0.0015 curie
The estimated amount of material released from drums containing solid waste is 25 percent of the
contents based on Mueller et al. (1996, page 94). Values from Mueller et al. (1996, all) were used for the
respirable airborne release fraction. For dry waste, the recommended respirable airborne release fraction
is 0.001. For grouted liquid waste, this fraction is determined by the following equation:
Respirable airborne release fraction = AxDxGxH
where:
A = constant (2.0 x 10"") (Mueller et al. 1996, page D-25)
D = material density [3.14 grams per cubic centimeter (196 pounds per cubic foot)]
(McFeely 1998, all)
G = gravitational acceleration [980 centimeters (32.2 feet) per second squared]
H = height of fall of the drum in the accident scenario
The assumed height of the fall is 2 meters (6.6 feet), which would be the approximate maximum lift
height when the drum was stacked on another drum or placed on a carrier for offsite transportation. This
same formula applies to drum puncture accident scenarios (Mueller et al. 1996, page D-30), and the
2-meter drop event would be equivalent in damage potential to a forklift impact at about 4.5 meters per
second (10 miles per hour). The respirable airborne release fraction for this case then becomes:
Respirable airborne release fraction = 2.0 x lO" x 3. 14 x 980 x 200
= 1.23x10'^
Based on these results, the worker risk would be dominated by accidents involving drums that contained
dry waste because both the frequency of the event [0.59 versus 0.46 (Section H.2.1.2)] and the release
fraction [1 x 10'^ versus 1.23 x 10"^ (derived above)] would be greater. The total amount of airborne
respirable material release (source term) for the risk-dominant dry waste accident scenario would be:
Cobalt-60 = 0.0035 curie (total drum inventory) x 0.25 (fraction released)
X 0.001 (respirable airborne release fraction)
= 8.5 X 10'^ curies
Cesium-137 = 0.0015 curie (total drum inventory) x 0.25 (fraction released)
X 0.001 (respirable airborne release fraction)
= 3.8 X 10"^ curies
The analysis assumed that, following normal industrial practice, workers would not be in the area beneath
suspended objects. Accordingly, the nearest worker was assumed to be 5 meters (16 feet) from the
impact area. Therefore, the volume assumed for dispersion of the material prior to reaching the worker
would be 125 cubic meters (4,400 cubic feet), which represents the immediate vicinity of the accident
H-26
:i
Potential Repository Accident Scenarios: Analytical Methods and Results
location [a volume approximately 5 meters (16 feet) by 5 meters by 5 meters]. The breathing rate of the
worker would be 0.00035 cubic meter (about 0.012 cubic foot) per second (ICRP 1975, page 346).
H.2.1 .5 Assessment of Accident Scenario Consequences
Accident scenario consequences were calculated as individual doses (rem), collective doses (person-rem),
and latent cancer fatalities. The receptors considered were (1) the maximally exposed offsite individual,
defined as a hypothetical member of the public at the point on the proposed repository land withdrawal
boundary who would receive the largest dose from the assumed accident scenario (a minimum distance of
1 1 kilometers (7 miles), (2) the maximally exposed involved worker, the hypothetical worker who would
be nearest the spent nuclear fuel or high-level radioactive waste when the accident occurred, (3) the
noninvolved worker, the hypothetical worker near the accident but not involved in handling the material,
assumed to be 100 meters (about 330 feet) from the accident, and (4) the members of the public who
reside within about 80 kilometers (50 miles) of the proposed repository.
For radiation doses below about 20 rem and low dose rates (below 10 rem per hour), potential health effects
would be those associated with a chronic exposure or an increase in the risk of fatal cancer (ICRP 1991,
Chapter 3) (see the discussion in Appendix F, Section F.l). The International Committee on Radiation
Protection has recommended the use of a conversion factor of 0.(XX)5 fatal cancer per person-rem for the
general population for low doses, and a value of O.CXXM fatal cancer per person-rem for workers for chronic
exposures. The higher value for the general population accounts in part for the fact that the general
population contains young people, who are more susceptible to the effects of radiation. These conversion
factors were used in the EIS consequence analysis. The latent cancer fatality caused by radiation exposure
could occur at any time during the remaining lifetime of the exposed individual. As dose increases above
about 15 rem over a short period (acute exposures), observable physical effects can occur, including
temporary male sterility (ICRP 1991, page 15). At even higher acute doses (above about 500 rem), death
within a few weeks is probable (ICRP 1991, page 16).
DOE used the MACCS2 computer program (Rollstin, Chanin, and Jow 1990, all; Chanin and Young
1998, all) and the radionuclide source terms for the identified accident scenarios in Section H.2.1.4 to
calculate consequences to receptors. This program, developed by the U.S. Nuclear Regulatory
Commission and DOE, has been widely used to compute radiological impacts from accident scenarios
involving releases of radionuclides from nuclear fuel and radioactive waste. DOE used this program for
offsite members of the public, the maximally exposed offsite individual, and the noninvolved worker.
The MACCS2 program calculates radiological doses based on a sampling of the distribution of weather
conditions for a year of site-specific weather data. Meteorological data were compiled at the proposed
repository site from 1993 through 1997. This analysis used the weather conditions for 1993. The
selection of 1993 was based on a sensitivity analysis that showed that, on the average, the weather
conditions for 1993 produced somewhat higher consequences than those for the other years for most
receptors, although the variation from year to year was small.
For exposure to inhaled radioactive material, it was assumed (in accordance with U.S. Environmental
Protection Agency guidance) that doses would accumulate in the body for a total of 50 years after the
accident (Eckerman, Wollbarst, and Richardson 1988, page 7). For external exposure (from ground
contamination and contaminated food consumption), the dose was assumed to accumulate for 70 years
(DOE 1993, page 21).
The MACCS2 program provides doses to selected receptors for a contiguous spectrum of site-specific
weather conditions. Two weather cases were selected for the EIS: (1) a median weather case (designated
at 50 f)ercent) that represents the weather conditions that would produce median consequences to the
H-27
Potential Repository Accident Scenarios: Analytical Methods and Results
receptors, and (2) a 95 percent weather case that provides higher consequences that would only be
exceeded 5 percent of the time.
The MACCS2 program is not suitable for calculating doses to receptors near the release point of
radioactive particles [within about 100 meters (330 feet)]. For such cases, the analysis calculated
involved worker dose estimates using a breathing rate of 0.00035 cubic meter (0.012 cubic foot) per
second (ICRP 1975, page 346). For involved worker dose calculations from accident scenarios in the
cask transfer and handling area, the analysis assumed that the worker would be a minimum of 4.6 meters
(15 feet) from the location of the cask impact with the floor during the accident (normal industrial
practice would preclude workers from being in the immediate vicinity of areas where heavy objects could
strike the floor during lifting operations). Because of the perceived hazard following a breached cask, the
analysis assumed that the worker would immediately vacate the area after observing that the cask had
ruptured. Accordingly, the analysis assumed that the worker would breathe air containing airborne
radioactive material from the ruptured cask for 10 seconds.
For involved worker doses from the drum handling accident scenario, the analysis assumed that the
worker (a forklift operator) would be 3 meters (10 feet) from the drum rupture location, and would
breathe air containing radioactive material from the ruptured drum for 30 seconds.
The involved worker dose estimates used the same dose conversion factors as those used by the MACCS2
program for inhalation exposure.
The analysis assumed that the population around the repository would be that projected for the year 2000
(see Appendix G, Table G-44). The exposed population would consist of individuals living within about
80 kilometers (50 miles) of the repository, including pockets of people who would reside just beyond the
80-kilometer distance. The dose calculations included impacts from the consumption of food
contaminated by the radionuclide releases. The contaminated food consumption analysis used site-
specific data on food production and consumption for the region around the proposed site (TRW 1997b,
all). For conservatism, the analysis assumed no mitigation measures, such as post-accident evacuation or
interdiction of contaminated foodstuffs. However, DOE would take appropriate mitigation actions in the
event of an actual release.
The results of the consequence analysis are listed in Tables H-7 (for 50-percent weather) and H-8 (for
95-percent weather). These tables list doses in rem for individual receptors and in person-rem (collective
dose to all exposed persons) for the 80-kilometer (50-mile) population around the site. For selected
receptors, as noted, the tables list estimated latent cancer fatalities predicted to occur over the lifetime of
the exposed receptors as a result of the calculated doses using the conversion factors described in this
section. These estimates do not consider the accident frequency. For comparison, in 1995 the lifetime
incidence of fatal cancer from all causes for Nevada residents was 0.24 (CDC 1998, page 215). Thus, the
estimated latent cancer fatalities for the individual receptors from accidents would be very small in
comparison to the cancer incidence from other causes. For the 28,000 persons living within 80 kilometer
of the site (see Appendix G), 6,720 (28,000 x 0.24) would be likely to die eventually of cancer. The
accident of most concern for the 95-percent weather conditions (earthquake. Table H-8, number 14)
would result only in an estimated 0.0072 latent cancer fatality for this same population.
H.2.2 NONRADIOLOGICAL ACCIDENT SCENARIOS
A potential release of hazardous or toxic materials during postulated operational accident scenarios at the i
repository would be very unlikely. Because of the large quantities of radioactive material, radiological
considerations would outweigh nonradiological concerns. The repository would not accept hazardous
waste as defined by the Resource Conservation and Recovery Act (40 CFR Parts 260 to 299). Some
f
H-28
Potential Repository Accident Scenarios: Analytical Methods and Results
Table H-7. Radiological consequences of repository operations accidents for median (50th-percentile)
meteorological conditions.
Maximally exposed Noninvolved
offsite individual Population worker Involved worker
Frequency Dose Dose Dose Dose
Accident scenario'^^'^ (per year)" (rem) LCFi"* (person-rem) LCFp"* (rem) LCFi (rem) LCFi
1. 6.9-meter drop of shipping 4.5x10*' 1.9xl0' l.OxlO' S.SxlO'^ 2.7x10-' 9.4xI0' S-SxlO-*' 76 3.0x10"^
caskinCTHA-61 BWR
assemblies-no filtration
2. 7.1 -meter drop of shipping 6.1x10'' 2.3x10 ' 1.2x10* 6.6x10'^ 3.3x10'' 1.1 4.4x10" 90 3.6x10^
cask in CTHA-26 PWR
assemblies-no filtration
3. 4. 1 -meter drop of shipping 1.4x10-' 1.3x10'' 6.5x10'"' 3.9x10'^ 2.0x10'' 5.7x10'' 2.3x10'" 46 1.8x10'^
caskinCTHA-61 BWR
assemblies- no filtration
4. 4.1 -meter drop of shipping 1.9xl0' 1.4x10'' 7.0x10'' 4.6x10'" 2.3x10'' 6.6x10'' 2.6x10'" 53 2.1x10'"
cask in CTHA-26 PWR
assemblies-no filtration
5. 6.3-meter drop of MCO in 4.5xI0" 3.7x10'' 1.9x10'"' l.lxlO' 5.3xlO' l.lxlO" 4.4x10'' (e) (e)
CTS-ION-Reactorfuel
canisters-filtration
6. 6.3-meter drop of MCO in 2.2x10'' 1.2x10'' 6.0x10'' 3.4x10'^ 1.7x10'' 3.6xlO' 1.4x10'" (e) (e)
CTS-lON-reactorfuel
canisters-no filtration
7. 5-meter drop of transfer basket 1.1x10^ 6.6x10 ' 3.3x10 '° 4.0xl0" 2.0x10' 1.7x10'" 6.8x10 ' (e) (e)
in ATS-8 PWR assemblies-
filtration
8. 5-meter drop of transfer basket 2.8x10' 5.6x10'" 2.8xl0' 1.7x10'^ 8.6x10* 1.6x10'' 6.4xl0' (e) (e)
in ATS-8 PWR assemblies-no
filtration
9. 7.6-meter drop of transfer 7.4x10'' 5.1x10'' 2.6x10'"' 2.9x10'" 1.5xlO' 1.3x10'" 5.2x10'^ (e) (e)
basket in ATS- 16 BWR
assemblies-filtration
10. 7.6-meter drop of transfer 1.9xl0' 6.1x10" 3.1x10'' 1.6x10'^ 8.2x10* 1.8x10' 7.2xl0' (e) (e)
basket in ATS-16 BWR fuel
assemblies-no filtration
11. 6-meter drop of disposal l.SxlO' 1.8x10* gOxlO'" l.OxlO' 5.2x10'' 5.0xl0" 2.0xl0' (e) (e)
container in DCHS-21 PWR
assemblies-filtration
12. 6-meter drop of disposal 8.6xlO' 1.7x10'' 8.5xl0' 5.1x10" 2.5xl0' 5.1x10'' 2.0xlO" (e) (e)
container in DCHS-21 PWR
fuel assemblies-no filtration
13. Transporter runaway and 1.2x10' 4.3x10'' 2.2x10* l.lxlO' 5.4xl0' 1.5 6.0xl0" (f) (0
derailment in access tunnel-21
PWR assemblies-filtration- 16-
meter drop height equivalent
14. Earthquake - 375 PWR 2.0x10'' 9.1x10'' 4.6x10* 3.6x10'' 1.8x10" 8.3 3.3x10'' (f) (0
assemblies
15. Earthquake w/fire in WTB 2.0xI0' 1.8xl0' 9.0x10'' 6.3x10" 3.2xlO' 5.2x10'' 2.1x10* (f) (0
16. LLW drum rupture in WTB 0.59 6.1xl0'° 3.1xlO" 2.1x10^ l.lxlO" 1.4xl0' 5.6x10" 7.0xl0' 2. 8x10'^
a. Source: Kappes (1998. all). These frequency estimates are highly uncertain due to the preliminary nature of the repository design and are
provided only to show potential accident sequence credibility. They represent conservative estimates based on the approach taken in
Kappes (1 998, all).
b. CTHA = Cask Transfer/Handling Area, CTS = Canister Transfer System, ATS = Assembly Transfer System, DCHS = Disposal Container
Handling System, WTB = Waste Treatment Building.
c. To convert meters to feet, multiply by 3.2808.
d. LCFi is the likelihood of a latent cancer fatality for an individual who receives the calculated dose. LCFp is the number of cancers probable
in the exposed population from the collective population dose (person-rem). These values were computed based on a conversion of dose in
rem to latent cancers as recommended by the International Council on Radiation Protection as discussed in this section.
e. For these cases, the involved workers are not expected to be vulnerable to exposure during an accident because operations are done
remotely. Thus, involved worker impacts were not evaluated.
f For these events, involved workers would likely be severely injured or killed by the event; thus, no radiological impacts were evaluated.
For the seismic event, as many as 39 people could be injured or killed in the Waste Handling Building, and as many as 36 in the Waste
Treatment Building based on current staffing projections (TRW 1998c, pages 17 and 18).
H-29
Potential Repository Accident Scenarios: Analytical Methods and Results
Table H-8. Radiological consequences of repository operations accidents for unfavorable (95th-
percentile) meteorological conditions.
Maximally exposed
ofFsite individual
Population
Noninvolved
worker Involved worker
Accident scenario^''''^
Frequency Dose Dose Dose Dose
(per year)' (rem) LCFi" (person-rem) LCFp'' (rem) LCFi (rem) LCFi
1 . 6.9-meter drop of shipping
caskinCTHA-61 BWR
assemblies-no filtration
2. 7.1 -meter drop of shipping
cask in CTHA-26 PWR
assemblies-no filtration
3. 4.1 -meter drop of shipping
caskinCTHA-61 BWR
assemblies-no filtration
4. 4.1 -meter drop of shipping
cask in CTHA-26 PWR
assemblies-no filtration
5. 6.3-meter drop of MCO in
CTS-ION-Reactorfiiel
canisters-filtration
6. 6.3-meter drop of MCO in
CTS-lON-reactorfiiel
canisters-no filtration
7. 5-meter drop of transfer
basket in ATS-8 PWR
assemblies- filtration
8. 5-meter drop of transfer
basket in ATS-8 PWR
assemblies-no filtration
9. 7.6-meter drop of transfer
basket in ATS- 1 6 BWR
assemblies-filtration
10. 7.6-meter drop of transfer
basket in ATS- 16 BWR ftiel
assemblies-no filtration
1 1 . 6-meter drop of disposal
container in DCHS-21 PWR
assemblies-filtration
12. 6-meter drop of disposal
container in DCHS-21 PWR
fiiel assemblies-no filtration
13. Transporter runaway and
derailment in access tunnel-
21 PWR assemblies-
filtration- 1 6-meter drop
height equivalent
14. Earthquake -375 PWR
assemblies
15. Earthquake w/fire in WTB
16. LLW drum rupture in WTB
4.5x10-^ 7.2x10-' 3.5x10"° 1.7
?.6xl0-^ 5.1 2.0x10' 76 3.0x10
-3
6.1x10"* 8.0x10-^ 4.0X10-* 2.1 1.1x10"' 5.9 2.4x10"' 90 3.6x10"^
1.4x10"' 4.3x10"' 2.2x10"* 1.3 6.5x10"* 3.1 1.2x10"' 46 1.8x10"^
1.9x10' 5.2x10"' 2.6x10"* 1.5 7.8x10"^ 3.5 1.4x10"' 53 2.1x10"-
4.5x10"^ 1.2x10"* 6.0x10"'° 2.6x10"* 1.3x10"' 3.3x101.3x10"' (e) (e)
2.2x10"' 4.3x10"' 2.2x10"* 8.6x10"' 4.3x10"^ 1.1 4.4x10"^ (e)
1.1x10"^ 2.5x10"* 1.3x10"' 3.3x10"^ 1.6x10"^ 4.6x101.8x10"' (e)
2.8x10"' 2.1x10"' 1.1x10"* 5.6x10"' 2.8x10"^ 4.6x10 1.8x10"^ (e)
7.4x10"' 2.1x10"* 1.1x10"'
2.4x10"^ 1.2x10"^ 3.8x101.5x10"' (e)
(e)
(e)
(e)
(e)
I
1.9x10"' 2.2x10"' 1.1x10"* 5.1x10"' 2.6x10"^ 5.1x10 2.0x10"^ (e) (e)
1.8x10"' 7.3x10* 3.7x10"' 8.6x10"^ 4.3x10"^ 1.3x10-5.2x10-' (e) (e)
8.6x10"' 6.1x10"' 3.1x10"* 1.6 8.0x10"^ 1.3 5.2x10-4 (e) (e)
1.2x10"' 1.3x10"^ 6.5x10* 3.2 1.6x10"' 3.9 1.6x10"' (f) (f)
2.0x10"' 3.2x10"^ 1.6x10"' 14
7.2x10"' 7.0 2.8x10"^ (f) (f)
I
2.0x10"^ 5.8x10"' 2.9x10"* 2.1 l.lxlQ-' 5.2x102.1x10"* (f) (f)
0.59 1.9x10"' 9.5x10"" 7.5x10"' 3.7x10"'" 1.4x10-5.6x10"" 7.0x10"' 2.8x10"*
a. Source: Kappes (1998, all). These frequency estimates are highly uncertain due to the preliminary nature of the repository design and are
provided only to show potential accident sequence credibility. They represent conservative estimates based on the approach taken in
Kappes (1998, all).
b. CTHA = Cask Transfer/Handling Area, CTS = Canister Transfer System, ATS = Assembly Transfer System, DCHS = Disposal Container
Handling System, WTB = Waste Treatment Building.
c. To convert meters to feet, multiply by 3.2808.
d. LCFi is the likelihood of a latent cancer fatality for an individual who receives the calculated dose. LCFp is the number of cancers probable
in the exposed population from the collective population dose (person-rem). These values were computed based on a conversion of dose in
rem to latent cancers as recommended by the International Council on Radiation Protection, as discussed in this section.
e. For these cases, the involved workers are not expected to be vulnerable to exposure during an accident since operations are done remotely.
Thus, involved worker impacts were not evaluated.
f For these events, involved workers would likely be severely injured or killed by the event; thus, no radiological impacts were evaluated.
For the seismic event, as many as 39 people could be injured or killed in the Waste Handling Building, and as many as 36 in the Waste
Treatment Building based on current staffing projections (TRW 1998c, pages 17 and 18).
H-30
Potential Repository Accident Scenarios: Analytical Methods and Results
potentially hazardous metals such as arsenic or mercury could be present in the high-level radioactive
waste. However, they would be in a solid glass matrix that would make the exposure of workers or
members of the public from operational accidents highly unlikely. Appendix A contains more
information on the inventory of potentially hazardous materials.
Some potentially nonradioactive hazardous or toxic substances would be present in limited quantities at
the repository as part of operational requirements. Such substances would include liquid chemicals such
as cleaning solvents, sodium hydroxide, sulfuric acid, and various solid chemicals. These substances are
in common use at other DOE sites. Potential impacts to workers from normal industrial hazards in the
workplace including workplace accidents were derived from DOE accident experience at other sites.
These impacts include those from accident scenarios involving the handling of hazardous materials and
toxic substances as part of typical DOE operations. Thus, the industrial health and safety impacts to
workers include impacts to workers from accidents involving such substances.
Impacts to members of the public would be unlikely because the hazardous materials would be mostly
liquid and solid so that a release would be confined locally. (For example, chlorine used at the site for
water treatment would be in powder form, so a gaseous release of chlorine would be unlikely.
Furthermore, the repository would not use propane as a heating fuel, so no potential exists for propane
explosions or fires.) The potential for hazardous chemicals to reach surface water during the Proposed
Action would be limited to spills or leaks followed immediately by a rare precipitation or snow melt event
large enough to generate runoff. Throughout the project, DOE would install engineered measures to
minimize the potential for spills or releases of hazardous chemicals and would comply with written plans
and procedures to ensure that, if a spill did occur, it would be properly managed and remediated. The
Spill Prevention Control and Countermeasures Plan that would be in place for Yucca Mountain activities
is an example of the plans DOE would follow under the Proposed Action.
The construction phase could generate as many as 3,500 drums [about 730 cubic meters (26,000 cubic
feet)] of solid hazardous waste, and emplacement operations could generate as much as 100 cubic meters
(3,500 cubic feet) per year (TRW 1999b, Section 6.1). Maintenance operations and closure would
generate similar or smaller waste volumes. DOE would accumulate this waste in onsite staging areas in
accordance with the regulations of the Resource Conservation and Recovery Act. Emplacement and
maintenance operation could generate as many as 2,700 liters (1,700 gallons) of liquid hazardous waste
annually (TRW 1999b, Section 6.1). The construction and closure phases would not generate liquid
hazardous waste. The generation, storage, packaging, and shipment off the site of solid and liquid
hazardous waste would present a very small potential for accidental releases and exposures of workers.
Although a specific accident scenario analysis was not performed for these activities, the analysis of
human health and safety (see Chapter 4, Section 4.1.7.3) included these impacts to workers implicitly
through the use of a data base that includes impacts from accidents involving hazardous and toxic
materials. Impacts to members of the public would be unlikely.
H.3 Accident Scenarios During Retrieval
During retrieval operations, activities at the repository would be essentially the reverse of waste package
emplacement, except operations in the Waste Handling Building would not be necessary because the
waste packages would not be opened. The waste packages would be retrieved remotely from the
emplacement drifts, transported to the surface, and transferred to a Waste Retriev ' Storage Facility
(TRW 1999b, Attachment I). This facility would include a Waste Retrieval Transfer Building where the
waste packages would be unloaded from the transporter, transferred to a concrete storage unit, and moved
to a concrete storage pad. The storage pad would be a 24- by 24-meter (80- by 80-foot) pad, about
1 meter (3.3 feet) thick, which probably would be about 3 kilometers (2 miles) over flat terrain from the
H-31
Potential Repository Accident Scenarios: Analytical Methods and Results
North Portal. Each storage pad would contain 14 waste packages. The number of pads required would
depend on how many waste packages would be retrieved.
Because retrieval operations would be essentially the reverse of emplacement operations, accidents
involving the disposal container during emplacement bound the retrieval operation. The bounding
accident scenario during emplacement of the disposal container would be transporter runaway and
derailment in the access tunnel (see Section H.2. 1.4). This accident scenario would also bound accident
scenarios during retrieval.
During storage, no credible accidents resulting in radioactive release of any measurable consequence
would be expected to occur. This prediction is based on an evaluation of above-ground dry storage
accident scenarios at the commercial sites under similar conditions, as evaluated in Appendix K.
In view of these considerations, DOE has concluded that the waste transporter derailment and the rockfall
accident scenarios analyzed in Section H.2 would bound accident impacts during retrieval.
H.4 Accident Scenarios During IVIonitoring and Closure
During monitoring and closure activities, DOE would not move the waste packages, with the possible
exception of removing a container from an emplacement drift for examination or drift maintenance. Such
operations could result in a transporter runaway and derailment accident, but the frequency of release
from such an event would be extremely low, as would the consequences, resulting in minimal risk. Thus,
DOE expects the radiological impacts from operations during monitoring and closure to be very small.
H.5 Accident Scenarios for Inventory Modules 1 and 2
Inventory Modules 1 and 2 are alternative inventory options that the EIS considers. These modules
involve the consideration of additional waste material for emplacement in the repository. They would
involve the same waste and handling activities as those for the Proposed Action, but the quantity of
materials received would increase, as would the period of emplacement operations. The analysis assumed
the receipt and emplacement rates would remain the same as those for the Proposed Action. Therefore,
DOE expects the accident impacts evaluated for the Proposed Action to bound those that could occur for
Inventory Modules 1 and 2 because the same set of operations would be involved.
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Kappes 1998
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Potential Repository Accident Scenarios: Analytical Methods and Results
NRC 1998
Ramsdell and Andrews 1986
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H-36
Potential Repository Accident Scenarios: Analytical Methods and Results
TRW 1998c
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H-37
'■TTT/T?^^
Appendix I
Environmental Consequences
of Long-Term Repository
Performance
■
Environmental Consequences of Long-Term Repository Performance
TABLE OF CONTENTS
Section Page
1.1 Long-Term Repository Performance Assessment Calculations I-l
1.2 Total System Performance Assessment Methods and Models 1-2
1.3 Inventory 1-5
1.3.1 Waterbome Radioactive Materials 1-6
1.3. 1 . 1 Reduction of the List of Radionuclides for Performance Assessment
Modeling 1-6
1.3.1.2 Radionuclide Inventory Used in the Performance Assessment Model 1-8
1.3.1.2.1 Commercial Spent Nuclear Fuel 1-9
1.3.1.2.2 DOE Spent Nuclear Fuel 1-9
1.3.1.2.3 High-Level Radioactive Waste 1-9
1.3. 1 .2.4 Greater-Than-Class-C and Special-Performance-Assessment-Required
Wastes 1-13
1.3.2 Waterbome Chemically Toxic Materials 1-13
1.3.2.1 Identification of Waterbome Chemically Toxic Materials 1-14
1.3.2.2 Screening Criteria 1-14
1.3.2.3 Screening Application 1-15
1.3.2.3.1 Solubility of Chemically Toxic Materials in the Repository 1-15
1.3.2.3.2 Well Concentration of Chemically Toxic Materials 1-18
1.3.2.3.3 Health Effects Screening for Chemically Toxic Materials 1-19
1.3.2.4 Chromium Inventory for Use in the Performance Assessement Model 1-20
1.3.2.5 Elemental Uranium Inventory for Use in the Performance Assessment Model 1-24
1.3.2.6 Molybdenum Inventory 1-24
1.3.3 Atmospheric Radioactive Materials 1-25
1.4 Extension of Total System Performance Assessment Methods and Models for
EIS Analyses 1-25
1.4.1 Repository Design for Alternative Thermal Loads 1-26
1.4.2 Thermal Hydrology Model 1-27
1.4.2.1 Thermal-Hydrologic Scenarios 1-27
1.4.2.2 Waste Package and Drift Geometry 1-28
1.4.2.3 Selection of Submodels 1-29
1.4.2.4 Hydrology and Climate Regime 1-31
1.4.2.5 Treatment of Edge Effects 1-32
1.4.2.5.1 Scaling Factors for Edge Effects 1-32
1.4.2.5.2 Definition of Thermal-Hydrologic Zones 1-33
1.4.2.6 Results 1-33
1.4.2.6.1 Variability Among the Waste Packages 1-33
1.4.2.6.2 Sensitivity to Thermal Loads 1-33
1.4.2.6.3 Comparison Between Center and Edge Locations 1-34
1.4.3 Waste Package Degradation Model 1-34
1.4.3. 1 WAPDEG Development and Application to Total System Performance
Assessment - Viability Assessment 1-34
1.4.3.2 Application of WAPDEG for the EIS 1-37
1.4.3.3 Results 1-38
1.4.3.4 Discussion 1-39
1.4.4 Waste Form Dissolution Models 1-40
1.4.4.1 Spent-Fuel Dissolution Model 1-40
1.4.4.2 High-Level Radioactive Waste Glass 1-41
1.4.4.3 Greater-Than-Class-C and Special-Performance-Assessment-Required Waste 1-41
1.4.5 RIP Model Modifications 1-41
1.4.5.1 Modifications to the RIP Model in the Repository Environment 1-41
I-iii
Environmental Consequences of Long-Term Repository Performance
Section Page
1.4.5.2 Modifications to Input and Output FEHM Model 1-42
1.4.5.3 Modifications to Saturated Zone Stream Tubes for Different Repository
Areas 1-43
1.4.5.4 Modifications to the Stream Tubes for Distances Other Than 20 Kilometers 1-45
1.4.5.5 Modifications to the RIP Model to Account for Unsaturated Zone and
Saturated Zone Particle Transport 1-47
1.4.5.6 Biosphere Dose Conversion Factors for Waterbome Radionuclides 1-48
1.5 Waterbome Radioactive Material Impacts 1-50
1.5.1 Total Releases During 10,000 Years and 1 Million Years 1-50
1.5.2 Apparent Anomalous Behavior Between Low and Intermediate Thermal
Load Results for Proposed Action Inventory I-5I
L5.2.1 Effect of the Dilution Factor 1-52
1.5.2.2 Effect of Waste Package Degradation 1-52
1.5.2.3 Effect of Percolation Flux Distribution 1-53
1.5.2.4 Conclusion 1-53
1.5.3 Sensitivity to Fuel Cladding Model 1-54
1.6 Waterbome Chemically Toxic Material Impacts 1-55
1.6.1 Chromium 1-55
1.6.1.1 RIP Model Adaptations for Chromium Modeling 1-55
1.6.1.2 Results for the Proposed Action 1-57
1.6.1.3 Results for Inventory Modules 1 and 2 1-59
1.6.2 Molybdenum 1-60
1.6.3 Uranium 1-60
1.6.3.1 RIP Model Adaptations for Elemental Uranium Modeling 1-60
1.6.3.2 Results for the Proposed Action 1-61
1.6.4 Results for Inventory Modules 1 and 2 1-62
1.7 Atmospheric Radioactive Material Impacts 1-62
1.7.1 Carbon-14 Releases to the Atmosphere 1-63
1.7.2 Atmosphere Consequences to the Local Population 1-64
1.7.3 Sensitivity to the Fraction of Early-Failed Cladding 1-66
References I-IU
I-iv
Environmental Consequences of Long-Term Repository Performance
LIST OF TABLES
Table Page
I-l Performance assessment model radionuclide inventory for commercial spent
nuclear fuel 1-9
1-2 Performance assessment model radionuclide inventory for DOE spent nuclear
fuel ~ MO
1-3 High-level radioactive waste mass and volume summary I-IO
1-4 Comparison of high-level radioactive waste inventories I-l 1
1-5 Nuclides at the Hanford Site for which Appendix A presents values greater than
those in the Characteristics Database I-ll
1-6 Nuclides for which Appendix A presents values lower than those in the
Characteristics Database 1-12
1-7 Nuclides at the Idaho National Engineering and Environmental Laboratory for
which Appendix A presents values greater than those in the Characteristics
Database 1-12
1-8 Greater-Than-Class-C low-level waste volumes by source 1-13
1-9 Performance assessement model radionuclide inventory for Greater-Than-Class-
C and Special-Performance-Assessment-Required waste 1-14
I- 10 Inventory of chemical materials placed in the repository under the Proposed
Action 1-15
I-l 1 Source concentrations of waterbome chemically toxic materials for screening
purposes 1-16
1-12 EQ6-modeled concentrations in solution from reaction of J13 water with carbon
steel and Alloy-22 1-18
1-13 Concentrations of waterbome chemically toxic materials for screening purposes 1-19
1-14 Human health hazard indices for chemically toxic materials 1-20
1-15 Chromium content of waste packages for the Proposed Action 1-21
1-16 Chromium content of waste packages for Inventory Module 1 1-21
1-17 Chromium content of waste packages for Inventory Module 2 1-22
1-18 Modeled waste package interior chromium inventory for Proposed Action 1-22
1-19 Modeled corrosion-resistant material (Alloy-22) chromium inventory for
Proposed Action 1-23
1-20 Modeled waste package interior chromium inventory for Inventory Module 1 1-23
1-21 Modeled corrosion-resistant material (Alloy-22) chromium inventory for
Inventory Module 1 1-24
1-22 Additional corrosion-resistant material (Alloy-22) chromium inventory for
Inventory Module 2 in excess of inventory for Module 1 1-24
1-23 Total elemental uranium inventory for Proposed Action and Inventory Modules 1
and 2 1-25
1-24 Total carbon- 14 inventory 1-25
1-25 Estimates of repository emplacement area 1-26
1-26 Waste package spacing for the Proposed Action inventory 1-29
1-27 Waste package spacing for Inventory Modules 1 and 2 1-29
1-28 Areas of submodels (stratigraphic columns) used in thermal-hydrologic
calculations 1-30
1-29 Uncertainty/variability splitting sets for corrosion rate of corrosion-resistant
material 1-36
1-30 Thermal-hydrologic and waste package degradation simulation matrix 1-38
1-31 Summary of fluxes from repository area to convolution stream tubes for
intermediate thermal load scenario with Proposed Action inventory 1-43
1-32 Summary of fluxes from repository area to convolution stream tubes for low
thermal load scenario with Proposed Action inventory 1-44
I-v
Environmental Consequences of Long-Term Repository Performance
Table Page
1-33 Summary of fluxes from repository area to convolution stream tubes for high
thermal load scenario with Inventory Modules 1 and 2 1-44
1-34 Summary of fluxes from repository area to convolution stream tubes for
intermediate thermal load scenario with Inventory Modules I and 2 1-44
1-35 Summary of fluxes from repository area to convolution stream tubes for low
thermal load scenario with Inventory Modules 1 and 2 1-44
1-36 Dilution factors for three thermal load scenarios and four exposure locations 1-47
1-37 Stochastic parameters for saturated zone flow and transport 1-48
1-38 Comparison of consequences for a maximally exposed individual from
groundwater releases of radionuclides using different fuel rod cladding models
under the high thermal load scenario 1-54
1-39 Chromium groundwater concentrations at 5 kilometers under Proposed Action
inventory using the high thermal load scenario and a two-stage RIP model 1-57
1-40 Peak chromium groundwater concentration under the Proposed Action inventory 1-58
1-41 Peak chromium groundwater concentration for 10,000 years after closure under
Inventory Module 1 1-59
1-42 Peak chromium groundwater concentration due only to Greater-than-Class-C and
Special-Performance-Assessment-Required wastes for 10,000 years after closure
under Inventory Module 2 1-59
1-43 Population by sector and distance from Yucca Mountain used to calculate
regional airborne consequences 1-64
1-44 Meteorologic joint frequency data used for Yucca Mountain atmospheric releases 1-65
I-vi
Environmental Consequences of Long-Term Repository Performance
LIST OF FIGURES
Figure Page
I-l Total system performance assessment model 1-67
1-2 Layout for Proposed Action inventory for high thermal load scenario 1-68
1-3 Layout for Inventory Modules 1 and 2 for high thermal load scenario 1-69
1-4 Layout for Proposed Action inventory for intermediate thermal load scenario 1-70
1-5 Layout for Inventory Modules 1 and 2 for intermediate thermal load scenario 1-71
1-6 Layout for Proposed Action inventory for low thermal load scenario 1-72
1-7 Layout for Inventory Modules 1 and 2 for low thermal load scenario 1-73
1-8 Relationship between the early performance assessment design and emplacement
block layout considered in this EIS performance assessment analysis 1-74
1-9 Development of thermal load scale factors on the basis of two-dimensional and
one-dimensional model comparisons using time history of temperature, liquid
saturation, and air mass fraction 1-75
I-IO Partition of repository area between center and edge regions 1-76
I-ll Temperature and relative humidity histories for all waste packages for high
thermal load scenario. Proposed Action inventory, and long-term average climate 1-77
1-12 Temperature and relative humidity histories for all waste packages, low thermal
load scenario. Proposed Action inventory, and long-term average climate 1-78
1-13 Temperature and relative humidity histories for the 21 pressurized-water-reactor
average waste packages, long-term average climate scenario, showing sensitivity
to waste inventory 1-79
I- 14 Temperature and relative humidity histories for the 21 pressurized-water-reactor
average waste packages, high thermal load scenario. Proposed Action inventory,
long-term average climate scenario, comparing the center and edge scenarios 1-80
1-15 WAPDEG input temperature and relative humidity histories for all thermal loads
with Proposed Action inventory 1-81
1-16 WAPDEG input temperature and relative humidity histories for all thermal loads
with Inventory Modules 1 and 2 1-82
1-17 Time to first breach of the corrosion-allowance material for low thermal load
scenario. Proposed Action inventory, all three stratigraphic columns, always-
dripping waste packages 1-83
1-18 Time to first breach of the corrosion-resistant material for low thermal load
scenario. Proposed Action inventory, all three stratigraphic columns, always-
dripping waste packages, and all nine uncertainty/variability splitting sets 1-83
1-19 Average number of patches failed per waste package as a function of time for
low thermal load scenario. Proposed Action inventory, all three stratigraphic
columns, always-dripping waste packages, and all nine uncertainty/variability
spHtting sets 1-84
1-20 Time to first breach of the corrosion-allowance material for high thermal load
scenario. Inventory Modules 1 and 2, center and edge regions for both
stratigraphic columns, always-dripping waste packages 1-84
1-2 1 Time to first breach of the corrosion-resistant material for high thermal load
scenario. Inventory Modules 1 and 2, center and edge regions for both
stratigraphic columns, always-dripping waste packages, and all nine
uncertainty/variability splitting sets 1-85
1-22 Average number of patches failed per package as a function of time for high
thermal load scenario. Inventory Modules 1 and 2, center and edge regions for
both stratigraphic columns, always-dripping waste packages, and all nine
uncertainty/variability splitting sets 1-85
I-vii
Environmental Consequences of Long-Term Repository Performance
Figure Page
1-23 Time to first breach of the corrosion-allowance material for all thermal loads and
inventories, all regions, always-dripping waste packages, uncertainty/variability
splitting set 5 1-86
1-24 Time to first breach of the corrosion-resistant material for all thermal loads and
inventories, all regions, always-dripping waste packages, uncertainty/variability
splitting set 5 1-86
1-25 Average number of patches failed per waste package as a function of time for all
thermal loads and inventories, all regions, always-dripping waste packages,
uncertainty /variability splitting set 9 1-87
1-26 Average number of patches failed per waste package as a function of time for all
thermal loads and inventories, all regions, always-dripping waste packages,
uncertainty/variability splitting set 5 1-87
1-27 Regions for performance assessment modeling. Option 1 , high thermal load
scenario. Proposed Action inventory 1-88
1-28 Regions for performance assessment modeling. Option 2, intermediate thermal
load scenario. Proposed Action inventory 1-89
1-29 Repository block areas for performance assessment modeling. Option 3, low
thermal load scenario with Inventory Module 1 , and intermediate thermal load
scenario with Inventory Module 1 cases 1-90
1-30 Regions for performance assessment modeling. Option 4, high thermal load
scenario. Proposed Action inventory 1-91
1-31 Regions for performance assessment modeling. Option 5, intermediate thermal
load scenario. Inventory Module 1 1-92
1-32 Repository block areas for performance assessment modeling. Option 6, low
thermal load scenario, Inventory Module 1 1-93
1-33 Capture regions for high and intermediate thermal load scenarios with Proposed
Action inventory 1-94
1-34 Capture regions for low thermal load scenario with the Proposed Action
Inventory and low and intermediate thermal load scenarios with Inventory
Modules 1 and 2 1-95
1-35 Capture regions for high thermal load scenario with Inventory Modules 1 and 2 1-96
1-36 Biosphere modeling components, including ingestion of contaminated food and
water, inhalation of contaminated air, and exposure to direct external radiation 1-97
1-37 Complementary cumulative distribution function of peak maximally exposed
individual radiological dose rates during 10,000 and 1 million years following
closure for high thermal load scenario with Proposed Action inventory 1-98
1-38 Complementary cumulative distribution function of peak maximally exposed
individual radiological dose rates during 10,000 and 1 million years following
closure for intermediate thermal load scenario with Proposed Action inventory 1-98
1-39 Complementary cumulative distribution function of peak maximally exposed
individual radiological dose rates during 10,000 and 1 million years following
closure for low thermal load scenario with Proposed Action inventory 1-99
1-40 Complementary cumulative distribution function of peak maximally exposed
individual radiological dose rates during 10,000 and 1 million years following
closure for high thermal load scenario with Inventory Module 1 1-99
1-4 1 Complementary cumulative distribution function of peak maximally exposed
individual radiological dose rates during 10,000 and 1 million years following
closure for intermediate thermal load scenario with Inventory Module 1 1-lOO
1-42 Complementary cumulative distribution function of peak maximally exposed
individual radiological dose rates during 10,000 and 1 million years following
closure for low thermal load scenario with Inventory Module I I-lOO
I-viii
Environmental Consequences of Long-Term Repository Performance
Figure Page
1-43 Comparison of low and intermediate thermal load scenarios total radiological
dose histories for the Proposed Action inventory 20 kilometers from the
repository I-lOl
I^ Waste package failure curves for low and intermediate thermal load scenarios I-lOl
1-45 Average percolation flux for repository blocks < 1-102
1-46 Neptunium-237 release rate at the water table for fixed long-term average climate
for low thermal load scenario during the first 1 million years following repository
closure I- 103
1-47 Neptunium-237 release rate at the water table for fixed long-term average climate
for intermediate thermal load scenario during the first 1 million years following
repository closure 1-103
1-48 Neptunium-237 release rate at the end of the saturated zone for fixed long-term
average climate for low thermal load scenario during the first I million years
following repository closure 1-104
1-49 Neptunium-237 release rate at the end of the saturated zone for fixed long-term
average climate for intermediate thermal load scenario during the fu^st 1 million
years following repository closure 1-104
1-50 Complementary cumulative distribution function of radiological doses with and
without cladding for a maximally exposed individual at 20 kilometers under the
Proposed Action 10,000 and 1 million years after repository closure 1-105
1-5 1 Average fractional release rate of corrosion-resistant material (Alloy -22) for
continually dripping and nondripping conditions computed from WAPDEG
modeling results for 400 simulated waste packages 1-105
1-52 Complementary cumulative distribution function of mean peak groundwater
concentrations of chromium during 10,000 years following closure under high
thermal load scenario with Proposed Action inventory 1-106
1-53 Complementary cumulative distribution function of mean peak groundwater
concentrations of chromium during 10,000 years following closure under
intermediate thermal load scenario with Proposed Action inventory 1-106
1-54 Complementary cumulative distribution function of mean peak groundwater
concentration of chromium during 10,000 years following closure under low
thermal load scenario with Proposed Action inventory 1-107
1-55 Complementary cumulative distribution function of mean peak groundwater
concentration of chromium during 10,000 years following closure under high
thermal load scenario with Inventory Module 1 1-107
1-56 Complementary cumulative distribution function of mean peak groundwater
concentration of chromium during 10,000 years following closure under
intermediate thermal load scenario with Inventory Module 1 1-108
1-57 Complementary cumulative distribution function of mean peak groundwater
concentration of chromium during 10,000 years following closure under low
thermal load scenario with Inventory Module 1 1-108
1-58 Complementary cumulative distribution function of mean peak groundwater
concentration of elemental uranium in water at 5 kilometers during 10,000 years
following closure under high thermal load scenario with Proposed Action
inventory 1-109
1-59 Fraction (patch area) of cladding that would fail using a zirconium-alloy
corrosion rate equal to 1.0 percent of that of Alloy-22 1-109
1-60 Release rate of carbon-14 from the repository to the ground surface I-UO
I-ix
Environmental Consequences of Long-Term Repository Performance
APPENDIX I. ENVIRONMENTAL CONSEQUENCES OF LONG-TERM
REPOSITORY PERFORMANCE
This appendix provides detailed supporting information on the calculation of the environmental
consequences of long-term (postclosure, up to 1 million years) repository performance. Chapter 5
summarizes these consequences for the Proposed Action, and Section 8.3 summarizes the cumulative
impacts of Inventory Modules 1 and 2.
Section 1. 1 introduces the bases for long-term performance assessment calculations. Section 1.2 provides
an overview of the use of computational models developed for the Total System Performance
Assessment - Viability Assessment used for this environmental impact statement (EIS). Section 1.3
identifies and quantifies the inventory of waste constituents of concern for long-term performance
assessment. Section 1.4 details the modeling extensions to the Viability Assessment base case (high
thermal load scenario with the Proposed Action inventory) developed to estimate potential impacts for
other thermal load scenarios and expanded inventories. Section 1.5 provides detailed results for
waterbome radioactive material impacts, while Section 1.6 provides the same for waterbome chemically
toxic material impacts. Section 1.7 describes atmospheric radioactive material impacts. To aid
readability, all the figures have been placed at the end of the appendix.
1.1 Long-Term Repository Performance Assessment Calculations
HOW ARE THE VIABILITY ASSESSMENT
AND THIS EIS PERFORMANCE
ASSESSMENT RELATED?
The long-term performance assessment for this
EIS builds incrementally on the Viability
Assessment (DOE 1998a, Volume 3, all; TRW
1998a,b,c,d,e,f,g,h,i,j,k. all).
This appendix reports only those aspects of the
EIS long-term performance assessment that are
incremental over the Viability Assessment. Only
those parts of the analysis unique to the EIS are
reported here, and the text refers to the
appropriate Viability Assessment documents for
information on the bases of the analyses.
This EIS analysis of postclosure impacts used
and extended the modeling work done for the
Total System Performance Assessment -
Viability Assessment, as reported in the U.S.
Department of Energy's (DOE's) Viability
Assessment of A Repository at Yucca Mountain,
Volume 3 (DOE 1998a, Volume 3, all) and in
the Total System Performance Assessment -
Viability Assessment (TSPA-VA) Analyses
Technical Basis Document (TRW
1998a,b,c,d,e,f,g,h,i,j,k, all). The Proposed
Action inventory under the high thermal load
scenario is identical to the Viability Assessment
base case, except that the Viability Assessment
only considered 20 kilometers (12 miles) from
the repository, while the EIS considers impacts
of radiological dose to maximally exposed
individuals through the groundwater pathway at
5, 20, 30, and 80 kilometers (3, 12, 19, and 50 miles) from the repository. The EIS analysis used a
repository integrated program computer model (Colder 1998, all) that DOE used for the total-system
model to calculate radiological doses through the groundwater pathway. This performance assessment
model and supporting Viability Assessment process models were extended to predict waterbome
chemically toxic material impacts. Additional calculations provided estimates of atmospheric
radiological doses to local and global populations.
The process of performing performance assessment analyses for this EIS required several steps. The EIS
analysis was designed to incorporate the Total System Performance Assessment - Viability Assessment
model of the base case repository configuration. Additional modeling (described in this appendix) was
performed to evaluate the impacts of alternative thermal load scenarios and expanded waste inventories.
The performance assessment model used for the Viability Assessment was expanded to accommodate
calculations of the radiological dose to people at distances other than those used in the Viability
I-l
Environmental Consequences of Long-Term Repository Performance
Assessment. Other adaptations to the model were made to calculate impacts from nonradiological
materials not considered in the Viability Assessment.
The performance assessment model simulates the transport of radionuclides away from the repository into
the unsaturated zone, through the unsaturated zone, and ultimately through the saturated zone to the
accessible environment. Performance assessment analyses depend greatly on the underlying process
models necessary to provide thermal-hydrologic conditions, near-field geochemical conditions,
unsaturated zone flow fields, and saturated zone flow fields as a function of time. Using these underlying
process models involves multiple steps that must be performed sequentially before performance
assessment modeling can begin.
Figure I-l shows the general flow of information between data sources, process models, and the total
system performance assessment model. (Note: Figures are on pages 1-67 to I- 110.) Several computer
models are identified in Figure I-l; these models are introduced in Section 1.2. The general purpose of
each of these computer models is described below its name in the figure. For example, TOUGH-2 is used
for the mountain-scale thermohydrology model and the drift-scale and mountain-scale unsaturated zone
flow model. The dashed box in the figure encompasses those portions of the performance assessment
model that are modeled within the repository integration program. Other functions are run externally as
"process models" to provide information to the repository integration program model. The ultimate result
sought from performance assessment modeling is a characterization of radiological dose to humans with
respect to time, which is depicted as the "Final Performance Measure" in the figure (the depiction is for
illustrative purposes only).
1.2 Total System Performance Assessment Methods and Models
DOE conducted analyses for this EIS to evaluate potential long-term impacts to human health from the
release of radioactive materials from the Yucca Mountain Repository. The analyses were conducted in
parallel with, but distinct from, the Total System Performance Assessment calculations for the Viability
Assessment (DOE 1998a, Volume 3, all). The methodologies and assumptions are detailed in the Total
System Performance Assessment - Viability Assessment Technical Bases Document (TRW
1998a,b,c,d,e,f,g,h,i,j,k, all). Extensions of the Viability Assessment analyses to meet distinct EIS
requirements were made using the same overall methodology.
The Total System Performance Assessment is a comprehensive systems analysis in which models of
appropriate levels of complexity represent all important features, events, and processes to predict the
behavior of the system being analyzed and to compare this behavior to specified performance standards.
In the case of the potential Yucca Mountain Repository system, a Total System Performance Assessment
must capture all of the important components of both the engineered and the natural barriers. In addition,
the Yucca Mountain Total System Performance Assessment must evaluate the overall uncertainty in the
prediction of waste containment and isolation, and the risks caused by the uncertainty in the individual
component models and corresponding parameters.
The components of the Yucca Mountain Repository system include five major elements that the Total
System Performance Assessment must evaluate:
• The natural environment unperturbed by the presence of underground openings or emplaced wastes
• Perturbations to the natural system caused by construction of the underground facilities and waste
emplacement
• The long-term degradation of the engineered components designed to contain the radioactive wastes
1-2
Environmental Consequences of Long-Term Repository Performance
• The release of the radionuclides from the engineered containment system
• The migration of these radionuclides through the engineered and natural barriers to the biosphere and
their potential uptake by people, leading to a radiation dose consequence
The processes that operate within these five elements are interrelated. To model the complexity of the
system efficiently, however, the following distinct process models were used in Total System
Performance Assessment - Viability Assessment and in performance assessment calculations for this EIS:
• The unsaturated-zone flow was modeled directly with a three-dimensional, site-scale, unsaturated
zone flow model, using the TOUGH2 program (Pruess 1991, all). Total System Performance
Assessment calculations modeled climate change by assuming a series of step changes in climatic
boundary conditions.
Drift-scale unsaturated zone thermal-
hydrology was modeled with the NUFT
program (Nitao 1998, all) in three
dimensions using a model domain that
contains discrete waste packages and
extends vertically from the water table to
the ground surface.
Waste package degradation was modeled
using the WAPDEG program (TRW
19981, all), which includes both
individual package variability and
package-to-package variability.
Waste-form and cladding degradation
was modeled in the repository integration
program model using empirical
degradation-rate formulas developed
CLIMATE CHANGE
The EIS performance assessment considered
three climate scenarios: (1) a present-day climate,
(2) a long-term average climate (wetter than the
present-day climate) scenario, and (3) a scenario
in which st/pen^/uv/a/ conditions (much wetter than
the present-day climate) are added at a short-
duration fixed interval on a periodic basis 100,000
years after waste emplacement. The climate
changes are step changes for the duration of the
climate periods, and the lengths of the sequences
are 10,000 years for the present-day dry climate
and the superpluvial climate, and 90,000 years for
the long-term average climate (DOE 1998a,
Volume 3, Section 5.1.1, page 5-1).
from available data. The model analyses used for the Total System Performance Assessment -
Viability Assessment and for this EIS included representation of the protective benefits of fuel
cladding for commercial spent nuclear fuel. The cladding failure model is described in detail in DOE
(1998a, Volume 3, Section 3.5.2, pages 3-100 to 3-103).
• Engineered barrier-system transport was modeled in the repository integration program model
(Colder 1998, all), using the program's cells algorithm. The transport modeling was based on an
idealized representation consisting of a linked series of equilibrium batch reactors, including the
waste form, waste package, corrosion products, and invert, and radionuclide transport through these
reactors (TRW 1998e, all).
• Unsaturated zone radionuclide transport was modeled directly with a three-dimensional site-scale
unsaturated zone-transport model using the FEHM model (Zyvoloski et al. 1995, all).
• Saturated zone flow and transport were modeled using a convolution method, in which the three-
dimensional, site-scale, saturated zone, flow-and-transport FEHM model (Zyvoloski et al. 1995, all;
TRW 1998g, all) was used to generate a library of solutions for translating time -varying mass inputs
to the saturated zone into water concentrations at exposure locations downgradient.
1-3
Environmental Consequences of Long-Term Repository Performance
The biosphere was modeled using biosphere dose-conversion factors that convert saturated zone
radionuclide concentrations to total radiological dose to an individual. The biosphere dose-
conversion factors were developed using the GENII-S program (Leigh et al. 1993, all). The total
radiological doses would be the final product of the Total System Performance Assessment
calculations.
The performance assessment calculations for both the Total System Performance Assessment - Viability
Assessment and this EIS were performed within a probabilistic framework combining the most likely
ranges of behavior for the various component models, processes, and related parameters. This appendix
presents the results in three main forms: (1) as probability distributions (for example, complementary
cumulative distribution functions) for peak radiological dose to a maximally exposed individual during
the 10,000 and 1 million years following repository closure; (2) as time histories of peak radiological dose
to a maximally exposed individual over 10,000 and 1 million years following repository closure; and
(3) in the case of this EIS only, as peak population radiological dose during 10,000 years for the local
population using contaminated groundwater. For maximally exposed individuals, the Viability
Assessment considered only a person 20 kilometers (12 miles) downgradient of the repository, while this
EIS considers individuals 5, 20, 30, and 80 kilometers (3, 12, 19, and 50 miles) downgradient from the
repository.
As noted above, the repository
integration program model
implements some of the individual
process models directly, while
other process models run outside
the repository integration program
model to produce abstractions in
the form of data tables, response
surfaces, or unit-response
functions. The repository
integration program model
provides a framework for
incorporating these abstractions,
integrating them with other
subsystem models. This is done in
a Monte Carlo simulation-based
methodology to create multiple
random combinations of the likely
ranges of the parameter values
related to the process models.
Probabilistic performance of the
entire waste-disposal system is
computed in terms of radiological
dose to individuals at selected
distances from the repository.
The EIS performance assessment
methodology draws on the
extensive analyses performed in
support of the Total System
Performance Assessment -
Viability Assessment. Most of the
process models (and their
THE COMPLEMENTARY CUMULATIVE
DISTRIBUTION FUNCTION
Example application for individual radiological dose
The value of many variables such as individual radiological
dose in the performance assessment models cannot be
known precisely, but they can be described in a statistical
sense. One of the statistical descriptions used is a
complementary cumulative distribution function. The function
for individual radiological dose is a curve that represents the
probability of exceeding various levels of radiological dose.
Although the complementary cumulative distribution function
is a curve, one can make probability statements for points on
the curve. For example, the stylized function for total
radiological dose to an individual shown here indicates that
there is a probability of 1 that radiological dose exceeds 0
millirem per year, a probability of 0.6 that radiological dose
exceeds 10 millirem per year, a probability of 0.1 that
radiological dose exceeds 20 millirem per year, and a
probability of 0 that radiological dose exceeds 39 millirem per
year.
Stylized Complementary Cumulative Distribution
Function of Individual Dose
J? c
3 O
11
O 3
3 o
§1
E €
E ^
o
O
1 ■
0.9-
^y
0.8-
\
0.7 -
\^^
0.6 -
^■^^^
0.5 -
^v
0.4-
^^
0.3-
^^
0.2 -
^^
0.1 -
0-
10
15
20 25 30 35 40
Dose (millirem per year)
1-4
Environmental Consequences of Long-Term Repository Performance
ABSTRACTION
Abstraction is the distillation of the essential
components of a process model into a
suitable form for use in a total system
performance assessment. The distillation
must retain the basic intrinsic form of the
process model but does not usually require
its original complexity. Model abstraction is
usually necessary to maximize the use of
limited computational resources while
allowing a sufficient range of sensitivity and
uncertainty analyses (DOE 1998a, Volume 3,
pageA-1).
MONTE CARLO METHOD:
UNCERTAINTY
An analytical method that uses random
sampling of parameter values available for
input into numerical models as a means of
approximating the uncertainty in the process
being modeled. A Monte Carlo simulation
comprises many individual runs of the
complete calculation using different values
for the parameters of interest as sampled
from a probability distribution. A different
final outcome for each individual calculation
and each individual run of the calculation is
called a realization (DOE 1998a, Volume 3,
page A-48).
j abstractions) developed for the Viability Assessment were used directly in the analyses described in this
I appendix. Only components that were modified to account for the additional analyses considered in this
EIS (but not the Viability Assessment) are described in this appendix.
i
1.3 Inventory
j The analyses of long-term performance considered the following waste categories for radioactive
! materials:
• Commercial spent nuclear fuel comprised of both conventional enriched uranium fuel and mixed-
oxide fuel using treated surplus fissile material that was reprocessed (consisting primarily of
plutonium)
• DOE spent nuclear fuel
• High-level radioactive waste (some of which contains immobilized surplus weapons-usable
plutonium)
• Greater-Than-Class-C waste and Special-Performance-Assessment-Required waste
The analysis assumed the waste would be in dual-shell waste packages. The outer shell would be
comprised of corrosion-allowance material (carbon steel) with an inner shell of corrosion-resistance
material (Alloy-22, a nickel-chromium alloy) (DOE 1998a, Volume 3, Figure 3-40, page 3-74). As
described in TRW (1997a, Section 2.6), it was assumed that the waste packages would contain fuel
assemblies from boiling-water reactors or pressurized-water reactors, naval ship or submarine reactors,
DOE research reactors, foreign research reactors, or vitrified high-level radioactive waste in canisters.
In addition, surplus plutonium not suitable for use in mixed-oxide fuel would be immobilized into
6.7-centimeter (2.6-inch)-diameter ceramic disks that would be packed in cylindrical cans, each
containing approximately I.O kilogram (2.2 pounds) of plutonium (see Appendix A). Twenty-eight of
these cans would be placed in a high-level radioactive waste canister and would occupy about 12 percent
of the volume of the canister. The remainder of each canister would be filled with vitrified high-level
radioactive waste. The plutonium encased in the high-level radioactive waste glass would then be
incorporated in standard waste packages. This analysis assumed that the high-level radioactive waste
would be in five-pack waste packages, each containing five high-level radioactive waste canisters and
disposed of with or without a canister of DOE spent nuclear fuel. The inventory used for this EIS
1-5
Environmental Consequences of Long-Term Repository Performance
assessment was the same as that used in the Viability Assessment (TRW 1998m, all), which also
considered more detailed sensitivity studies concerned with ceramic waste forms, alternative waste
package configurations, individual fuel assembly configurations, and mixed waste forms (DOE 1998a,
Volume 3, Section 5.5).
Thirty-nine radionuclides were included in the initial estimates of total inventories using the 0RIGEN2
program (Croff 1980, all). In the Viability Assessment and the EIS performance assessment model, the
list of 39 radionuclides was reduced to nine, based on the screening criteria discussed in this section and
observing the nuclides that contributed most to total radiological dose as calculated in the performance
assessment models. These nine radionuclides are carbon-14, iodine-129, neptunium-237, protactinium-
231, plutonium-239, plutonium-242, selenium-79, technetium-99, and uranium-234.
This section discusses the inventories of waterbome radioactive materials used to model impacts and of
some nonradioactive, chemically toxic waterbome materials used in the repository environment that could
present health hazards. This section also discusses the inventory of atmospheric radioactive materials.
1.3.1 WATERBORNE RADIOACTIVE MATERIALS
There would be more than 200 radionuclides in the materials to be placed in the repository (see
Appendix A). Because some of the radionuclides have a small inventory and some have short half-lives,
this analysis did not need to consider all of these radionuclides when estimating long-term repository
performance. Therefore, a screening analysis was performed to choose a subset of these radionuclides for
further analysis.
1.3.1 .1 Reduction of the List of Radionuclides for Performance Assessment Modeling
This evaluation of postclosure performance reduced the number of radionuclides considered by
eliminating any radionuclides that:
• Have short half-lives and are not decay products of long-lived radionuclides
• Have high chemical sorption such that long travel times to a human exposure location would result in
extremely low concentrations due to radioactive decay (unless the radionuclide has a large inventory
and the potential for colloidal transport)
• Have low biosphere dose-conversion factors
Any one or any combination of these factors would result in a diminished contribution by the
radionuclide to the total radiological dose; thus, eliminating that radionuclide from consideration would
not reduce estimates of radioactive material impacts. Based on these considerations and previous
performance analysis results (TRW 1995, all), DOE selected nine dominant radionuclides for analysis and
focused on those radionuclides that would have the most impact on human health, thereby enhancing
modeling efforts.
Two other factors were a part of the decision to reduce the list of radionuclides explicitly modeled in
performance assessment calculations. First, there was a need to reduce the number of radionuclides in
order to focus on only those radionuclides with the greatest impact on human health. Large
multidimensional flow-and-transport models such as the unsaturated zone and saturated zone particle-
tracking and transport models that are part of the repository integration program model require extensive
computer time (days or weeks). Hence, it was necessary to focus on those radionuclides that would have
the most impact on human health. The reduced list of radionuclides adequately characterized the impacts
without requiring an unnecessary computer modeling effort. Second, knowledge and experience gained
from earlier assessments (Wilson et al. 1994, all; TRW 1995, all), as well as the experience of other
1-6
Environmental Consequences of Long-Term Repository Performance
i
organizations (Wescott et al. 1995, all), were incorporated into the choice of radionuclides included for
analysis. To be included for the Total System Performance Assessment - Viability Assessment, a
radionuclide had to pass the elimination process performed under the basic criteria described above. It
also had to have an overall larger inventory than a similar radionuclide with similar performance
importance, or it had to have been identified as important in earlier studies.
The following is a discussion of the further rationale for the final selection of the specific radionuclides to
model.
Selected Radionuclides
• Carbon-14, technetium-99, and iodine-129. These radionuclides are highly soluble and exhibit
little or no chemical sorption. Technetium-99 and iodine-129 were major radiological dose
contributors in previous Total System Performance Assessments (Barnard et al. 1992, all; Wilson et
al. 1994, all). Carbon-14 and iodine-129 could be liberated from the waste packages as gases and
subsequently dissolved in water.
• Selenium-79, protactinium-231 , uranium-234, and neptunium-237. These radionuclides are
relatively soluble and have relatively low chemical sorption. Selenium-79 is the major radiological
dose contributor through a cow's liver pathway. Protactinium-231 has a relatively high sorption
coefficient, but because it is a decay product of uranium-235, it should be transported relatively
quickly and have a long residence time. Uranium-234 has a large inventory, is a decay product of
uranium-238, and has a high biosphere dose conversion factor. Neptunium-237 has been the most
important radionuclide in previous Total System Performance Assessments for exposure periods
between 20,000 and 1,000,000 years after repository closure.
• Plutonium-239 and plutonium-242. Although these plutonium isotopes are highly sorbing, they
were included on the list because of their large inventory and the possibility that they might migrate
by colloidal transport. These radionuclides would be among the most important radionuclides
involved in colloid-facilitated transport, if colloidal transport of plutonium were determined to be
important. Plutonium-242 was selected over plutonium-240 because of its longer half-life, thus
making it more likely to reach the accessible environment (especially via colloidal transport).
Radionuclides Not Selected
• Curium-246, curium-245, americium-241, americium-243, plutonium-240, uranium-238,
thorium-230, radium-226, lead-210, cesium- 137, cesium- 135, niobium-94, and nicl<el-59.
These radionuclides were among those selected by the U.S. Nuclear Regulatory Commission for its
Iterative Performance Assessment (Wescott et al. 1995, page 5-5). The Viability Assessment did not
include curium isotopes because of their similarity to plutonium. Americium isotopes were not
included directly because they have short half-lives, americium-243 was included in the plutonium-
239 inventory, and the activity of americium-241 was included in the neptunium-237 inventory.
Plutonium-240 was not selected because it is highly sorbing (although plutonium-242 was selected to
address colloidal transport). Uranium-238 was not selected because its decay product uranium-234
was chosen. Ingrowth of uranium-238 was compensated for by increasing the uranium-234
inventory. Thorium, radium, lead, cesium, niobium, and nickel were generally not included because
they are highly sorbing. In addition, lead-210, cesium-137, and radium-226 have relatively short
half-lives, while cesium-135, nickel-59, and niobium-94 have low inventories. For these reasons,
none of these radionuclides would contribute significantly to radiological dose (that is, including
these radionuclides in the calculations would not change the estimates of dose within the number of
significant figures reported for results).
Using only a subset of the radionuclides leads to potential underestimates of impacts to humans. The
modeling results reported in Chapters 5 and 8 show that in the first 10,000 years, the radiological dose is
1-7
Environmental Consequences of Long-Term Repository Performance
dominated by technetium-99, iodine-129, and carbon-14. These radionuclides all have relatively high
solubility and little chemical sorption. There are no other radionuclides with a meaningful inventory in
the proposed repository that share these characteristics. Thus, the error introduced by excluding other
radionuclides is very small in the first 10,000 years after repository closure.
The potential for underestimating impacts increases with time periods greater than 10,000 years after
repository closure. The possible error is largely due to the modeling of a few nuclides without modeling
the entire decay chain for the nuclide. Based on decay equilibrium calculations for the first 1,000,000
years after repository closure, the error from neglecting all other nuclides is about 5 percent of the total
radiological dose rate (DOE 1998a, Appendix C, page C6-2 and Figure C6-1).
The inventories for the categories of spent nuclear fuel and high-level radioactive waste described in the
following paragraphs include these nine radionuclides. The inventories of these radionuclides were used
in the performance assessment model to estimate the impacts to people.
The Viability Assessment and these EIS performance assessment calculations included only certain
nuclides of prominent decay chains. To account for the lack of ingrowth of decay products, modifications
were made to the nine radionuclide inventories for commercial spent nuclear fuel, DOE spent nuclear
fuel, and high-level radioactive waste. These modifications helped produce conservative estimates of the
activities of these nuclides (that is, estimates of the inventory would be equal to or greater than the real
inventory, so that any uncertainty would tend to overpredict impacts), which were then used by the
performance assessment model to determine impacts to individuals at specific exposure locations. Three
of the radionuclide inventories were modified as follows:
• The amount of protactinium-23 1 was entered in the repository integration program model as grams
per waste package of protactinium-23 1 rather than as curies per waste package, which allowed the
inventory of protactinium-23 1 to be modeled in secular equilibrium with its parent nuclide uranium-
235.
• The estimated activities of neptunium-237 and uranium-234 were increased by 58 percent and
13 percent, respectively. The increase in the activity of neptunium included the activity of the
precursors califomium-249, curium-245, plutonium-241, and americium-241 in the performance
assessment model. Neptunium-237 transports faster than the precursor radionuclides, so putting the
entire inventory in neptunium-237 would not underestimate the radiological dose. The increase of
activity in uranium-234 included the activity of precursors such as califomium-250, curium-246,
plutonium-242, americium-242, curium-242, uranium-238, and plutonium-238.
1.3.1.2 Radionuclide Inventory Used In the Performance Assessment Model
Radioactive material inventories were included in the performance assessment model for Total System
Performance Assessment calculations by the following waste categories: commercial spent nuclear fuel,
high-level radioactive waste, and DOE spent nuclear fuel. For each waste category, an abstracted waste
package was represented with an average radionuclide inventory for the nine radionuclides selected in the
screening analysis (see Section 1.3.1.1).
The quantity of abstracted packages was determined, in part, by averaging the characteristics of the
several different types of actual waste packages planned for each waste category and, in part, by demands
for a symmetrical, replicating arrangement of waste packages necessary for efficient thermal-hydrologic
modeling. Therefore, the quantity of abstracted packages in the performance assessment model differed
slightly from the actual quantity of waste packages identified in Appendix A and elsewhere. Other
inventory differences between the performance assessment model and Appendix A, and the associated
implications, are discussed in this section.
1-8
Environmental Consequences of Long-Term Repository Performance
ABSTRACTED WASTE PACKAGES
The number of waste packages used in the performance assessment simulations do not exactly
match the number of actual waste packages specified in TRW (1998n, all).
The performance assessment model uses three types of abstracted waste packages, representing
the averaged inventory of all the actual waste packages used for a particular waste category
(commercial spent nuclear fuel, DOE spent nuclear fuel, or high-level radioactive waste).
While the number of abstracted waste packages might vary from TRW (1998n, all), the total
radionuclide inventory (activity) represented by all of the abstracted waste packages collectively is
equivalent to the total inventory given in Appendix A, unless otherwise noted.
1.3.1.2.1 Commercial Spent Nuclear Fuel
The commercial spent nuclear fuel inventory is discussed in detail in Appendix A. The quantities and
activities were weighted according to the contributors and the expected waste package configurations.
Using these data, the analysis established an abstracted waste package commercial spent nuclear fuel
radionuclide inventory for the Total System Performance Assessment - Viability Assessment and EIS
performance assessment modeling (TRW 1998m, page 5-10). Table I-l lists the radionuclide inventory
for commercial spent nuclear fuel used for both the EIS and Viability Assessment analyses.
Table I-l. Performance assessment model radionuclide
inventory (curies per waste package) for commercial spent
nuclear fuel.'
Nuclide Inventory
Carbon- 14 12
Iodine- 129 0.29
Neptunium-237 1 1
Protactinium-231'' 5.1
Plutonium-239 3,100
Plutonium-242 17
Selenium-79 3.7
Technetium-99 120
Uraniuni-234 21
a. Source: DOE (1998a, Volumes, page 3-96).
b. Protactinium-23 1 is listed in grams per package to facilitate
modeling as an equilibrium decay product of uranium-235. The
specific activity of protactinium-23 1 is 0.0000022 curies per gram.
1.3.1.2.2 DOE Spent Nuclear Fuel
The DOE spent nuclear fuel inventory is discussed in detail in Appendix A. Table 1-2 lists the abstracted
waste package radionuclide inventory for DOE spent nuclear fuel used for the Viability Assessment and
the EIS analyses for the Proposed Action.
1.3.1 .2.3 High-Level Radioactive Waste
High-level radioactive waste is the highly radioactive material resulting from the reprocessing of spent
nuclear fuel, and the inventory for its disposal is presented in Appendix A. The high-level radioactive
waste inventory assembled for Total System Performance Assessment - Viability Assessment and EIS
performance assessment modeling was derived from the inventories of high-level radioactive waste at the
Hanford Site, Savannah River Site, Idaho National Engineering and Environmental Laboratory, and West
1-9
Environmental Consequences of Long-Term Repository Performance
Table 1-2. Performance assessment model radionuclide
inventory (curies per waste package) for DOE spent
nuclear fuel/
Nuclide Inventory
Carbon-14 0.31
Iodine- 129 0.0057
Neptunium-237 0.15
Protactinium-231'' 0.66
Plutonium-239' 155
Plutonium-242 0.1 1
Selenium-79 0.089
Technetium-99 2.6
Uranium-234 054
a. Source: DOE (1998a, Volume 3, page 3-96).
b. Protactiniuni-23 1 is listed in grams per package to facilitate
modeling as an equilibrium decay product of uranium-235. The
specific activity of protactinium-23 1 is 0.0000022 curies per gram.
c. Inventory for plutonium-239 is correct; DOE (1998a, Volume 3,
page 3-96) contains a typographical error.
Valley Demonstration Project. This inventory was established from the National Low-Level Waste
Database and weighted for the expected contributions from the four principal high-level radioactive waste
sites listed above using quantities calculated in the Waste Quantity, Mix and Throughput Report (TRW
1997a, all). This inventory is listed in Table 1-3 for the nine modeled radionuclides.
Table 1-3. High-level radioactive waste mass and volume summary.
Parameter EIS analyses Appendix A
Mass (metric tons)
NA^
58,000
Volume (cubic meters)
18,000
21,000
Number of canisters
19,234
22,280
Waste packages (5-packs)
3,848
4,456"
a. NA = not applicable.
b. Derived from data presented in fi
mpendix A.
These data were included in the high-level radioactive waste inventory for the Viability Assessment base
case (TRW 1998o, all); long-term performance assessment analyses for this EIS used this same inventory.
Recent updates of the waste inventories from the DOE sites are in Appendix A. The most recent
estimates from these sites indicated a higher total volume of high-level radioactive waste but with an
overall lower activity. Appendix A provides a 1998 summary of the potential total mass, volume, and
number of canisters of high-level radioactive waste that would be available to the Yucca Mountain
Repository from the principal waste sites.
These performance assessment analyses did not use the most recent information reported in Appendix A,
because the more recent estimates of high-level radioactive waste activity were received too late for
inclusion in the Viability Assessment and EIS performance assessment calculations (see TRW I998f,
page 6-16). A sensitivity analysis of high-level radioactive waste was performed by comparing the high-
level radioactive waste inventory used in EIS analyses to the inventory in Appendix A. The results of the
analysis showed that the estimate of total radiological dose to maximally exposed individuals at
20 kilometers (12 miles) from the Yucca Mountain Repository, using the high-level radioactive waste
base case inventory for the Viability Assessment, led to higher amounts of radionuclides contributing to
radiological dose than those calculated using the revised data from Appendix A. Therefore, actual
impacts would be lower than estimated if the more recent information were used. Table 1-4 compares the
nine radionuclide inventories used in the Viability Assessment and EIS analyses with those used in the
Appendix A inventory. Note that the nine modeled radionuclides do not contribute equally to radiological
I-IO
Environmental Consequences of Long-Term Repository Performance
Table 1-4. Comparison of high-level radioactive waste inventories
(curies per package).
TSPA-VA inventory ^ Appendix A inventory
Nuclide (3,848 packages) (4,456 packages)
Carbon- 14
0
0.032
Iodine- 129
0.000042
0.0085
Neptunium-237
0.74
0.13
Protactinium-231''
0.036
0.82
Plutonium-239
24
68
Plutoniuni-242
0.02
0.014
Selenium-79
0.29
0.49
Technetium-99
30
13
Uranium-234
0.9
0.15
a. Source: TSPA-VA = (Total Systems Performance Assessment - Viability
Assessment); DOE (1998a, Volume 3, page 3-96).
b. Protactinium-23 1 is listed in grams per package to facilitate modeling as an
equilibrium decay product of uranium-235. The specific activity of
protactinium-23 1 is 0.0000022 curies per gram.
dose, so a comparison of the inventories in Table 1-4 can be misleading. For example, neptunium-237
typically contributes more than 90 percent of the dose in the 1 -million-year period, so the larger inventory
of neptunium-237 in the Total Systems Performance Assessment - Viability Assessment inventory
column is more important that the smaller inventory of other radionuclides relative to the Appendix A
inventory column. Similarly, iodine-129 and technetium-99 inventories contribute most of the dose in the
10,(X)0-year period, so difference in those inventories are most important in that case.
The source used for the Viability Assessment to establish the inventory of high-level radioactive waste
was the Characteristics Database (DOE 1992, all). Appendix A contains data submitted by the individual
sites in response to an EIS data call. The differences in the data from each source are listed below by site.
Discussion of differences is limited to the nine radionuclides modeled in the performance assessment
analyses.
Hanford Site
• The Characteristics Database (DOE 1992, all) assumes 1,650 kilograms (3,630 pounds) of glass per
canister.
• Appendix A reports the mass of glass per canister as 3,040 kilograms (6,700 pounds). Values in
Appendix A are generally higher than those presented in the Characteristics Database (DOE 1992,
all); these values are listed in Table 1-5. Nuclide values which are generally lower in Appendix A
than the Characteristics Database are presented in Table 1-6.
Table 1-5. Nuclides at the Hanford Site for
which Appendix A presents values greater than
those in the Characteristics Database.''
Nuclide Factor
Iodine-129
100
Protactinium-231
100,000
Plutonium-239
2.5
Selenium-79
8
Uranium-234
5
a. Source: DOE (1992, all).
I-ll
Environmental Consequences of Long-Term Repository Performance
Table 1-6. Nuclides for which Appendix A
presents values lower than those in the
Characteristics Database."
Nuclide Factor
Neptunium-237 100
Technetium-99 3
a. Source: DOE (1992, all).
Idaho National Environmental and Engineering Laboratory
• The Characteristics Database (DOE 1992, all) inventory numbers do not include the projected high-
level radioactive waste inventory from the Argonne National Laboratory-West ceramic and metal
waste matrices (approximately 102 canisters).
• Appendix A reported values for carbon-14 and iodine-129 (0.000083 and 0.017 curie per canister,
respectively), while the Characteristics Database (DOE 1992, all) reported no values for these
nuclides.
• The Characteristics Database (DOE 1992, all) reported 0.08 curie per canister for selenium-79;
however, no value is reported for use in Appendix A.
• For the other nuclides, the values reported in Appendix A are greater by a variety of factors, as listed
in Table 1-7.
Table 1-7. Nuclides at the Idaho National
Engineering and Environmental Laboratory
for which Appendix A presents values greater
than those in the Characteristics Database."
Nuclide Factor
Neptunium-237 270
Plutonium-239 2.25
Plutonium- 242 1.65
Technetium-99 3.7
Uranium-234 200,000
a. Source: DOE (1992, all).
Savannah River Site
• In general, the Appendix A values for the other nuclides are slightly smaller (generally less than
1 percent) than those presented in the Characteristics Database (DOE 1992, all). The uranium-234
value reported in Appendix A is 77 percent less; however, most of the other nuclides are within
1 percent of the values in the Characteristics Database.
West Valley Demonstration Project
• The Characteristics Database (DOE 1992, all) does not include data for carbon-14 or iodine-129;
Appendix A uses approximately 0.53 and 0.00081 curie per canister, respectively, for these nuclides.
• Neptunium-237, plutonium-239, plutonium-242, and protactinium-23 1 differ slightly in Appendix A
(by about 1 percent) due largely to the difference in reporting accuracy (Appendix A reports two
significant figures; the Characteristics Database reports three).
•
Uranium-234 is increased by about 15 percent in Appendix A.
Technetium-99 and selenium-79 are both higher in Appendix A by a factor of approximately 15.
1-12
Environmental Consequences of Long-Term Repository Performance
1.3.1.2.4 Greater-Than-Class-C and Special-Performance-Assessment-Required Wastes
Wastes with concentrations above Class-C limits (shown in 10 CFR Part 61.55, Tables 1 and 2 for long
and short half-life radionuclides, respectively) are called Greater-Than-Class-C low-level waste. These
wastes generally are not suitable for near-surface disposal. The Greater-Than-Class-C waste inventory is
discussed in detail in Appendix A.
DOE Special-Performance-Assessment-Required low-level radioactive waste could include production
reactor operating wastes, production and research reactor decommissioning wastes, non-fuel-bearing
components of naval reactors, sealed radioisotope sources that exceed Class-C limits for waste
classification, DOE isotope production related wastes, and research reactor fuel assembly hardware. The
Special-Performance-Assessment-Required waste inventory is discussed in detail in Appendix A.
The final disposition method for Greater-Than-Class-C and Special-Performance-Assessment-Required
low-level radioactive waste is not known. If these wastes were to be placed in a repository, they would be
placed in canisters before shipment. This appendix assumes the use of a canister similar to the naval
dual-purpose canister described in Section A.2.2.5.6.
Table 1-8 lists existing and projected volumes through 2055 for the three Greater-Than-Class-C waste
sources. DOE conservatively assumes 2055 because that year would include all Greater-Than-Class-C
low-level waste resulting from the decontamination and decommissioning of commercial nuclear reactors.
The projected volumes conservatively reflect the highest potential volume and activity expected based on
inventories, surveys, and industry production rates.
Table 1-8. Greater-Than-Class-C low-level waste volumes (cubic
meters)^ by source.^
Source
1993
2055
Nuclear electric utility
26
1,300
Sealed sources
40
240
Other
74
470
Totals
140
2,010
a. To convert cubic meters to cubic feet, multiply by 35.314.
b. Source: DOE (1994, Tables 6-1 and 6-3).
The data concerning the volumes and projections of Greater-Than-Class-C low-level waste are from
Appendix A-1 of the Greater-Than-Class-C Low-Level Radioactive Waste Characterization: Estimated
Volumes, Radionuclide Activities, and Other Characteristics (DOE 1994, all). This appendix provides
detailed radioactivity reports for such waste currently stored at nuclear utilities. Table 1-9 summarizes the
radioactivity data for the nine radionuclides modeled in performance assessment calculations, decayed to
2055.
1.3.2 WATERBORNE CHEMICALLY TOXIC MATERIALS
Waterbome chemically toxic materials that could present a human health risk would be present in
materials disposed of in the repository. The most abundant of these chemically toxic materials would be
nickel, chromium, and molybdenum, which would be used in the waste package, and uranium in the
disposed waste. Uranium is both a chemically toxic and radiological material. Screening studies were
conducted to determine which, if any, of these or other materials could be released in sufficient quantities
to have a meaningful impact on groundwater quality.
1-13
Environmental Consequences of Long-Term Repository Performance
Table 1-9. Performance assessment model radionuclide
inventory (curies per waste package) for Greater-Than-
Class-C and Special-Performance-Assessment-Required
waste/
Nuclide
Inventory
Carbon- 14
38
Iodine- 129
0.000000012
Neptunium-237
0.000000052
Protactinium-231''
0.0000015
Plutonium-239
48
Plutonium-242
0.0000040
Selenium-79
0.0000010
Technetium-99
2.6
Uranium-234
0.00000062
a. Source: TRW (1999a, Table 2.2-6, page 2-10).
b. Protactinium-23 1 is listed in grams per package to facilitate
modeling as an equilibrium decay product of uranium-235. The
specific activity of protactinium-23 1 is 0.0000022 curies per
gram.
1.3.2.1 Identification of Waterborne Chemically Toxic Materials
An inventory of chemical materials to be placed in the repository under the Proposed Action was
prepared. The inventories of the chemical components in the repository were combined into four groups:
• Materials outside the waste packages (concrete, copper bus bars, structural members, emplacement
tracks and supports, etc.)
• Carbon steel in the outer layer of the waste packages
• Alloy-22 in the inner layer of the waste packages
• Materials internal to the waste packages
These materials were organized into groups with similar release times for use in the screening study.
Table I-IO lists the chemical inventories. Plutonium is not listed in Table I-IO because, while it is a heavy
metal and therefore could have toxic effects, its radiological toxicity far exceeds its chemical toxicity
(DOE 1998b, Section 2.6.1) (see Section 1.5 for more information). Also, while there are radiological
limits set for exposure to plutonium, no chemical toxicity benchmarks have been developed. Therefore,
because of this lack of data to analyze chemical toxicity, plutonium was not analyzed for the chemical
screening.
1.3.2.2 Screening Criteria
Only those chemicals likely to be toxic to humans were carried forward in the screening study. Uranium
was an exception; it was carried forward due to its high inventory and also to serve as a check on the
screening study. Chemicals included in the substance list for the U.S. Environmental Protection
Agency's Integrated Risk Information System (EPA 1999, all) were evaluated further to determine a
concentration that would be found in drinking water in a well downgradient from the repository. The
chemicals on the Integrated Risk Information System substance list that would be in the repository are
barium, boron, cadmium, chromium, copper, lead, manganese, mercury, molybdenum, nickel, selenium,
uranium, vanadium, and zinc.
1-14
Environmental Consequences of Long-Term Repository Performance
Table I-IO. Inventory (kilograms/ of chemical materials placed in the repository under the Proposed
Action.
Inventory
High-level
Outside
radioactive
Element
package
Carbon steel
Alloy-22
Internal
waste
Totals
Aluminum
0
0
0
1,205,000
0
1,205,000
Barium
0
0
0
0
19,000
19,000
Boron
0
0
0
223,000
0
223,000
Cadmium
0
0
0
0
43,000
43,000
Carbon
286,000
796,000
8,000
5,000
0
1,096,000
Chromium
0
0
9,670,000
3,903,000
0
13,573,000
Cobalt
0
0
1,357,000
27,000
0
1,384,000
Copper
1,135,000
0
0
3,000
0
1,139,000
Iron
91,482,000
320,089,000
2,171,000
9,000
0
413,751,000
Lead
0
0
0
0
2,000
2,000
Magnesium
0
0
0
12,000
0
12,000
Manganese
234,000
3,007,000
271,000
2,000
0
3,514,000
Mercury
0
0
0
0
200
200
Molybdenum
0
0
5,934,000
302,000
0
6,236,000
Nickel
0
0
29,727,000
5,563,000
0
35,290,000
Phosphorus
37,000
114,000
11,000
0
0
161,000
Selenium
0
0
0
0
300
300
Silicon
361,000
943,000
43,000
7,000
0
1,354,000
Sulfur
46,000
114,000
11,000
0
0
170,000
Titanium
0
0
0
2,000
0
2,000
Tungsten
0
0
1,628,000
0
0
1,628,000
Uranium
0
0
0
70,000,000
0
70,000,000
Vanadium
0
0
190,000
0
0
190,000
Zinc
0
0
0
3,000
0
3,000
a. To convert kilograms to pounds,
, multiply by 2,2046.
1.3.2.3 Screening Application
The screening calculations for chemically toxic materials assume that groundwater would move through
the repository, dissolving and transporting the potentially chemically toxic materials. This analysis
treated the repository materials and the carbon-steel layer of the waste package as simultaneously
degrading in the groundwater. After the carbon-steel layer of the waste degraded, the Alloy-22
corrosion-resistant material would start degrading. Finally, once the waste package was breached, the
materials inside the waste packages would become available for dissolution and transport.
1.3.2.3.1 Solubility of Chemically Toxic Materials in the Repository
The release of chemically toxic materials to the accessible environment depends on the solubility of the
materials in water. Table I-l 1 lists the solubility values used for the screening study.
Maximum source concentrations for materials in the repository that are not a part of the waste package
materials were calculated as solubilities of an element in repository water. This calculation would
provide the maximum possible concentration of that element in water entering the unsaturated zone if it
dissolved at a sufficiently high rate. The solubilities were obtained by modeling with the EQ3 code
(Wolery 1992, all). The simulations were started with water from well J-13 near the Yucca Mountain site
(Harrar et al. 1990, all). EQ3 calculates chemical equilibrium of a system so that by making successive
runs with gradually increasing aqueous concentrations of an element, eventually a result will show the
saturation of a mineral in that element. That concentration at which the first mineral saturates is said to be
1-15
Environmental Consequences of Long-Term Repository Performance
SCREENING ANALYSIS
A screening analysis is a method applied to avoid unnecessary calculations and focus on potentially
large impacts.
The repository would contain many materials that could result in impacts to human health. However,
most of these materials would either not be present in large enough quantities or not dissolve readily
enough in water to pose a risk.
To evaluate the potential risk posed by so many materials, an analysis could either rigorously
evaluate every material at great cost, or could apply a screening analysis to identify those materials
with too little inventory or too little solubility to be of concern. The screening analysis applied for the
EIS was a simplified scoping calculation which resulted in a short list of materials that merited further
consideration. Any preliminary concentrations predicted under the simplified assumptions of the
screening analysis were treated as conservative estimates used only to determine if the material
should be rigorously modeled again using the performance assessment model. For those materials
that the screening analysis indicated must be evaluated further, more realistic concentrations and
impacts were computed with the performance assessment model and are reported in Sections 1.5
and 1.6.
Table I-ll. Source concentrations" (milligrams per liter) '' of waterbome chemically toxic materials for
screening purposes.
Element
Concentration
Aqueous species
Reference
Boron
6,400
B(OH)3aq
Chromium
300
Cr04"-
Copper
0.018
CuOH\ Cu(C03)aq
Cu'"
Manganese
4.40x10"
Mn^
Molybdenum
218
Mo04-
Nickel
1.00 X 10*
N-r
Uranium
0.6
U02(OH)2aq
Vanadium
4.8
V030H"",HV04""
Zinc
63
Zn++
Solubility in repository water by EQ3'^
simulation
EQ6 "* simulation of Alloy 22 corrosion
Solubility in repository water by EQ3 '^
simulation
EQ6'' simulation of Alloy 22 corrosion
EQ6 ^ simulation of Alloy 22 corrosion
EQ6 '' simulation of Alloy 22 corrosion
Derived from TRW (1997b), Figure C-3,
page C-8'
EQ6'' simulation of Alloy 22 corrosion
Solubility in repository water by EQ3 "^
simulation
Concentration at the point where the chemical enters unsaturated zone water, controlled by solubility or local chemistry of
dissolution and interaction with tuff. Note that these concentrations are not used for transpwrt modeling (which is discussed
in Section 1.6) but are used only for screening analysis purposes. Refer to Section 1.6 for groundwater concentrations of
chemically toxic materials that were selected for further consideration based on the screening analysis.
To convert milligrams per liter to pounds per cubic foot, multiply by 0.00000624.
EQ6 code. Version 7.2b (Wolery and Daveler 1992, all).
EQ3 code, Version 7.2b (Wolery 1992, all).
For ph=8 and Co2=10 atmospheric partial pressure.
the "solubility." For example, the solubility of copper (from the bus bars left in the tunnels) would be
obtained by increasing copper concentrations in successive runs of EQ3. At a concentration of 0.0181 1
milligram per liter, tenorite (CuO) would be saturated. This mineral would then be in equilibrium with
dissolved copper existing in approximately equal molar parts as CuOH*, Cu(C03)aq, and Cu^*. The
aqueous concentration was then reported in Table I-ll as a "solubility" of copper for the purposes of
screening the potentially toxic chemicals.
The largest quantities of potentially toxic materials come from the construction materials of the waste
packages themselves. The main source is the Alloy-22 material used in the corrosion-resistant layer. The
possible maximum concentrations of these materials (chromium, nickel, molybdenum, manganese, and
vanadium) were developed by examining the corrosion process. Corrosion was modeled in the EQ6 code
1-16
Environmental Consequences of Long-Term Repository Performance
(Wolery and Daveler 1992, all), starting with the same repository water as used in the solubility
calculations described above. In the corrosion step, EQ6 was run in the titration mode (that is, a confined
area in which essentially stagnant water reacts with iron from existing corrosion-allowance material
fragments and Anoy-22). Oxygen is fixed at atmospheric fugacity (which is analogous to partial pressure
adjusted for nonidealities). After a few hundred years, the chemistry of the resultant solution stays
relatively constant for a long period. Following that, ionic strength eventually exceeds limits for +EQ6.
The chemistry during this "flat period" was used as the resultant solution, which contained very high
quantities of dissolved chromium (as hexavalent chromium), nickel, and molybdenum, and small
dissolved quantities of manganese and vanadium. The reaction of this solution with tuff was then
modeled. The resultant solution showed that essentially all of the nickel and manganese were precipitated
and that the original dissolved concentrations of chromium, molybdenum, and vanadium remained.
Two types of geochemical analyses were performed. The first was an analysis of the solution
concentration obtained when J-13 water, adjusted for the presence of repository materials such as
concrete (that is, the same water chemistry used for other process modeling work supporting the Total
System Performance Assessment-Viability Assessment), reacts with a large mass of carbon steel and
Alloy-22 for an extended period. The second was an analysis of the reaction of the solution from the first
analysis with volcanic tuff. The resultant solution from the second analysis would represent a bounding
value for the source term solution at the floor of the emplacement drift.
At each step of the reaction progress in which the titration mode of EQ6 was used, a small quantity of
reactants (steel and Alloy-22) was added to the solution (starting as J-13 water). After each addition, the
increment of reactant dissolves and all product phases would reequilibrate with the aqueous solution.
After a long time, this process would produce a bounding concentration for the solution. This would be
the case if the water had a very long contact time with the metals and a very limited amount of water was
used.
The composition of J-13 water was taken from earlier studies (TRW 1997b, page A-5). The carbon
dioxide and oxygen levels are maintained at atmospheric conditions during the reaction. This process
promotes the formation of the chromate (Cr04..) ion, which represents the hexavalent (and most toxic)
state of chromium. The complete oxidation of chromium and the formation of chromate creates a very
low pH environment in the area immediately adjacent to the corrosion process. The result of a low pH
level in the presence of sufficient oxygen would be dissolved chromium existing in the hexavalent state.
Large amounts of soluble hexavalent molybdenum are also formed.
Once the corrosion solution left the waste package, it would quickly encounter rock material. The second
analysis evaluated the effect of rock on the solution. The analysis used the option for a "Fluid-Centered
Flow-Through Open System" in EQ6. In this type of simulation the solution is permitted to react with
solid materials (in this case, the tuff) for some specified interval (either time or reaction progress). The
solution is then moved away from the solid reaction products that would be created and allowed to react
with the same initial solids for a further interval. In this way, the model simulates reaction of the solution
as it percolates through a rock.
This analysis simulated the tuff rock with the elemental composition characteristics of volcanic tuff.
Earlier waste package criticality studies used this formulation for tuff reactants (TRW 1997c, page 17).
The resultant solution from the simulated reaction of J-13 water with carbon steel and Alloy-22 has a very
low pH and a high concentration of dissolved chromium, molybdenum, and nickel. The resulting pH 2.0
solution would have the elemental concentrations listed in the second column of Table 1-12. When the
solution from corrosion contacts the rock, it would be neutralized to a pH of 8. The availability of silica
in the rock would promote the formation of silicates, which would precipitate most of the nickel and
manganese but virtually none of the chromium, molybdenum, or vanadium. Some chromium would
change to Cr207 (still hexavalent and very soluble). The molybdenum would behave in a very similar
1-17
Environmental Consequences of Long-Term Repository Performance
Table 1-12. EQ6-modeled concentrations (milligrams per liter)" in solution from
reaction of J-13 water with carbon steel and Alloy-22.
Element After corrosion of Alloy-22 After reaction with tuff rock
Chromium 299 299
Manganese 32 4.40x10""
Molybdenum 218 218
Nickel 750 9.9x10"'
Vanadium 4^8 4^8
a. To convert milligrams per hter to pounds per cubic foot, multiply by 0.00000624.
fashion and remain in solution as hexavalent species. The resultant solution would have the elemental
concentrations listed in the third column of Table 1-12.
The mechanism for mass loss of the Alloy-22 remains an issue at this time. There is no reliable evidence
to support or refute the idea that the chromium that is carried away from Alloy-22 is dissolved hexavalent
chromium. What is known fairly well is that trivalent chromium is the likely constituent (as Cr203) of the
passivation film and that it has a very low solubility. It is not known whether the film grows thick until it
sloughs off or if the film oxidizes in place so that it loses hexavalent chromium into solution. It is also
not known if the film would oxidize and dissolve if it did slough off. EQ6 simulates a process whereby
the trivalent chromium oxidizes to hexavalent chromium by reaction with O2. It is well known that if
chromium is in solution, the predominant species will be hexavalent chromium, especially in oxidizing
conditions. At the Eh for atmospheric oxygen, it is known that the ratio of hexavalent chromium to For
purposes of analysis, DOE assumes hexavalent chromium is mobilized as a dissolved constituent, and its
source term is represented by 0.22 times the bulk loss rate of Alloy-22. A parallel assumption has been
made about hexavalent molybdenum, which is also present in meaningful quantities in the results of the
corrosion simulation.
1.3.2.3.2 Well Concentration of Chemically Toxic Materials
After the materials would begin to be released from the repository, they would be transported through the
unsaturated zone to the saturated zone and on to the accessible environment. The screening study
assumed that the chemicals would flow to a well from which an individual received all of their drinking
water. Table 1-13 lists the concentrations for the chemically toxic materials.
The well concentrations listed in Table 1-13 were based on a series of simple calculations. First, the
release concentrations for each material were calculated. The release rate for the material in the carbon
steel is based on a degradation rate of 0.025 millimeter (0.001 inch) per year and a thickness of
100 millimeters (3.9 inches); thus, the annual fractional release rate for carbon steel is 0.00025. The
degradation rate for Alloy-22 is 0.000006 millimeter (0.00000024 inch) per year and the material
thickness is 20 millimeters (0.79 inch); the resulting annual fractional release rate is 0.0000003. The
internal materials were assumed to be released at the same rate as the carbon steel (a conservative
assumption). The release rate for the high-level radioactive waste was taken from earlier studies (TRW
1998f, Section 6.4). The annual fractional release rate for the high-level radioactive waste is 0.000054.
The well concentrations in Table 1-13 are very conservative concentration estimates that are not used
directly for impact estimates. Instead, they are used to screen potentially toxic chemicals for more
detailed analyses. These estimates were then compared to the Maximum Contaminant Levels for each
material, if available (40 CFR 141.2). Some of the estimated concentrations were orders of magnitude
below their respective Maximum Contaminant Levels. As a result of this screening study, barium,
copper, lead, mercury, and selenium were eliminated from further detailed analysis. All the other
chemically toxic materials, including boron, cadmium, chromium, manganese, molybdenum, nickel,
uranium, vanadium, and zinc, were carried forward for further detailed analysis (see Chapter 5,
Section 5.6.1).
1-18
Environmental Consequences of Long-Term Repository Performance
Table 1-13. Concentrations (milligrams per liter)" of waterbome chemically toxic materials for screening
K
purposes
Concentration
Release concentration
Well
Maximum
Non-
Carbon
contaminant
Element
limit
package
steel
AlIoy-22
Internal
HLW
Maximum concentration
level'
Barium
0.00412
0
0
0
0
0.99
0.00412
1.5x10'
2.0
Boron
6,400
0
0
0
50
0
52
1.9x10'
NA''
Cadmium
23
0
0
0
0
2.2
2.2
7.7x10'
0.005
Chromium
300
0
0
2.7
940
0
300
1.1
0.1
Copper
0.018
0.018
0
0
0
0
0.018
6.4x10'
1.3
Lead
NA
0
0
0
0
0.09
0.09
3.2x10''
0.015
Manganese
4.4x10"
4.4x10"
707
0.077
0.44
0
4.4x10"
1.6x10"
NA
Mercury
NA
0
0
0
0
0.01
0.01
3.6x10'
0.002
Molybdenum
218
0
0
2.07
71
0
71
2.5x10"'
NA
Nickel
1.0x10*
0
0
8.4
1,310
0
1.0x10*
3.5x10'
NA
Selenium
NA
0
0
0
0
0.014
0.014
4.9x10'
0.05
Uranium
0.0023
0
0
0
16,500
0
0.0023
8.2x10*
NA
Vanadium
4.8
0
0
0.054
0
0
0.054
1.9x10"
NA
Zinc
63
0
0
0
0.73
0
0.73
2.6x10'
NA
a.
b.
c.
d.
To convert grams per cubic meter to pounds per cubic foot, multiply by 0.00000624.
Note that these concentrations are not used for transport modeling (as discussed in Section 1.6), but only for screening analysis
purposes. Refer to Section 1.6 for groundwater concentrations of chemically toxic materials that were selected for further
consideration based on the screening analysis.
Maximum contaminant levels are specified in 40 CFR 141.2.
NA = not available (no Maximum Contaminant Level established by the U.S. Environmental Protection Agency for this element).
For the chemicals in the nonpackaged materials, the degradation was assumed to be limited by the
solubility of the chemical in water. The release concentration (in grams per cubic meter) was assumed to
be equal to the elemental solubility for those chemicals with a nonzero inventory in the nonpackaged
materials. For the remaining material categories, all part of the waste packages, the release concentration
was calculated based on the per-package inventory and the release rate from a waste package.
The per-package inventory (in grams for each material category) was calculated by dividing the total
inventory (in grams) of the material type by the total number of waste packages in the repository
(assumed to be 1 1,969). The release of material per cubic meter would be the fractional release rate
divided by the rate of water flow past a waste package, based on an average 20-millimeter (0.79-inch)
annual water flow rate through the repository. The release concentration is the per-package inventory in
grams multiplied by the release per cubic meter.
To estimate the concentration in a well, two steps were performed. First, the maximum release
concentration from the four material groups was selected. Then, two dilution factors were applied to the
maximum release concentration. An unsaturated zone dilution factor was calculated as the ratio of the
total cross-sectional area of all waste packages to the total surface area of the repository. Each of the
1 1,969 waste packages would have a cross-sectional area of 8.9 square meters (96 square feet), and the
assumed repository surface area would be about 3 square kilometers (740 acres). This calculation
resulted in an unsaturated zone dilution factor of 0.035. A dilution factor of 10 was applied to the
saturated zone so the dilution factor, when combined for the unsaturated and saturated zones, would be
0.0035.
1.3.2.3.3 Health Effects Screening for Chemically Toxic Materials
The potential for human health impacts was estimated using a hazard index. The hazard index was
determined by dividing the intake of a chemical by the oral reference dose for that chemical. A hazard
index of 1.0 or above indicated the potential for human health impacts. Table 1-14 lists the human health
hazard indices.
1-19
Environmental Consequences of Long-Term Repository Performance
ORAL REFERENCE DOSE
The oral reference dose is based on the assumption that thresholds exist tor certain toxic effects
such as cellular necrosis. This dose is expressed in units of milligrams per kilogram per day. In
general, the oral reference dose is an estimate (with uncertainty spanning perhaps an order of
magnitude) of a daily exposure to the human population (including sensitive subgroups) that is likely
to be without an appreciable risk of deleterious effects during a lifetime (EPA 1999, all).
Table 1-14. Human health hazard indices for chemically toxic materials.
Element (millierai
Boron
Cadmium
Chromium
Manganese
Molybdenum
Nickel
Uranium
Vanadium
Zinc
0.0053
0.00022
0.030
4.5 X 10-'^
0.0072
1.0x10'°
0.00000023
0.0000054
0.000074
r kilogram
per day)
Hazard index
0.09
0.059
0.0005
0.44
0.005
6.1
0.14
3.2x10-''
0.005
1.4
0.02
5.1 X 10'
0.003
0.000078
0.007
0.00078
0.3
0.00025
a. Source: EPA (1999, all).
Intake was based on a 2-liter (0.53-gallon) daily consumption rate of drinking water, at the concentrations
in the well, by a 70-kilogram (154-pound) adult. The oral reference doses were from the Integrated Risk
Information System (EPA 1999, all), with the exception of doses for uranium (EPA 1994, all) and
vanadium (International Consultants 1997, all).
Of the proposed chemically toxic materials in the repository, only chromium and molybdenum have a
hazard index above 1.0. Because the inventories of a given material category in the repository should no
more than double under any of the inventory modules, all chemically toxic materials (except chromium
and molybdenum) can be eliminated from detailed analyses. However, the analysis also considered
uranium in recognition of the special attention this element attracts and as a check for the screening
analyses.
1.3.2.4 Chromium Inventory for Use in the Performance Assessment Model
The Alloy-22 that would comprise the inner corrosion-resistant material layer of the waste packages for
the Yucca Mountain Repository design would contain 21.25 percent chromium and 55 percent nickel. In
addition, stainless-steel containers and fuel cladding would contribute a meaningful but much smaller
quantity of chromium. Table 1-15 lists the chromium that would be present in the waste packages under
the Proposed Action. Tables 1-16 and 1-17 list the chromium that would be present in the waste packages
under Inventory Modules 1 and 2, respectively.
The performance assessment model simulates a number of abstracted waste packages for each waste
category with a generalized inventory. Tables 1-18 and 1-19 summarize the assignment of the chromium
inventory under the Proposed Action derived from the actual inventory listed in Table 1-15 to the number
of abstracted waste packages simulated with the model. The inventory is separated between interior
stainless steel (Table 1-18) and waste package Alloy-22 (Table 1-19) because these two portions of the
chromium inventory are modeled separately in a two-step process (see Section 1.6 for details). Similarly,
Tables 1-20 and 1-21 summarize the assignment of the chromium inventory derived from the actual
inventory under Inventory Module 1, listed in Table 1-16, to the number of abstracted waste packages
1-20
Environmental Consequences of Long-Term Repository Performance
Table 1-15. Chromium content (kilograms) of waste packages for the Proposed Action."
Alloy-22 per
ss/b''
alloy per
Chromium
Quantity
actual waste
waste
package
waste
package
mass per
Alloy
Chromium
Alloy
Chromium
waste package
Waste category
Waste package type'
packages'"
mass
mass"
mass
mass'
type
Commercial spent
21 PWR UCF (no absorber)
1,369
4,458
947
0
0
1,296,888
nuclear fuel
21 PWR UCF (absorber plates)
2,641
4,458
947
1,883
546
3,944,056
21 PWR UCF (control rods)
169
4,458
947
0
0
160,098
12 PWR UCF (high heat)
394
3,282
697
0
0
274,785
12 PWR UCF (South Texas)
179
3,717
790
1,071
311
196,981
44 BWR UCF (no absorber)
773
4,261
905
0
0
699,923
44 BWR UCF (absorber plates)
2,024
4,261
905
3,999
1,160
4,179,909
24 BWR UCF (thick absorber)
93
3,342
710
2,141
621
123,789
High-level
5 HLW co-disposal
1,270
4,066
864
0
0
1,097,312
radioactive waste
5 HLW long co-disposal
1,007
5,687
1,208
0
0
1,216,947
DOE spent
Navy SNF long
300*
6,306
1,340
0
0
381,907
nuclear fuel
Totals
10^04
13,572^95
a. To convert kilograms to pounds, multiply by 2.2046.
b. SS/B = stainless-steel boron.
c. Abbreviations: PWR = pressurized-water reactor; UCF = uncanistered fuel; BWR = boiling-water reactor; HLW = defense
high-level radioactive waste; SNF = spent nuclear fuel.
d. Source: TRW (1999b, pages 6-5 to 6- 12); quantities of waste packages modeled for results reported in Section 1.6 differ
slightly (because of the use of earlier estimates), resulting in a total chromium inventory about 1 percent less than indicated
in this table. Final chromium impacts were not expected to differ because the inventory would not be exhausted during the
period simulated.
e. Chromium constitutes 2 1 .25 percent of Alloy-22.
f. Chromium constitutes 29 percent of SS/B alloy.
g. The analysis used 285 Navy SNF long waste packages in models for results discussed in Section 1.6. The difference resulted
in a chromium inventory that was about an additional 0.02 percent less than indicated in this table.
Table 1-16. Chromium content (kilograms) of waste packages for Inventory Module 1.
Alloy-2::
! per waste
SS/B"
alloy per
Chromium
Quantity
actual waste
package
waste
; package
mass per
waste package
Alloy
Chromium
Alloy
Chromium
Waste category
Waste package type"
packages'"
mass
mass'
mass
mass'
type
Commercial spent
21 PWR UCF (no absorber)
2,339
4,458
947
0
0
2,215,793
nuclear fuel
21 PWR UCF (absorber plates)
4,228
4,458
947
1,883
546
6,314,074
21 PWR UCF (control rods)
314
4,458
947
0
0
297,460
12 PWR UCF (high heat)
646
3,282
697
0
0
450,537
12 PWR UCF (South Texas)
428
3,717
790
1,071
311
470,994
44 BWR UCF (no absorber)
1,242
4,261
905
0
0
1,124,584
44 BWR UCF (absorber plates)
3,195
4,261
905
3,999
1,160
6.598,226
24 BWR UCF (thick absorber)
186
3,342
710
2,141
621
247,578
High-level
5 HLW co-disposal
1,557
4,066
864
0
0
1,345,287
radioactive waste
5 HLW long co-disposal
3,000
5,687
1,208
0
0
3,625,463
DOE spent nuclear
fuel
Totals
Navy SNF Long
300
6,306
1,340
0
0
402,008
17,435
23,092,003
a. To convert kilograms to pounds, multiply by 2.2046.
b. SS/B = stainless-steel boron.
c. Abbreviations: PWR = pressurized-water reactor; UCF = uncanistered fuel; BWR = boiling-water reactor; HLW = defense
high-level radioactive waste; SNF = spent nuclear fuel.
d. Source: TRW (1999b, pages 6-5 to 6- 12); quantities of waste packages modeled for results reported in Section 1.6 differ
slightly (because of the use of earlier estimates), resulting in a total chromium inventory about 1 percent less than indicated
in this table. Final chromium impacts were not expected to differ because the inventory would not be exhausted during the
period simulated.
e. Chromium constitutes 2 1 .25 percent of Alloy-22.
f Chromium constitutes 29 percent of SS/B alloy.
1-21
Environmental Consequences of Long-Term Repository Performance
Table 1-17. Chromium content (kilograms) of waste packages for Inventory Module 2/
Chromium
mass per
waste
Chromium package
Waste
Quantity
actual
waste
category
Waste package type
packages
Alloy-22 per
waste package
Alloy
mass
SS/B" alloy per
waste package
Chromium
mass'
Alloy
mass
mass
type
Commercial
spent nuclear
fuel
High-level
radioactive
waste
DOE spent
nuclear fuel
GTCC and
SPAR«
Totals
21 PWR UCF (no absorber)
21 PWR UCF (absorber plates)
21 PWR UCF (control rods)
12 PWR UCF (high heat)
12 PWR UCF (South Texas)
44 BWR UCF (no absorber)
44 BWR UCF (absorber plates)
24 BWR UCF (thick absorber)
5 HLW co-disposal
5 HLW long co-disposal
Navy SNF long
5 HLW long co-disposal
2,339
4,228
314
646
428
1,242
3,195
186
1,557
3,000
300
608
18,043
4,458
4,458
4,458
3,282
3,717
4,261
4,261
3,342
4,066
5,687
6,306
5,687
947
947
947
697
790
905
905
710
864
1,208
1,340
1,208
0
1,883
0
0
1,071
0
3,999
2,141
0
0
0
0
0
546
0
0
311
0
1,160
621
0
0
0
0
2,215,793
6,314,074
297,460
450,537
470,994
1,124,584
6,598,226
247,578
1,345,287
3,625,463
402,008
734,760
23,826,763
a. To convert kilograms to pounds, multiply by 2.2046.
b. SS/B = stainless-steel boron.
c. Abbreviations: PWR = pressurized-water reactor; UCF = uncanistered fuel; BWR = boiling-water reactor; HLW = defense
high-level radioactive waste; SNF = spent nuclear fuel.
d. Source: TRW (1999b, pages 6-5 to 6-12); quantities of waste packages modeled for results reported in Section 1.6 differ
slightly (because of the use of earlier estimates), resulting in a total chromium inventory about 1 percent less than indicated
in this table. Final chromium impacts were not expected to differ because the inventory would not be exhausted during the
period simulated.
e. Chromium constitutes 2 1 .25 percent of Alloy-22.
f. Chromium constitutes 29 percent of SS/B alloy.
g. GTCC = Greater- Than-Class-C waste; SPAR = Special-Performance-Assessment-Required waste.
Table 1-18. Modeled waste package interior chromium inventory for Proposed Action (kilograms)."
Number of Mass per
Mass per Mass per abstracted abstracted
waste package waste waste waste
Waste category
Waste package type
type
category packages package
Commercial spent
nuclear fuel
High-level
radioactive waste
DOE spent nuclear
fuel
Totals
21 PWR UCF (no absorber)
21 PWR UCF (absorber plates)
21 PWR UCF (control rods)
12 PWR UCF (high heat)
12 PWR UCF (South Texas)
44 BWR UCF (no absorber)
44 BWR UCF (absorber plates)
24 BWR UCF (thick absorber)
5 HLW co-disposal
5 HLW long co-disposal
Navy SNF long
0
1,442,171
0
0
55,596
0
2,347,253
57,743
0
0
0
3,902,762
3,902,762
7,760
0 1,663
0 2,546
3,902,762 11,969
503
0
0
a. To convert kilograms to pounds, multiply by 2.2046.
b. Abbreviations: PWR = pressurized-water reactor; UCF = uncanistered fuel; BWR :
high-level radioactive waste; SNF = spent nuclear fuel.
c. Source: Table 1-15.
boiling-water reactor; HLW = defense
1-22
Environmental Consequences of Long-Term Repository Performance
Table 1-19. Modeled corrosion-resistant material (Alloy-22) chromium inventory (kilograms) for
Proposed Action/^
Mass per
Mass per
Number of
Mass per
waste package
waste
abstracted
abstracted
Waste category
Waste package type""
type^
category
waste packages
waste package
Commercial spent
21 PWR UCF (no absorber)
1,296,888
6,973,667
7,760
899
nuclear fuel
21 PWR UCF (absorber plates)
21 PWR UCF (control rods)
12 PWR UCF (high heat)
12 PWR UCF (South Texas)
44 BWR UCF (no absorber)
44 BWR UCF (absorber plates)
24 BWR UCF (thick absorber)
2,501,885
160,098
274,785
141,385
699,923
1,832,656
66,046
High-level
5 HLW co-disposal
1,097,312
2,314,259
1,663
1,392
radioactive waste
5 HLW long co-disposal
1,216,947
DOE spent nuclear
fuel
Totals
Navy SNF long
381,907
381,907
2,546
150
9,669,833
9,669,833
11,969
a. To convert kilograms to pounds, multiply by 2.2046.
b. Abbreviations: PWR = pressurized-water reactor; UCF =
uncanistered fuel
BWR = boiling-water reactor;
HLW = defense
high-level radioactive waste; SNF = spent nuclear fuel.
c. Source: Table 1-15.
Table 1-20. Modeled waste package interior chromium inventory (kilograms)
for Inventory
Module 1.'
Mass per
Mass per
Number of
Mass per
waste package
waste
abstracted
abstracted
Waste category
Waste package type''
type^
category
waste packages
waste package
Commercial spent
21 PWR UCF (no absorber)
0
6,262,475
12,932
484
nuclear fuel
21 PWR UCF (absorber plates)
21 PWR UCF (control rods)
12 PWR UCF (high heat)
12 PWR UCF (South Texas)
44 BWR UCF (no absorber)
44 BWR UCF (absorber plates)
24 BWR UCF (thick absorber)
2,308,784
0
0
132,933
0
3,705,273
115,486
High-level
5 HLW co-disposal
0
0
4,456
0
radioactive waste
5 HLW long co-disposal
0
DOE spent nuclear
fuel
Totals
Navy SNF long
0
0
4,340
0
6,262,475
6,262,475
21,728
a. To convert kilograms to pounds, multiply by 2.2046.
b. Abbreviations: PWR = pressurized-water reactor; UCF =
high-level radioactive waste; SNF = spent nuclear fuel.
c. Source: Table 1-16.
uncanistered fuel; BWR = boiling-water reactor; HLW = defense
simulated with the performance assessment model for interior stainless steel and corrosion-resistant
material, respectively.
Inventory Module 2 is simulated as an incremental impact over Inventory Module 1, where the difference
is in the Greater-Than-Class-C and Special-Performance-Assessment-Required wastes added under
Inventory Module 2. Table 1-22 summarizes the assignment of the additional chromium inventory
derived from the actual inventory for Inventory Module 2 to the number of abstracted waste packages
simulated with the performance assessment model. No interior stainless steel would be included in the
additional waste packages under Inventory Module 2.
1-23
Environmental Consequences of Long-Term Repository Performance
Table 1-21. Modeled corrosion-resistant material (Alloy -22) chromium inventory (i<;ilograms) for
Inventory Module 1 ."
Mass per
Mass per
Number of Mass per
waste package
waste
abstracted abstracted
Waste category
Waste package type*"
type'
category
waste packages waste package
Commercial spent
21 PWR UCF (no absorber)
2,215,793
11,456,771
12,932 886
nuclear fuel
21 PWR UCF (absorber plates)
21 PWR UCF (control rods)
12 PWR UCF (high heat)
12 PWR UCF (South Texas)
44 BWR UCF (no absorber)
44 BWR UCF (absorber plates)
24 BWR UCF (thick absorber)
4,005,290
297,460
450,537
338,061
1,124,584
2,892,953
132,093
High-level
5 HLW co-disposal
1,345,287
4,970,749
4,456 1,116
radioactive waste
5 HLW long co-disposal
3,625,463
DOE spent nuclear
fuel
Totals
Navy SNF long
402,008
402,008
4,340 93
16,829,528
16,829,528
21,728
a. To convert kilograms to pounds, multiply by 2.2046.
b. Abbreviations: PWR = pressurized-water reactor; UCF = uncanistered fuel; BWR = boiling-water reactor; HLW = high-
level radioactive waste; SNF = spent nuclear fuel.
c. Source: Table 1-17.
Table 1-22. Additional corrosion-resistant material (Alloy-22) chromium inventory for Inventory
Module 2 in excess of inventory for Module 1 (kilograms)."
Number of
Mass per
Mass per
Mass per
abstracted
abstracted
waste package
waste
waste
waste
Waste category
Waste package type*"
type'
category
packages
package
GTCC+SPAR''
5 HLW long co-disposal
734,760
734,760
1,642
447
a. To convert kilograms to pounds, multiply by 2.2046.
b. Abbreviations: HLW = high-level radioactive waste.
c. Source: Table 1-17.
d. GTCC = Greater- Than-Class-C waste; SPAR = Special-Performance-Assessment-Required waste.
1.3.2.5 Elemental Uranium Inventory for Use In the Performance Assessment Model
Table 1-23 lists the total inventory of elemental uranium (that is, all isotopes of uranium) for consideration
as a chemically toxic material for the Proposed Action and Inventory Modules 1 and 2. The total uranium
inventory for both Inventory Modules 1 and 2 would be about 70 percent greater than that for the
Proposed Action. The uranium content in high-level radioactive waste was set to the equivalent of metric
tons of heavy metal (MTHM) for this analysis, though much of the uranium would have been removed
during reprocessing operations. The elemental uranium inventory for Modules 1 and 2 would be
essentially equivalent because Greater-Than-Class-C and Special-Performance-Assessment-Required
wastes (the only additional waste in Module 2 over Module 1) do not contain substantial quantities of
uranium.
1.3.2.6 Molybdenum Inventory
The Alloy-22 used for the corrosion-resistant material contains 13.5 percent molybdenum. During the
corrosion of the Alloy-22, molybdenum behaves almost the same as chromium. Due to the corrosion
conditions, molybdenum also dissolves in a highly soluble hexavalent form. Therefore, the source term
for molybdenum will be exactly 13.5/21.25 times the source term for chromium (or 64 percent) from
Alloy-22 only.
1-24
Environmental Consequences of Long-Term Repository Performance
Table 1-23. Total elemental uranium inventory (kilograms)^ for Proposed Action and Inventory Modules
1 and Z."-'"
Inventory
Commercial SNF*
HLW^
DOESNfF
Totals
Proposed Action
Modules 1 and 2^
63,000,000
105,000,000
4,700,000
13,000,000
2,300,000
2,500,000
70,000,000
120,000,000
a. To convert kilograms to pounds, multiply by 2.2046.
b. The uranium content in high-level radioactive waste was set to the MTHM equivalent for this analysis, even though much of
the uranium would have been removed during reprocessing operations.
c. Rounded to two significant figures.
d. Source: Appendix A, Tables A-12, A-13, A-19, A-29 to A-34.
e. SNF = spent nuclear fuel.
f. HLW = high-level radioactive waste.
g. Inventory Module 1 and 2 will have the same total uranium inventory because Greater-Than-Class-C and Special-
Performance- Assessment-Required waste (the only additional waste in Module 2 over Module 1) does not contain a
substantial quantity of uranium.
1.3.3 ATMOSPHERIC RADIOACTIVE MATERIALS
The only radionuclide that would have a relatively large inventory and a potential for gas transport would
be carbon- 14. Iodine- 129 can exist in a gas phase, but it is highly soluble and therefore likely to dissolve
in groundwater rather than migrate as a gas. After carbon- 14 escaped from the waste package, it could
flow through the rock in the form of carbon dioxide. About 2 percent of the carbon- 14 in commercial
spent nuclear fuel occurs in a gas phase in the space (or gap) between the fuel and the cladding around the
fuel (Oversby 1987, page 92). The gas-phase inventory consists of 0.23 curie of carbon-14 per
commercial spent nuclear fuel waste package. Table 1-24 lists the total carbon-14 inventory for the
repository under the Proposed Action and Inventory Modules 1 and 2.
Table 1-24. Total carbon-14 inventory (curies)."
Inventory
Solid'' Gaseous"
Totals"
Proposed Action
Module 1
Module 2
92,000 1,800
150,000 3,200
240,000 3,200
93,000
160,000
240,000
a. Source: Appendix A, Table A- 10.
b. Impacts of carbon-14 in solid form are addressed as waterbome
radioactive material impacts.
c. Based on 0.234 curie of carbon-14 per commercial spent nuclear fuel
waste package.
d. Totals are rounded to two significant figures.
1.4 Extension of Total System Performance Assessment Methods and
Models for EiS Analyses
DOE conducted analyses for the Total System Performance Assessment - Viability Assessment to
evaluate potential long-term impacts to human health from the release of radioactive materials from the
Yucca Mountain Repository. The analyses for this EIS were conducted in conjunction with, but distinct
from, the calculations for the Viability Assessment (DOE 1998a, Volume 3, all). The methodologies and
assumptions for the Viability Assessment are detailed in TRW (1998a,b,c,d,e,f,g,h,i,j,k, all). Extensions
of the Viability Assessment analyses to meet distinct EIS requirements (for example, consideration of
different thermal load scenarios or inventories) were made using the same overall methodology, and
details of these extensions are provided in this section. Additional information on EIS performance-
assessment analyses can be found in TRW (1999a, all).
1-25
Environmental Consequences of Long-Term Repository Performance
1.4.1 REPOSITORY DESIGN FOR ALTERNATIVE THERMAL LOADS
The spatial density at which the waste packages are emplaced in the repository is generally quantified
using thermal load, which is the MTHM emplaced per acre of repository area. The higher the thermal
load, the smaller the spacing between waste packages, resulting in a higher thermal output per unit area.
The area required for emplacement is based on the target thermal loads attained by varying the spacing
between the waste packages and the distance between the emplacement drifts. The commercial spent
nuclear fuel heat output dominates the overall heat load and thus the total emplacement area required.
Thus, for purposes of thermal modeling, the Proposed Action inventory implies the nominal value of
63,000 MTHM commercial spent nuclear fuel, whereas Inventory Modules 1 and 2 have the same
expanded inventory of 105,000 MTHM commercial spent nuclear fuel.
Table 1-25 gives the estimates of repository area required for the emplacement of wastes, ranging from a
low of 740 acres for the high thermal load scenario with the Proposed Action inventory case to a high of
4,200 acres for the low thermal load scenario with the Inventory Module 1 or 2 case. Most of the options
require waste emplacement in areas beyond the primary, or upper, emplacement block, which is
juxtaposed between the Solitario Canyon Fault and the Ghost Dance Fault. The upper emplacement block
is the reference repository region in the Viability Assessment base case facility design (63,000 MTHM
high thermal load scenario). Selection of potential expansion blocks near the upper block was carried out
using several criteria:
• Availability of 200 meters (660 feet) of overburden
• Consistency of elevation and dip with the upper block
• Distance from the saturated zone
• Favorable excavation characteristics
These considerations are described in detail in TRW (1999b, all).
Table 1-25. Estimates of repository emplacement area."
Area (acres)''
Thermal load Drift spacing
(MTHM per acre) (meters)'^
Proposed Action
Inventory Modules
land 2
85 28
60 40
25 38"
740
1,050
2,520
1,240
1,750
4,200
Source: TRW (1999a, Table 2.3-1, page 2-12) based on 63,000 MTHM of commercial spent
nuclear fuel.
b. To convert acres to square miles, divide by 640.
c. To convert meters to feet, multiply by 0.3048.
d. Under the low thermal load, the waste packages would be placed in an approximately square
pattern so that the thermal load was distributed evenly. To accomplish this, the emplacement
drift spacing and the spacing of the waste packages in the emplacement drift would be
approximately equal (TRW 1999c, page F-2).
The selected inventory layouts for the Proposed Action and Inventory Modules 1 and 2 for the high,
intermediate, and low thermal load scenarios are shown in Figures 1-2 through 1-7. These layouts,
simplified from the original engineering layouts presented in TRW (1999c, Figures 3.3-1 through 3.3-6),
indicate that the wastes for these thermal loads can be accommodated within the upper blocks, the lower
block, and one additional region (Block la) to the west of the Solitario Canyon Fault.
1-26
Environmental Consequences of Long-Term Repository Performance
As described in TRW (1999c, all), additional subsurface blocks for emplacement of waste according to
intermediate and low thermal load scenarios were identified by:
• Expanding the upper block to the north and south
• Expanding the lower block to the north and east
• Lowering the elevation of Block la, combining it with Block lb, and designating the combined area
as Block 5
• Raising the elevation of Block 2 by 15 meters (50 feet) and designating it as Block 6
Raising the elevation of Block 3 by 12 meters (39 feet) and designating it as Block 7
•
• Raising the elevation of Block 4 by 2 meters (6.6 feet), extending the area to the south, and
designating it as Block 8
The corresponding layouts for the low thermal load scenario for the Proposed Action and for Inventory
Modules 1 and 2 are shown in Figures 1-6 and 1-7, respectively. Figure 1-8 shows the relationship
between the early Proposed Action designs and the design areas considered in these EIS analyses.
1.4.2 THERMAL HYDROLOGY MODEL
Evaluation of the intermediate (60 MTHM per acre) and low (25 MTHM per acre) thermal load scenarios
for this EIS diverged from the high thermal load base case evaluated in the Viability Assessment.
Extensions of the thermal-hydrologic modeling supporting the total systems performance assessment
model were required to evaluate these additional thermal load scenarios. These extensions are detailed in
this section.
1.4.2.1 Thermal-Hydrologic Scenarios
The analysis of waste package degradation and engineered barrier system release for the EIS requires
information regarding waste package temperature and relative humidity, and liquid saturation and
temperature within the repository invert. These data were derived from the development and application
of a suite of three-dimensional, drift-scale models for predicting the thermal-hydrologic environment near
the waste packages. Six sets of calculations were carried out to handle the two inventory options (63,000
and 105,000 MTHM) and the three thermal load scenarios (85, 60, and 25 MTHM per acre). The
simulations were performed using NUFT, an integrated finite-difference code capable of modeling
multidimensional fluid flow, solute migration, and heat transfer in porous and/or fractured media (Nitao
1998, all).
These calculations closely parallel the thermal-hydrologic modeling study performed in support of Total
System Performance Assessment - Viability Assessment (TRW 1998c, all). The main difference
between the two studies is in the treatment of thermal-hydrologic conditions at the edge of the repository.
In Total System Performance Assessment - Viability Assessment, a hybrid methodology with
complementary thermal-hydrologic and thermal conduction models is used to delineate different thermal-
hydrologic zones within the repository horizon (TRW 1998c, all). In this study, a less detailed scaling
methodology is used to divide the repository into center and edge regions because of the computational
complexities associated with larger inventories and expanded emplacement regions. This less detailed
scaling methodology is not expected to adversely impact the results.
1-27
Environmental Consequences of Long-Term Repository Performance
1.4.2.2 Waste Package and Drift Geometry
Following the approach taken in Total System Performance Assessment - Viability Assessment, the basic
three-dimensional drift-scale model was developed around a discrete waste package symmetry element.
This model extends:
• In the x-direction, from the drift centerline to the midpoint between adjacent drifts
• In the y-direction, over a representative number of packages to capture the package-to-package
variability in heat output
• In the z-direction, from the ground surface to the water table
The vertical discretization between the ground surface and the water table was chosen to be consistent
with the Lawrence Berkeley National Laboratory three-dimensional, site-scale unsaturated flow model
(Bodvarsson, Bandurraga, and Wu 1997, all). The basis for the model discretization in the other two
dimensions is described in the following paragraphs.
The Proposed Action inventory consists of 63,000 MTHM of commercial spent nuclear fuel, 4,667
MTHM of high-level radioactive waste, and 2,333 MTHM of DOE spent nuclear fuel. As described in
DOE (1998a, Volume 3, Figure 3-18, page 3-31), the corresponding symmetry element contains seven
packages:
• Three 2 l-pressurized-water-reactor waste packages
• Two 44-boiling-water-reactor waste packages
• One-half of a 12-pressurized-water-reactor waste package
• One-half of a direct-disposal waste package (containing four DOE spent nuclear fuel N-reactor
canisters)
• One co-disposal waste package (containing five high-level radioactive waste glass-filled canisters
with or without a DOE spent nuclear fuel canister)
Inventory Module 1 consists of 105,000 MTHM of commercial spent nuclear fuel, 12,600 MTHM of
high-level radioactive waste (based on MTHM equivalency discussion in Section A.2.3.1 of Appendix A
of this EIS), and 2,500 MTHM of DOE spent nuclear fuel. Accordingly, the expanded inventory
symmetry element was created using a total of nine packages:
• Three and one -half 2 l-pressurized-water-reactor waste packages
• Two and one-half 44-boiling-water-reactor waste packages
• One 12-pressurized-water-reactor waste package
• Two co-disposal waste packages containing five high-level radioactive waste glass-filled canisters
(with or without a DOE spent nuclear fuel canister)
Note that this symmetry element model maintains the relative percentage (and heat output) of different
package types while minimizing the total number of discrete packages for computational convenience.
This package discretization model was deemed adequate from the standpoint of thermal-hydrologic
modeling, although it is only an approximation of the true inventory.
1-28
Environmental Consequences of Long-Term Repository Performance
For the high (85 MTHM per acre) and intermediate (60 MTHM per acre) thermal load scenarios, the
waste package arrangement within the drifts was kept constant, and the drift spacing was adjusted to
attain the correct thermal load levels. Thus, the high thermal load scenario yields drift spacing of
28 meters (about 92 feet) and the intermediate thermal load scenario yields drift spacing of 40 meters
(about 130 feet). For the low (25 MTHM per acre) thermal load scenario, maintaining the same waste
package arrangement as for the high and intermediate thermal load scenarios would have required the
drifts to be spaced too far apart in the x-direction, resulting in localized heating effects. Therefore, the
package-to-package spacing in the y-direction was increased for the low thermal load scenario to create
an approximately square symmetry element, including drift spacing of 38 meters (about 120 feet). Waste
package spacing for the Proposed Action and for Inventory Modules 1 and 2 is summarized in Table 1-26
and Table 1-27, respectively.
Table 1-26. Waste package spacing for the Proposed Action inventory."
Spacing of gap after given package
(meters)
Waste package
Waste package
High and intermediate
type
width (meters)
thermal load
Low thermal load
12-PWR
V2 (5.87)
6.021
26.424
21-PWR
5.3
9.276
31.215
21-PWR
5.3
2.949
15.415
Co-disposal
5.37
2.2535
13.676
21-PWR
5.3
8.929
30.345
44-PWR
5.3
7.98
27.969
44-BWR
5.3
1.305
11.2996
Direct-disposal
V2 (5.37)
a. Source; TRW (1999a, Table 3.2-1, page 3-3).
b. To convert meters to feet, multiply by 0.3048.
Table 1-27. Waste package spacing for Inventory Modules 1 and 2."
Spacing of gap after given package
(meters)
Waste package
Waste packc
ige
High
1 and intermediate
type
width (meters)
thermal load
Low thermal load
21-PWR
V^(5.3)
2.949
11.3055
Co-disposal
5.37
2.2535
17.79
21-PWR
5.3
9.95
32.902
21-PWR
5.3
10.02
33.081
21-PWR
5.3
7.39
26.9175
12-PWR
5.87
6.368
24.3615
44-PWR
5.3
7.98
27.969
44-BWR
5.3
1.305
12.599
Direct-disposal
5.37
1.305
10.0
44-BWR
>/2 (5.3)
a. Source: TRW (1999a, Table 3.2-2,
page
3-4).
b. To convert meters to feet, multiply
by 0.3048.
1.4.2.3 Selection of Submodels
Engineering layouts developed for waste emplacement were shown in Figures 1-2 through 1-7. These
layouts suggest that multiple, discontinuous heated regions will develop in the postclosure period for
some of the options. A full three-dimensional representation of all heated regions (such as emplacement
areas) was not considered computationally practical. Therefore, for modeling purposes each region was
treated as an isolated entity by assuming that boundaries existed for no heat flow and no fluid flow
between the regions. Furthermore, to capture the effects of varying stratigraphy and variable surface
1-29
Environmental Consequences of Long-Term Repository Performance
infiltration on the thermal-hydrology response at the repository, each emplacement block was modeled by
a representative stratigraphic column or submodel. These submodel solution assumptions are unlikely to
affect adversely the results reported in this EIS.
Based on the original design layouts (see Figure 1-2), each thermal load scenario was to be modeled using
some combination of each of the following seven stratigraphic columns:
Upper Block (stratigraphic column 1)
Lower Block (stratigraphic column 2)
Block la (stratigraphic column 3)
Block lb (stratigraphic column 7)
Block 2 (stratigraphic column 5)
Block 3 (stratigraphic column 6)
Block 4 (stratigraphic column 4)
These submodels were used for the high and intermediate thermal load scenarios. However, because of
the large areal extent required for the low thermal load scenario, the engineering layout changed for those
two design options. In the new design layout. Block lb has been combined with part of Block la to form
Block 5, while part of Block la has been combined with Block 4 to form Block 8. These two new areas
can be represented by two existing submodels: stratigraphic column 7 for Block 5 and stratigraphic
column 4 for Block 8. This information is summarized in Table 1-28 and shown on Figure 1-8.
Table 1-28. Areas of submodels (stratigraphic columns) used in thermal-hydrologic calculations."
Thermal-
hydrologic
scenario
Loading
(MTHM
per acre)
Waste package
inventory
module
Emplacement
block
Stratigraphic
column
number
Actual area
(acres)
Percent
of area
1
85
Proposed Action
Upper Block
1
740
100.0
2
60
Proposed Action
Upper Block
1
1,050
100.0
3
25
Proposed Action
Upper Block
1
1,110
44.0
Lower Block
2
596
23.7
Block 5
7
814
32.3
4
85
Inventory
Upper Block
1
1,180
95.5
Modules 1 and 2
Lower Block
2
55
4.5
5
60
Inventory
Upper Block
1
1,180
67.4
Modules I and 2
Lower Block
2
380
21.7
Block la
3
190
10.9
6
25
Inventory
Upper Block
1
1,110
26.4
Modules 1 and 2
Lower Block
2
596
14.2
Block 5
7
814
19.4
Block 6
5
420
10.0
Block?
6
440
10.5
Block 8
4
820
19.5
a. Source: TRW (1999a, Table 3.2-3, page 3-5).
For all submodels, the vertical stratigraphic data for the model stratigraphic columns were extracted from
the Lawrence Berkeley National Laboratory site-scale model (Bodvarsson, Bandurraga, and Wu 1997,
all), with the exception of Block 2 and Block 3, which lie outside the boundaries of the site-scale model.
The geologic framework model (TRW 1997d, all) was used to develop the stratigraphy for the columns
corresponding to Block 2 and Block 3 even though very little information is available regarding the
stratigraphy, hydrology, and infiltration conditions in this sector of the Yucca Mountain site. Thermal-
hydrologic simulations were carried out with these two submodels for the low thermal load with
expanded inventory scenario, but the simulations were not used for the subsequent total-system
calculations. It was assumed that the thermal-hydrologic results for these regions could be approximated
by the neighboring regions within the Berkeley model domain. Thus, the submodel for Block 8
1-30
Environmental Consequences of Long-Term Repository Performance
(stratigraphic column 4) was assumed also to represent Block 3, and the submodel for Block 5
(stratigraphic column 7) was assumed also to represent Block 2.
1.4.2.4 Hydrology and Climate Regime
Hydrologic properties for the thermal-hydrologic models were taken to be the same as the Total System
Performance Assessment - Viability Assessment base case (TRW 1998c, Section 3.5). These properties
include matrix and fracture characteristics describing capillary retention and relative permeability for a
dual-permeability model, including fracture-matrix-interaction area-reduction factor terms that were
adjusted to match observed borehole saturations. As described in RamaRao, Ogintz, and Mishra (1998,
pages 1 16 to 118), the dual-permeability model parameters have been adjusted for the present study using
the "satiated saturation" concept in the generalized equivalent continuum model. Using a porosity-
weighted average, the dual-permeability model fracture and matrix parameters (porosity and
permeability) are combined to create corresponding parameters for the generalized equivalent continuum
model, while the satiated saturation concept is used to set the threshold for the initiation of flow in
fractures (before the attainment of full matrix saturation). Subsequently, the composite medium capillary
characteristics are generated by a porosity-weighted average of the individual media curves. These
hydrologic properties, as well as other thermal properties used in the thermal-hydrologic calculations, are
discussed in TRW (1998c, Section 3.2.1, pages 3-21 to 3-26).
This EIS performance assessment considered three climate scenarios: present-day, long-term average
(wetter than the present-day climate), and superpluvial, which are added at short-duration, fixed intervals
on a periodic basis during the 100,000-year period after waste emplacement. In the performance
assessment model, the initial conditions (that is, the present-day climate) are multiplied by 5.45 to obtain
the long-term average climate and by 14.30 to obtain the super-pluvial climate (DOE 1998a, Volume 3,
Figure 4.2, page 4-4). The climate changes are measured in step-changes for the duration of the climate
periods, and the sequence lengths are 10,000 years for the present-day dry climate and the super-pluvial
climate, and 90,000 years for the long-term average climate. The sequence of climate changes used for
expected-value simulations (which use the mean value of probabilistically defined input variables) is:
0 to 5,000 years - present-day (dry) climate
5,001 to 95,(X)0 years - long-term average climate
95,(X)1 to 105,000 years - present-day (dry) climate
105,001 to 195,000 years - long-term average climate
195,001 to 205,000 years - present-day (dry) climate
205,001 to 285,000 years - long-term average climate
285,001 to 295,000 years - super-pluvial climate
295,001 to 305,000 years - present-day (dry) climate
This sequence is repeated for the duration of the simulation period.
Expected-value simulations were carried out for the first 1 million years after closure, to include the
complete decay of waste heat caused by radioactive decay and a return to ambient conditions. To
establish appropriate initial conditions for the thermal-hydrologic simulations, the nominal present-day
(dry) climate scenario, as used in the Viability Assessment base case (TRW 1998c, Section 3.5), was used
for the ambient hydrologic calculations. A separate set of thermal-hydrologic simulations was then
performed for each climate condition, as required. This approach is consistent with that used in the
Viability Assessment, in which climate effects on thermal hydrology for the entire period were included
by making three sets of calculations (for present-day, long-term average, and superpluvial climates). The
influence of climate change on thermal-hydrologic system response was then approximated in the
performance assessment model total-system simulator by switching from one set of results to the other at
the time of climate change.
1-31
Environmental Consequences of Long-Term Repository Performance
For both the present-day and long-term average climate, the infiltration flux at the top of each
representative column was extracted from the flux associated with the nearest element in the Lawrence
Berkeley National Laboratory site-scale model (Bodvarsson, Bandurraga, and Wu 1997, all). However,
there was no infiltration information available for stratigraphic columns 5 and 6, which are located
outside the Berkeley model boundary. Therefore, the infiltration fluxes for these columns were assumed
to be equal to the fluxes at the nearest element within the Berkeley model boundary. Note that these
infiltration rates were assumed to be constant throughout the 1 -million-year postemplacement period with
climate changes implemented by multiplying the infiltration rate as described above.
1.4.2.5 Treatment of Edge Effects
The drift-scale modeling results, developed using a representative symmetry element with periodic lateral
boundary conditions, best represents the conditions at the center of the repository. To account for the
edge-cooling effects experienced by exterior drifts located near unheated rock mass, a scaling
methodology was developed based on the hypothesis that the repository can be divided into at least two
thermal-hydrologic regions for grouping waste packages, a center region and an edge region. The center
region was designed so periodic boundary conditions (no-flow thermal and hydrologic boundaries) could
be assigned in a lateral direction. The edge region has a more complicated response because of edge-
cooling effects. However, it is believed that the thermal-hydrologic response at the edge is similar to that
for the center, albeit at a lower thermal load. Thus, the objective of the scaling methodology was
two-fold:
1 . Devise a strategy for generating the thermal load scale factors so models representative of the center
can be used to simulate the edge response.
2. Estimate the fraction of the repository area enclosed within the center or edge regions.
The following sections briefly describe the development and testing of the components of this scaling
methodology.
1.4.2.5.1 Scaling Factors for Edge Effects
Based on the conceptual model that the edge response is similar to the center response at a lower thermal
load, two-dimensional results from an east-west cross-section scale model of the mountain were
compared to a set of one-dimensional runs representing the edge at a series of different thermal loads.
The objective was to find a scaling factor for the thermal loads which would provide agreement between
the two-dimensional and one-dimensional runs with respect to (I) time history of temperature, liquid
saturation, and the mass fraction of air at the repository horizon; and (2) vertical profiles of temperature,
liquid saturation, and the mass fraction of air at different points in time.
These calculations were carried out for the base case hydrologic properties and infiltration regime
described earlier. The selection of the optimal scaling factor was performed by visual examination and
restricted to one scaling factor for the early-time period (0 to 1,000 years) and a second scaling factor for
the late-time period (1,000 years to 100,000 years).
Figure 1-9 shows the comparison between the two-dimensional and one-dimensional model results using
scale factors of 0.8 and 0.6. This comparison suggests that a scale factor of 0.8 is more appropriate for
the early-time period, and a scale factor of 0.6 is more suitable for the late-time period. Although not
shown here, examining vertical profiles of the primary variables at two different points in time (100 years
and 10,000 years) yielded similar observations. Note that a single scaling factor can only provide a gross
average match of all stated variables; thus, the match between two-dimensional and scaled one-
dimensional results is never perfect. Furthermore, categorization of only two scale factors (early-time and
late-time periods) is primarily for computational convenience. These simplifications notwithstanding, the
1-32
Environmental Consequences of Long-Term Repository Performance
scaling methodology appears to be a reasonable and practical strategy for generating the edge response
without resorting to more complex three-dimensional models containing both heated drifts and unheated
rock mass.
1.4.2.5.2 Definition of Thermal-Hydrologic Zones
The spatial division of the repository into center and edge regions is based on the approximation of the
diffusive temperature profile at the repository by a step function. The temperature profile at selected time
steps was extracted and fitted with equivalent step functions. The fraction of area enclosed within the
temperature discontinuity was then taken as the fraction of repository belonging to the center region. This
process is schematically demonstrated for the high thermal load scenario in Figure I- 10.
The fractional areas were found to be time-dependent. For the high thermal load scenario, the thermal-
hydrologic response is nearly the same for the entire repository as long as the boiling period is active.
Thereafter, for all practical purposes, the fraction belonging to the center stabilizes at about 0.66 (this is
the recommended fraction to be used at all times for waste package degradation calculations). For the
intermediate thermal load scenario, the fractional area belonging to the center region is found to be close
to unity at early- and late-time periods, dropping to approximately 0.6 at intermediate times. Therefore, a
time-averaged value of 0.8 is recommended as the fractional area belonging to the center for this thermal
load. Edge effects are not considered important for the low thermal load scenario, because the use of
multiple emplacement blocks will tend to elevate the temperature between adjacent blocks, thus
minimizing edge-cooling effects.
1.4.2.6 Results
As mentioned earlier, thermal-hydrologic modeling results in the form of waste package temperature and
relative humidity are required for waste package degradation calculations in WAPDEG. In addition,
temperature and liquid saturation within the invert supporting the waste packages is required for
Engineered Barrier System release calculations in the repository integration program model. Such
information is extracted from NUFT output files and archived in tabular form for input to WAPDEG and
the repository integration program model. In this section, a brief discussion of the sensitivity of the
thermal-hydrologic simulation results to various design options and natural-system uncertainties will be
presented.
1.4.2.6.1 Variability Among the Waste Pacliages
Figures I-l 1 and 1-12 show the temperature and relative humidity histories for the various waste package
types for the Proposed Action inventory at high and low thermal loads, respectively. For the high thermal
load scenario, the highest peak temperature would result from the use of the 2 1 -pressurized-water-reactor
design package, whereas the lowest peak temperature would result from the use of the direct disposal
package. These peaks differ by approximately 80°C (I76°F). The temperature history for the
21 -pressurized-water-reactor average waste package falls near the middle of this range. Note, however,
the convergence in temperature and relative humidity for all packages as the temperature drops below the
nominal boiling point [100°C (212°F)]. The small differences in temperature and relative humidity
histories for the waste packages from this time onward would not affect the WAPDEG-predicted package
degradation rates in a meaningful manner. Therefore, results from only the 21 -pressurized-water-reactor
average waste package are provided as representative inputs to WAPDEG.
1.4.2.6.2 Sensitivity to Thermal Loads
Figure 1-13 shows the temperature and relative humidity histories for the three thermal loads and both
Proposed Action and Inventory Modules 1 and 2 scenarios. As expected, the relative peak temperatures
correspond to the magnitude of the thermal loads. For each thermal load, the expanded inventory gives a
1-33
Environmental Consequences of Long-Term Repository Performance
slightly higher peak temperature result, but the two inventories converge quickly at later times.
Calculations for the high and intermediate thermal load scenarios result in similar curves, both in terms of
temperature and relative humidity. For the low thermal load scenario, the shape of the curve is much
flatter and the temperature drops below 100°C (212°F) much earlier than the other scenarios.
1.4.2.6.3 Comparison Between Center and Edge Locations
Figure 1-14 shows a comparison between temperature and relative humidity histories calculated for the
high thermal load scenario using both center and edge models. The edge model is essentially the center
model with a lower heat load. As described in Section 1.4.2.5, the heat flux for the center model is scaled
by 0.8 prior to 1,000 years and by 0.6 after 1,000 years, to provide the thermal input for the edge model.
As expected, the temperature history for the edge model falls below, and the relative-humidity history lies
above, the response for the center model.
1.4.3 WASTE PACKAGE DEGRADATION MODEL
Evaluation of Inventory Modules I and 2 for this EIS diverged from the Proposed Action, or base case,
inventory evaluated in the Viability Assessment. Extensions of the waste package degradation modeling
supporting the total systems performance assessment model were required to evaluate the additional
inventories. These extensions are detailed in this section.
One component of the EIS and Total System Performance Assessment - Viability Assessment
performance assessments pertains to quantifying the degradation of the metallic waste packages. A waste
package would be a double-walled disposal container consisting of an outer 10-centimeter (4-inch)-thick
layer of carbon steel (the corrosion-allowance material), and an inner 2-centimeter (0.8-inch)-thick layer
of chromium-molybdenum Alloy-22 (the corrosion-resistant material) (DOE 1998a, Volume 3, page
3-74). A statistically based waste package degradation numerical code, WAPDEG (TRW 19981, all), was
developed to quantify the ranges in expected degradation of the waste packages. The corrosion rates for
the corrosion-allowance materials and corrosion-resistant materials included in the code were abstracted
from several sources (TRW 1998e, pages 5-1 1 to 5-16). The development of WAPDEG indicated that the
major environmental factors in waste package degradation were temperature and moisture availability.
These data were input into WAPDEG after conducting thermal-hydrologic modeling to establish the
temperature and relative humidity histories, as described in Section 1.4.2.
1.4.3.1 WAPDEG Development and Application to Total System Performance
Assessment - Viability Assessment
The EIS WAPDEG calculations were based on the Total System Performance Assessment - Viability
Assessment model configuration of this code (TRW 1998e, page 5-3). The performance assessment
analysis conducted for the Total System Performance Assessment - Viability Assessment considered a
repository thermal load of 85 MTHM per acre, with the base case waste inventory of 63,(X)0 MTHM
commercial spent nuclear fuel and 7,000 MTHM DOE spent nuclear fuel and high-level radioactive
waste. Numerical thermal-hydrologic modeling was conducted to generate transient temperature and
relative humidity histories within the emplacement drift. These histories were then used as input into the
WAPDEG code to determine the time of initiation, type, and rate of waste package corrosion during a
100,000-year simulation. The WAPDEG simulations generated a suite of waste package failure
distributions that were incorporated into the Total System Performance Assessment - Viability
Assessment model.
Two corrosion modes were implemented by the WAPDEG code for each waste package, general
corrosion and localized corrosion. These modes were applicable to both the corrosion-allowance-material
outer wall/barrier and the corrosion-resistant-material inner wall/barrier. The conditions under which the
1-34
Environmental Consequences of Long-Term Repository Performance
corrosion modes applied in WAPDEG depended primarily on temperature, relative humidity, the
geochemistry of the water, and the presence or absence of dripping or pooled water.
The corrosion-allowance material undergoes general corrosion according to one of two models, a humid-
air corrosion model and an aqueous corrosion model, depending on the relative humidity at the waste
package surface. Both models are based on statistical analysis of corrosion data observed for carbon-steel
corrosion (DOE 1998a, Volume 3, pages 3-81 to 3-82). However, neither corrosion model will be
applicable if the temperature at the waste package surface is too high. The thermal calculations for the
potential repository typically show an initial postclosure increase in repository temperature due to
radioactive decay, followed by a cooling period that eventually reaches ambient temperature. Laboratory
and modeling studies indicate that general corrosion of the corrosion-allowance material can only start
when the temperature cools to a value near the boiling point of water (DOE 1998a, Volume 3, page 3-82).
The temperature-dependent corrosion data are input into the model and applied to waste packages based
on a user-defmed temperature threshold either in the form of a fixed value or a probability distribution
that is sampled for each package.
Relative humidity generally increases as the temperature cools and vaporized moisture condenses. If the
relative humidity is sufficiently high and the temperature threshold is met, the corrosion-allowance
material can undergo humid-air corrosion. An input to the model is the relative humidity threshold
sufficient for initiation of humid-air general corrosion either as a fixed value or a probability distribution
that is sampled for each package.
The relative humidity may rise sufficiently to cause a thin film of water to form on the waste package
surface. At that point, the aqueous corrosion model more appropriately describes general corrosion. The
relative humidity threshold is input either as a fixed value or a probability distribution that is sampled for
each package. When the relative humidity exceeds the threshold, WAPDEG transitions from the humid-
air corrosion model to the aqueous corrosion model.
Neither general corrosion model for corrosion-allowance materials is expected to behave in a uniform
manner over the entire waste package surface. WAPDEG includes a provision for nonuniform corrosion
in two ways; it discretizes the waste package surface into segments called patches with roughness factors
applied to each patch. The number of patches per waste package and the roughness factors are input, with
the latter either as a fixed value or a probability distribution. WAPDEG obtains a statistical sample of the
distribution (if provided) to be used for each patch on the package. The product of the general corrosion
depth at a given time and the roughness factor gives the total corroded depth at a particular location on the
patch at that time. When the corroded depth at any point on a patch equals or exceeds the thickness of the
corrosion-allowance material, WAPDEG assumes that the patch has failed.
When a patch is breached on the corrosion-allowance material, WAPDEG assumes that part of the surface
area of the corrosion-resistant material is then subject to corrosion. In fact, there is a one-to-one
correspondence of patches for corrosion-allowance material and corrosion-resistant material. Even
though only a fraction of the corrosion-allowance material patch may be breached, the crevice between
the two materials will likely grow over time to allow water and air to access the entire corrosion-resistant
material patch. WAPDEG conservatively assumes that the entire area of this patch is immediately subject
to corrosion upon breach of its overlying corrosion-allowance material patch.
The general corrosion of the two materials differs due to the composition of the two waste package wall
materials. The general corrosion rate applied by WAPDEG to the corrosion-resistant material was
derived from data gained from the Waste Package Degradation Expert Elicitation. A compilation of the
elicited results was then used to create a cumulative distribution function for general corrosion rates of
corrosion-resistant materials at temperatures of 25°C, 50°C, and 1(X)°C (77°F, 122°F, and 212°F,
respectively) (DOE 1998a, Volume 3, pages 3-85 to 3-88). WAPDEG samples a corrosion rate from each
cumulative distribution function for a package in such a manner that, if the points were joined on a plot
1-35
Environmental Consequences of Long-Term Repository Performance
comparing corrosion rates and temperatures, the curve for a waste package is parallel to the curves for all
the other waste packages. When WAPDEG encounters a temperature between the specified temperatures,
it linearly interpolates the logarithm of the corrosion rate versus the reciprocal of the temperature to
estimate the corrosion rate at the given temperature.
According to a follow-up question for the Waste Package Degradation Expert Elicitation, the spread of
the general corrosion rates at a given temperature was due to a combination of uncertainty and natural
variability. Waste Package Degradation Expert Elicitation panelists estimated the Alloy-22 general
corrosion rate and the allocation of the total variance to its variability and uncertainty. The effect of the
corrosion rate variability among waste packages, patches, and the corrosion rate uncertainty on waste
package failure and, ultimately, radiological dose was evaluated by splitting the total variance into three
different variability and uncertainty combinations: 75-percent variability and 25-percent uncertainty;
50-percent variability and 50-percent uncertainty; and 25-percent variability and 75-percent uncertainty.
Uncertainty was interpreted as the uncertainty of the mean of the distribution. To capture this uncertainty,
a given percentage was used to establish three possible values for the mean which were based on the 5th,
50th, and 95th percentiles of the uncertainty about the global mean. Three uncertainty splits, combined
with these three estimates of the mean, produced nine new cumulative distribution functions for general
corrosion rate, which implied nine WAPDEG runs. These runs are summarized in Table 1-29.
Table 1-29. Uncertainty/variability splitting sets for corrosion rate
of corrosion-resistant material."
Uncertainty/variability splitting
ratios
Percentile
25% and 75%
50% and 50%
75% and 25%
5th
50th
95th
Setl
Set 4
Set 7
Set 2
Sets
Sets
Set 3
Set 6
Set 9
a. Source: TRW (1999a, Table 3.3-1, page 3-12).
In the presence of water or water vapor, localized corrosion could occur on the corrosion-resistant
material in the form of pitting or crevice corrosion. Information from the Waste Package Degradation
Expert Elicitation indicates that localized corrosion would begin only if the temperature was sufficiently
high. The user supplies the temperature threshold for initiating pitting either in the form of a fixed value
or a probability distribution that is sampled for each waste package. If pitting is allowed to begin as the
result of sufficient water and heat levels, WAPDEG implements an Arhennius model for pit growth.
Thus, the corrosion-resistant material could be breached either by the general corrosion of patches on the
waste package surface or by pit penetration. WAPDEG output files indicate the number of patch failures
and pit penetrations over time for each waste package.
The local environment in the waste-emplacement areas could differ from package to package, a factor
treated as variability in WAPDEG. To implement this concept, WAPDEG assumes that the variances of
the probability distributions that describe general corrosion are due to spatial variability and the variances
should be allocated. Using the treatment described above for splitting the cumulative distribution
functions for general corrosion of the corrosion-resistant material, the variance of each of the resulting
nine distributions is due to natural variability. Some variance accounts for package-to-package
variability, and the rest accounts for variable conditions along a waste package (patch-to-patch
variability). The user supplies the fraction of variance to be shared by the waste packages, and the
remaining fraction is applied to patches. In the Viability Assessment analysis, variance between packages
and between patches is 35 percent/65 percent for patches dripped on and 50 percent/50 percent otherwise.
In practice, WAPDEG samples a corrosion parameter using the global distribution but with only a
fraction of its variance. The sampled value is then treated as the mean value for the patches on that waste
package. For each patch, WAPDEG samples the distribution using the waste package mean and the
remaining variance. The results are used to model general corrosion for the patch. WAPDEG also
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Environmental Consequences of Long-Term Repository Performance
applies this variance-sharing technique to the general corrosion of the corrosion-allowance material and to
the temperature threshold for pitting initiation on the corrosion-resistant material.
One difference between waste package environments would be the presence or absence of dripping or
pooled water. WAPDEG allows the user to specify the fraction of patches that contact such water, either
as a fixed value or using a probability distribution. The user can also specify when drips start, stop, or
experience a change in water chemistry. For dripping conditions, model inputs can be used to specify
roughness factors on the corrosion-allowance material, the cumulative distribution functions of general
corrosion rates for corrosion-resistant material, and all the temperature and relative humidity thresholds as
different from those for nondripping conditions. WAPDEG determines if an individual patch is dripped
on or not and uses the appropriate model parameters.
For the Total System Performance Assessment - Viability Assessment configuration, waste package
failure distributions were generated based on always-dripping or no-dripping conditions. For each
infiltration (I) case where I varied from I multiplied by 3 to I divided by 3 (I, I x 3, and I - 3), nine
simulations were conducted based on the always-dripping corrosion rates. Because of the small number
of failures for the no-dripping case, only one case was simulated (Set 6).
1.4.3.2 Application of WAPDEG for the EIS
This EIS analyzes the effects of three different thermal loads (high, intermediate, and low) and three
waste inventories (Proposed Action, Inventory Module 1, and Inventory Module 2) to determine their
impact, if any, on total system performance. The comparison of thermal output versus time for the
Inventory Module I and Inventory Module 2 waste inventories were considered identical for the thermal-
hydrologic modeling (see Section 1.4.2). Therefore, only the Proposed Action inventory and Inventory
Module 1 (the expanded inventory) were considered.
Section 1.4.2 describes the number of repository regions that were simulated depending on the thermal
load requirements for each scenario. To incorporate the potential cooling effects around the edges of a
repository region, some regions were simulated using a conceptualized center and edge, resulting in
multiple NUFT simulations for certain regions. Table 1-30 lists the number of individual simulations
conducted for each thermal load/inventory combination, for each climate scenario.
As with the Total System Performance Assessment-Viability Assessment analyses, only the long-term
average climate scenario was used in the EIS WAPDEG simulations. Therefore, the six thermal-
hydrologic scenarios listed in Table 1-30 were used in the generation of an equal suite of WAPDEG
simulations that assumed long-term average infiltration conditions. Table 1-30 lists 18 total individual
thermal-hydrologic simulations for the six scenarios. WAPDEG simulations were performed using the
temperature and relative humidity histories generated from each of the 18 simulations. Each set of
WAPDEG simulations consisted of nine always-dripping and one no-dripping case, based on
uncertainty/variability splitting.
The EIS analyses used one always-dripping case and the no-dripping base case input files from the Total
System Performance Assessment - Viability Assessment as starting points. The EIS models used the
same corrosion model configuration and the same corrosion rate probability distribution functions as
those used in the Total System Performance Assessment - Viability Assessment base case configuration.
However, the EIS analysis used a lower, fixed relative humidity threshold for corrosion initiation of the
corrosion-resistant material than that used in the Total System Performance Assessment - Viability
Assessment analysis. The threshold used in the EIS analysis is based on a better understanding of the
factors that initiate corrosion. This difference resulted in an earlier estimate of failure of the corrosion-
resistant material for the EIS analysis. This earlier failure is evident in the results of the 10,000-year
analysis but does not affect the 1 -million-year analysis.
1-37
Environmental Consequences of Long-Term Repository Performance
Table 1-30.
Thermal-hydrologic
and waste
package degradation simulation matrix."
Thermal -
hydrology
scenario
Inventory
module
Thermal load Repository
(MTHM per acre) block(s)
Stratigraphic
column
Block
simulation
location
WAPDEG
simulation
number
1
Proposed Action
85
Upper Block
1
Center
Edge
1-10
11-20
2
Proposed Action
60
Upper Block
1
Center
Edge
21-30
31-40
3
Proposed Action
25
Upper Block
Lower Block
Block 5
1
2
7
Center
Center
Center
41-50
51-60
61-70
4
Inventory
Modules 1 and 2
85
Upper Block
Lower Block
1
2
Center
Edge
Center
Edge
71-80
81-90
91-100
101-110 J
5
Inventory
Modules 1 and 2
60
Upper Block
Lower Block
Block la
1
2
3
Center
Center
Center
111-120
121-130
131-140 ^
6
Inventory
Modules 1 and 2
25
Upper Block
Lower Block
Block 8
Block 5
1
2
4
7
Center
Center
Center
Center
141-150
151-160
161-170
171-180
a. Source: TRW (1999a, Table 3.3-2, page 3-13).
Each WAPDEG run generated a failure curve that contained a probability distribution function of the
first corrosion-resistance-material breach, average pit failures, and average patch failures (as a function i
time). These files were transferred to the repository integration program model.
1.4.3.3 Results
Figure 1-15 shows the temperature and drift relative humidity history curves, respectively, for all three
thermal loads (high, intermediate, and low) with the Proposed Action inventory. Figure 1-16 shows the
temperature and relative humidity history curves, respectively, for all three thermal load scenarios with
the expanded inventory (Inventory Modules 1 and 2). These figures show that when the temperature
threshold [100°C (212°F)] for corrosion initiation is met, the relative humidity within the drifts for most]
of the runs is within the range of aqueous corrosion (80 to 100 percent). The time to reach the
temperature threshold is less for the low thermal load scenario (less than 100 years) than for the high and
intermediate thermal load scenarios (200 to 700 years). Corrosion of the corrosion-allowance material foT"
the low thermal load scenario is initiated sooner but only by a few hundred years. This difference will
become relatively small when discussing the differences in package failure rates at times greater than
10,000 years.
The thermal histories generated from the thermal-hydrologic modeling indicate that the hottest and
coolest thermal histories correspond to the high thermal load, expanded-inventory scenario and the low
thermal load, Proposed Action inventory scenarios, respectively. Thus, the results from these two
configurations bound the range of potential WAPDEG failure responses. In addition, the waste package
failure results were dominated by the packages that were dripped on; therefore, the failure results for the
packages that were not dripped on are not presented.
1-38
A
Environmental Consequences of Long-Term Repository Performance
WAPDEG simulations for the low thermal load with Proposed Action inventory case were generated for
three repository regions corresponding to the upper (primary) block, lower block, and Block 5. The
thermal output for this layout did not include edge effects (see Section 1.4.2); therefore, only one thermal
simulation per repository block was generated. Temperature and relative humidity histories generated
from each repository block were used to define the conditions within the drifts. Figure 1-17 shows the
time to first breach or failure of the corrosion-allowance material for the always-dripping packages in
each of the three emplacement blocks. The failures of the corrosion-allowance material are very similar
for all three stratigraphic columns, with failures starting at approximately 800 years and extending
approximately 4,000 years. Figure 1-18 shows the time to first breach of the corrosion-resistant material
for the always-dripping packages in each of the three emplacement blocks, for each of the nine
uncertainty/variability splitting sets (defined in Table 1-29). The failure of the corrosion-resistant material
barriers in the three regions were very similar, given the same uncertainty/variability splitting set (set 5).
For example, the responses observed for stratigraphic columns 2 and 7 overlie each other. The variability
in the failure of the corrosion-resistant material in a particular region (for example, stratigraphic
column 1), due to the introduction of the uncertainty/variability splitting, ranges from a few thousand
years (set 7) to no failures within 1 million years (set 3).
Given the relatively cool thermal history for the low thermal load scenario and the 70,000 MTHM
inventory, no pits (localized corrosion) would penetrate through the corrosion-resistant material for the
always-dripping packages for all three emplacement blocks. All failures (see Figure 1-18) would be due
to general corrosion because the temperature threshold for localized corrosion was not reached. Figure
1-19 shows the average number of patches penetrated through the corrosion-resistant material as a
function of time for the always-dripping packages, all three emplacement blocks, and the
uncertainty/variability splitting sets. Figures 1-20 through 1-22 show that the variability in the results of
the failure for the three emplacement blocks is dominated by the corrosion rate uncertainty/variability
splitting of corrosion-resistant material, with little variability attributed to the different thermal-hydrologic
inputs.
WAPDEG simulations for the high thermal load scenario with the expanded inventory were generated for
the upper (primary) and lower repository blocks. The repository blocks were simulated with both a center
and an edge region (see Section 1.2). Figure 1-20 shows the time to first breach or failure of the corrosion-
allowance material for the always-dripping packages, for all four simulations, and the
uncertainty/variability splitting sets. Figure 1-21 shows the time to first breach of the corrosion-resistant
material. Figure 1-22 shows the average number of patches penetrated through the corrosion-resistant
material as a function of time. Previous analyses have shown that the releases from the waste packages
are dominated by advection through the patch. Therefore, the patch failure history is a representative
indicator of the overall performance. The results shown in Figure 1-22 also show that the variability in
the failures for the four center and edge simulations is dominated by the uncertainty/variability splitting,
with little variability attributed to the different thermal-hydrologic inputs.
These results show that the variability in the corrosion-resistant material failures as a function of time has
a greater dependency on the variability/uncertainty splitting associated with the corrosion-resistant
material corrosion rate than on the variation in the temperature and relative humidity histories. The
results for the high and intermediate thermal load scenarios for the Proposed Action inventory and the
intermediate and low thermal load scenarios for the expanded inventory simulations showed similar
behavior to the results discussed above.
1.4.3.4 Discussion
Corrosion of the corrosion-allowance material is not initiated until the waste package temperature
decreases below the thermal threshold selected for the model [100°C (212°F)]. For the majority of the
thermal-hydrologic simulations conducted for the EIS, once the thermal threshold is satisfied, the humid-
air corrosion is initiated. Figure 1-23 shows the time to the first breach of the corrosion-allowance
1-39
Environmental Consequences of Long-Term Repository Performance
material for all expected-value always-dripping WAPDEG simulations (Set 5). The time to first breach of
the corrosion-allowance material is earliest for the low thermal load scenarios as expected from the
temperature profiles shown in Figures 1-15 and 1-16. Because the thermal threshold is satisfied sooner,
corrosion of the corrosion-allowance material is initiated sooner.
Figure 1-23 also shows that by 5,000 years, almost each waste package has had at least a single corrosion-
allowance material failure, thereby allowing corrosion of corrosion-resistant material. Figure 1-24 shows
the time to the first breach of the corrosion-resistant material for all expected-value always-dripping
WAPDEG simulations (Set 5). The first corrosion-resistant-material breach for most scenarios occurs
between 20,(XX) to 30,(KX) years, with the high thermal load, expanded-inventory scenario having a very
low fraction of packages failing within 10,000 years. Figure 1-24 also shows that the higher thermal loads
generate the earliest corrosion-resistant material failures, even with later corrosion-allowance material
failures. This behavior is due to the temperature-dependent, corrosion-resistant-material corrosion
models, which have higher corrosion rates at higher temperatures. The thermal profiles in Figures 1-23
and 1-24 show that temperature is lower for the lower thermal load scenarios, resulting in slower
corrosion rates and delayed failure relative to the higher loads.
Figure 1-25 shows the average number of patches that failed per package as a function of time for all
thermal loads and inventories, all regions, always-dripping, and uncertainty/variability splitting (set 9).
Figure 1-26 shows the average number of patches that failed per package as a function of time for all
thermal loads and inventories, all regions, always-dripping, and uncertainty/variability splitting (set 5).
These plots show a factor-of-five difference between the failure results for the two different
uncertainty/variability-splitting sets.
The degradation results show that for each thermal-hydrologic scenario, the variability in the failures due
to the uncertainty/variability splitting in the corrosion rate of the corrosion-resistant material would be
considerably greater than the variability due to the different thermal histories. Therefore, for each thermal
and inventory scenario, a set of -failure distributions from a single region was selected and included in the
RIP model simulations.
1.4.4 WASTE FORM DISSOLUTION MODELS
Evaluation of Inventory Modules 1 and 2 for this EIS diverged from the Proposed Action, or base case,
inventory evaluated in the Viability Assessment. Specifically, additional waste forms were included in
Inventory Modules 1 and 2 that were not considered in the Viability Assessment base case, and waste
form dissolution models were required to model these additional waste forms. Extensions of the waste
form dissolution modeling that supported the Total Systems Performance Assessment model were
required to evaluate the additional inventories. These extensions are detailed in this section.
1.4.4.1 Spent-Fuel Dissolution Model
A semi-empirical model for intrinsic dissolution (alteration) rate of the spent fuel matrix was developed
from experimental data (TRW 1995, page 6-2). If the postclosure environment inside the potential
repository can be assumed to maintain the atmospheric oxygen partial pressure of 0.2 atmosphere (TRW
1995, page 6-1), the dissolution model becomes a function of temperature, total carbonate concentration,
and pH of contacting water. The dissolution rate strongly depends on temperature and total carbonate
concentration but is less influenced by pH. The spent fuel dissolution rate increases with temperature and
is enhanced by the total carbonate concentration of the contacting water, although to a smaller extent than
by temperature. The mixed oxide spent nuclear fuel from plutonium disposition was modeled as
commercial spent nuclear fuel.
i
1-40
Environmental Consequences of Long-Term Repository Performance
1.4.4.2 High-Level Radioactive Waste Glass
As in the spent fuel alteration/dissolution modeling discussed above, the entire surface area of defense
high-level radioactive waste in the glass waste form is assumed to be exposed to the near-field
environment as soon as the first pit penetrates the waste package. The waste forms are assumed to be
covered by a "thin" water film when the water contacts the glass and the alteration/dissolution processes
are initiated. The "can-in-canister" ceramic from plutonium disposition was modeled as high-level
radioactive waste.
High-Level Radioactive Waste Glass Dissolution Model
Details concerning the intrinsic glass dissolution rate model, as a function of temperature and pH, are
presented along with rate data (TRW 1995, pages 6-4, 6-5, and 6-37). The relationship indicates that the
rate model represented by the equation predicts a monotonically increasing dissolution rate with
temperature.
This dissolution conceptualization contains several assumptions and limitations. The radionuclides are
assumed to be released as fast as the glass structure breaks down, which is a conservative assumption
because it does not account for solubility-limited radionuclides. No credit is taken for the fact that
"experiments have shown that the actinides more commonly are included in alteration phases at the
surface of the glass either as minor components of other phases or as phases made up predominantly of
actinides" (TRW 1995, page 6-5). The model includes neither solution chemistry (other than pH and
dissolved-silica concentration) nor vapor-phase alteration of the glass. Glass has been observed to
undergo hydration in a humid environment and, on subsequent contact with water, radionuclide releases
from a hydrated glass layer were several orders of magnitude higher than those from an unhydrated
(firesh) glass waste form (TRW 1995, page 6-5).
1.4.4.3 Greater-Than-Ciass-C and Special-Performance-Assessment-Required Waste
The alteration/dissolution processes for Greater-Than-Class-C and Special-Performance-Assessment-
Required waste forms were assumed to be similar to those for high-level radioactive waste glass.
1.4.5 RIP MODEL MODIFICATIONS
The EIS RIP model simulations are based on the Total System Performance Assessment - Viability
Assessment (Revision 1) base case RIP model (TRW 1998n, all). To perform the EIS performance
assessment analyses, the base case model was modified primarily to allow input of the different repository
areas corresponding to the thermal load scenarios and the expanded waste inventories of Modules I and 2,
and the repository-block configurations used in the thermal-hydrologic modeling. The EIS analysis also
considered the impact to individuals at distances other than the 20 kilometers (12 miles) used for the
Viability Assessment. Therefore, the analysis expanded the saturated-zone convolution model used in the
Viability Assessment to include development of convolution stream tubes from the repository to distances
of 30 kilometers (19 miles) and 80 kilometers (50 miles) and postprocessing of the 20-kilometer output to
extract the radiological dose to individuals at the 5-kilometer (3-mile) distance described in Section
1.4.5.4. This section describes the modifications. Knowledge and understanding of the RIP model
(Colder 1998, all) and the Viability Assessment model (TRW 1998a,b,c,d,e,f,g,h,i,j,k, all) are necessary
to fully understand the differences discussed in this section.
1.4.5.1 Modifications to the RIP Model in the Repository Environment
The RIP model conceptualization for the Yucca Mountain Repository performance assessment considers
waste forms in discrete regions of the repository as source terms for flow and transport. The RIP model
conceptualization for the Viability Assessment considered the primary repository block, corresponding to
the high thermal load scenario, to be comprised of six regions. For any particular case analyzed for the
1-41
Environmental Consequences of Long-Term Repository Performance
EIS, the EIS thermohydrologic simulations were used to determine the number of repository regions used.
In adapting the Viability Assessment base case as the model for the EIS analyses, the repository regions
had to conform to the center/edge model conceptualization. For each of the unused Viability Assessment
regions, the source terms (commercial spent nuclear fuel, high-level radioactive waste, and DOE spent
nuclear fuel) and all associated RIP model cells were removed from the model, and the remaining source
terms and associated connecting cells were adapted to the center/edge model. In all cases, a total of
60 concentration parameters and all of the "connection" groups, except the 10 groups that provided total
radiological dose at various points, were removed from the model. Then, the new region-specific
connection groups were added as appropriate to account for the calculation of advective and diffusive
releases from the center and edge regions of the EIS simulations. The calculated flux data, developed
from the Lawrence Berkeley National Laboratory hydrologic model of the repository area (Bodvarsson,
Bandurraga, and Wu 1997, all), was used to modify the flux into and fluid saturations applicable to the
various source terms in the EIS RIP model.
Another modification resulted from the fact that although the Total System Performance Assessment -
Viability Assessment considered sensitivity variations in the infiltration to the repository, the EIS
simulations used only the infiltration (I) option. This was done to reduce the number of calculations,
because the three thermal loads and two extra inventories greatly multiplied the number of cases to be
simulated. The (I x 3) and (I x 3) options of the Total System Performance Assessment - Viability
Assessment were not considered. Therefore, only the WAPDEG results for the "always-dripping" and
"no-drip" scenarios were selected for model input. This change resulted in appropriate changes to the
fraction-of-packages-failed parameters to allow the appropriate (I) WAPDEG to be incorporated into the
model. To accommodate these differences to the RIP model, the fraction-of-packages-failed parameters
for the (I X 3) and (I x 3) options were redirected to call the applicable WAPDEG tables for the long-term
average climate case. The effect of neglecting this variation is minor. Sensitivity studies with the
ViabiUty Assessment model for the high thermal load scenario (DOE 1998a, Volume 3, pages 5-3 to 5-5)
showed that the 10,000-year peak dose is actually decreased by 30 percent for the I x 3 case, while the
peak is moved back from 10,000 years to about 5,000 years and the 1 -million-year peak dose is increased
about 30 percent.
The EIS simulations used only one thermal table rather than the six used in the Viability Assessment base
case. Therefore, the thermal parameters were updated to refer to only one unique thermal table for each
of the thermal load scenarios and inventory combinations:
• High thermal load. Proposed Action inventory
• Intermediate thermal load. Proposed Action inventory
• Low thermal load. Proposed Action inventory
• High thermal load. Inventory Modules 1 and 2
• Intermediate thermal load. Inventory Modules 1 and 2
• Low thermal load. Inventory Modules 1 and 2
The thermal hydrology modeling indicated that a single invert saturation was sufficient for all regions and
all layers of the invert. Based on this information, all invert saturation parameters were fixed to a value of
0.993.
1.4.5.2 Modifications to Input and Output FEHIVI Model
The particle-tracking files used in the Viability Assessment (TRW 1998g, all) were modified for each EIS
case to allow a different number of FEHM input regions to be used, depending on the number of input
regions used in the engineered barrier system model. The "Zone 6" interface file was modified for each
EIS case by changing the FEHM nodes to be used for input of mass from the engineered barrier system.
The FEHM nodes were chosen to correspond to the coordinates of the EIS repository emplacement
1-42
Environmental Consequences of Long-Term Repository Performance
blocks. For the low thermal load scenario for Inventory Modules 1 and 2 shown in Figure 1-7, proposed
f IpBlocks 6 and 7 fell outside the model boundaries. To allow the unsaturated zone particle tracker in the
FEHM model to account for all mass in the repository, the mass from areas 6 and 7 were allocated to
Blocks 5 and 8, respectively. Figures 1-27 through 1-32 show the repository emplacement blocks used for
each case.
The "Zone 6" interface file was also modified for each EIS case by defining the saturated zone area that
would capture the mass coming out of the FEHM model. It was necessary to modify the capture regions
in order to ensure inclusion of all of the mass and to distribute the mass amongst the six stream tubes
based on its repository emplacement block of origin. For the high and intermediate thermal load
scenarios with Proposed Action inventories, the same regions were used for this EIS as were used for the
Viability Assessment base case (Figure 1-33). Figure 1-34 shows the capture regions used for the low
thermal load scenario with the Proposed Action inventory; the low thermal load scenario with Inventory
Modules 1 and 2, and the intermediate thermal load scenario with Inventory Modules 1 and 2. Figure
1-35 shows the capture regions used for the high thermal load scenario with Inventory Modules 1 and 2.
1.4.5.3 Modifications to Saturated Zone Stream Tubes for Different Repository Areas
The saturated zone stream tubes consist of a unit-breakthrough curve and a scaling factor. The unit-
breakthrough curves are all the same for a given radionuclide at a given distance. The scaling factor is the
product of the flux coming from the repository and a dilution factor. The dilution factor is a lumped
parameter that is used to account for mixing and lateral dispersion. For the multiple-realization cases, the
dilution factor is assumed to have lognormal distribution with a mean value of ten.
In order to use the stream tubes for different repository regions, flux multiplier values were calculated for
each stream tube. The flux multiplier value is the ratio of the new flux into a stream tube to the flux into
that stream tube in the base case (Proposed Action inventory, high thermal load scenario). The saturated
zone module of RIP requires the concentration of water entering the saturated zone from the unsaturated
zone, so the water flux at this interface is needed to compute the mass concentration of contaminants in
the water. The resulting flux multiplier is used to scale the water flux predicted by the FEHM transport
module in RIP to properly account for the larger capture zone areas for other cases. Each stream tube is
associated with one of the unsaturated zone capture regions described above. The flux into a given stream
tube is the sum of the fluxes from the repository regions that are in that capture region. The high thermal
load scenario with Proposed Action-inventory used the same fluxes as the Viability Assessment base
case. Tables 1-3 1 and 1-32 list the contribution to each of the stream tubes from each of the repository
areas for the intermediate and low thermal load scenarios with Proposed Action inventory, respectively.
The same information is provided for the high, intermediate, and low thermal load scenarios with
Inventory Modules 1 and 2 inventory, respectively, in Tables 1-33 through 1-35. The fluxes used in these
tables were obtained from the results of the base case Lawrence Berkeley National Laboratory site-scale
unsaturated zone flow model (Bodvarsson, Bandurraga, and Wu 1997, all).
Table 1-31. Summary of fluxes (cubic meters per year) from repository area to convolution stream tubes
for intermediate thermal load scenario with Proposed Action inventory."
Flux from each
repository
area into each stream tube
85-MTHM-per
Upper
Lower
Blocks
Blocks
acre, base case
Stream tube
block
block
5&6
7&8
Total flux
inventory flux
Flux multiplier
1
6,410
0
0
0
6,410
3,162
2.03
2
3,480
0
0
0
3,480
3,482
LOO
3
3,990
0
0
0
3,990
3,993
LOO
4
4,060
0
0
0
4,060
4,060
1.00
5
8,090
0
0
0
8,090
10,103
0.801
6
5,320
0
0
0
5,320
2,077
2.56
a. Source: TRW (1999a, Table 3.5-1, page 3-19).
1-43
Environmental Consequences of Long-Term Repository Performance
Table 1-32. Summary of fluxes (cubic meters per year) from repository area to convolution
stream tubes for low thermal load scenario with Proposed Action inventory."
Flux from each repository area into each stream tube
Stream Upper Lower Blocks Blocks
tube block block 5&6 7&8
Total flux
85-MTHM-per
acre, base case Flux
inventory flux multiplier
1
2
3
4
5
6
16,570
16,570
0
0
0
0
0
0
5,250''
5,250"
0
0
0
0
0
0
6,750
6,750
0
16,570
3,162
5.24
0
16,570
3,482
4.76
0
5,250"
3,993
0.131
0
5,250"
4,060
0.129
0
6,750
10,103
0.668
0
6,750
2,077
3.25
a. Source: TRW (1999a, Table 3.5-2, page 3-19).
b. Typographical error in source document.
Table 1-33. Summary of fluxes (cubic meters per year) from repository area to convolution
stream tubes for high thermal load scenario with Inventory Modules 1 and 2."
Flux from each repository area into each stream tube
Stream Upper Lower Blocks Blocks
tube block block 5 & 6 7 & 8 Total flux
85-MTHM-per
acre, base case Flux
inventory flux multiplier
1
7,050
0
2
7,050
0
3
7,050
0
4
7,050
0
5
7,050
0
6
0
969
0
0
0
0
0
0
7,050
3,162
2.23
7,050
3,482
2.02
7,050
3,993
1.77
7,050
4,060
1.74
7,050
10,103
0.698
969
2,077
0.466
a. Source: TRW (1999a, Table 3.5-3, page 3-20).
Table 1-34. Summary of fluxes (cubic meters per year) from repository area to convolution
stream tubes for intermediate thermal load scenario with Inventory Modules 1 and 2."
Flux from each repository area into each stream tube
Stream Upper Lower Blocks Blocks
tube block block 5&6 7&8 Total flux
85-MTHM-per
acre, base case Flux
inventory flux multiplier
1
17,620
0
2
17,620
0
3
0
3,350
4
0
3,350
5
0
0
6
0
0
0
0
0
0
0
0
0
17,620
3,162
5.57
0
17,620
3,482
5.06
0
3,350
3,993
0.838
0
3,350
4,060
0.824
4,090
4,090
10,103
0.404
4,090
4,090
2,077
1.97
a. Source: TRW (1999a, Table 3.5-5, page 3-20).
Table 1-35, Summary of fluxes (cubic meters per year) from repository area to convolution
stream tubes for low thermal load scenario with Inventory Modules 1 and 2."
Flux from each repository
area into each stream tube
85-MTHM-per
acre, base case
Stream Upper
Lower
Blocks
Blocks
Flux
tube block
block
5&6
7&8
Total flux
inventory f
lux
multiplier
1 17,620
0
0
0
17,620
3,162
5.57
2 17,620
0
0
0
17,620
3,482
5.06
3 0
5,250
0
0
5,250
3,993
1.31
4 0
5,250
0
0
5,250
4,060
1.29
5 0
0
10,240
0
10,240
10,103
1.01
6 0
0
10,240
54,200
64,440
2,077
31.0
a. Source: TRW (1999a, Table 3.5-4, page 3-20).
1-44
Environmental Consequences of Long-Term Repository Performance
REPOSITORY SIZE AND SATURATED ZONE DILUTION FACTORS
Increasing repository size could cause either a reduction or no change in the relative lateral
dispersive effects of saturated zone transport. Consider a rectangular repository oriented normal to
the direction of flow in the saturated zone. The cross-sectional area of the resultant contaminant
plume at a downstream well would be larger than that at the cross-sectional area of the plume at the
source (below the repository), causing dilution of the radionuclide concentration at the downstream
well. However, if the area of the repository was doubled, the plume at the exposure location would
increase, but by less than twice. Hence, lower dilution factors would occur for larger repositories.
Analytical modeling provides quantification for lower dilution factors.
The validity of using lower dilution factors for larger repositories can be illustrated by considering two
hypothetical repositories with equal waste inventory, one having twice the emplacement area of the
other. The concentration at the base of the unsaturated zone below the larger repository would be -
half the concentration below the smaller repository (a direct result of different spacing of the waste).
Using a one-dimensional saturated zone transport model without dilution, for times far greater than
the groundwater travel time, the concentrations at a downstream well would be equal to those at the
base of the unsaturated zone (provided the contaminant release was continuous). If the same
dilution factor was applied in both cases, the downstream well concentrations for the larger
repository would be half those in the smaller repository. On the other hand, if the repository was
treated as a point source in each case, the dilution factor for the larger repository would be half that
of the smaller repository, resulting in equal concentrations at a downstream well. These two
outcomes correspond to two alternative ways of doubling the repository area. Thus, the dilution
factors for expanded area repositories can be lower or equal to those of the base-case repository.
1.4.5.4 Modifications to the Stream Tubes for Distances Other Than 20 Kilometers
One-dimensional stream-tube runs for the saturated zone were conducted for generating unit-
breakthrough curves at distances of 30 and 80 kilometers (19 and 50 miles) downstream from the
repository. This was accomplished using the Los Alamos National Laboratory simulator FEHM
(Zyvoloski et al. 1995, all) and developing a finite-element mesh that extended beyond the 25-kilometer
(16-mile) mesh previously used to develop the 20-kilometer (12 mile) stream tube used for the Viability
Assessment. The sets of transport parameters used in the previous model runs were also applied in the
extended mesh simulations for distances up to 25 kilometers. Beyond 25 kilometers, the model properties
were made identical to those assigned to the undifferentiated valley fill. On completing the FEHM runs
for each of nine radionuclides, model output was postprocessed to take into account mass loadings from
the unsaturated zone to each of six different stream-tube capture areas and to adjust model results for
dilution attributed to transverse dispersion. This last step involved the determination of distance-
dependent dilution factors by using dilution information previously developed from exposure
concentrations at the 20-kilometer distance. An analytical transport solution in the program 3DADE
(Leij, Scaggs, and van Genuchten 1991, all) was used to determine dispersion coefficients that resulted in
dilution factors of 10, 50, and 100 at 20 kilometers and to determine corresponding dilution factors at
distances of 30 and 80 kilometers. The resulting data indicated a logarithmic relationship between the
20-kilometer dilution factors and those occurring at the longer distances, making it possible to determine
appropriate dilution parameters used in postprocessing of the extended-distance FEHM runs.
The saturated zone transport in the Viability Assessment is essentially based on a one-dimensional
analysis that precludes lateral dispersion in the y and z directions. To simulate the realistic results of
three-dimensional transport, the results of the one-dimensional analysis are divided by a dilution factor.
Thus, the dilution factor accounts for attenuation of concentrations caused by the spread of the
contaminant plume as the result of lateral dispersion. The dilution factor approximates numerical
dispersion for the one-dimensional saturated zone model, as can be achieved using a three-dimensional
advective-dispersive numerical model. This simulates the real dilution in the system.
1-45
Environmental Consequences of Long-Term Repository Performance
The Viability Assessment dilution factors were based on the results of the Expert Elicitation Panel Project
(TRW 1998h, Section 8.2.3.2), which assigned a median value of 10, a maximum value of 100, and a
minimum value of 1.0 (no dispersion). Consideration of Inventory Modules 1 and 2 and/or the reduced
thermal load resulted in a larger-area repository than that considered in the Viability Assessment analysis.
Simplified logical models were developed to study the impact of the larger-area repository configurations
for this EIS. In general, a larger inventory at the same thermal load results in lower concentrations at the
base of the unsaturated zone (barring some exceptionally adverse infiltration conditions) because the
spacing between disposal blocks results in the additional amount of waste being spread over a larger area,
The larger size of the repository also tends to cause a reduction in the lateral dispersive effects of
saturated zone transport, implying lower dilution factors for larger repository configurations. If the
dilution factors of the Viability Assessment were to be used in this EIS, the dose rates would be predicted
(albeit erroneously) to be lower than their true values for cases with expanded repository areas.
IS
i
The dilution factors appropriate for the larger-area repository configurations were computed for the EIS
analyses. The analytical solution for the three-dimensional transport in a one-dimensional flow field
(Leij, Scaggs, and van Genuchten 1991, all) was used to relate the lateral dispersion lengths (in the y andj
directions) and the dilution factors. Considering a rectangular source oriented normally to the flow
direction, the steady-state concentrations at the locations [5, 20, 30, and 80 kilometers (3, 12, 19, and 50
miles)] were computed based on the assumed dispersion lengths described below.
The ratio between the concentration from the one-dimensional and three-dimensional analyses gives the
dilution factor, which enables a "translation" of the Saturated Zone Expert Elicitation Panel's dilution
factors to "dispersion lengths." The Panel's dilution estimates were for a 25-kilometer (16 miles) distance
and the Viability Assessment adjusted this estimate for estimates at 20 kilometers (12 miles). The
dispersion lengths so derived for the Viability Assessment are assumed to remain the same for larger
repository configurations. Using the same dispersion lengths, as implied in the Viability Assessment, the
dilution factors for the larger repository configurations were computed using the analytical solution. The
Darcy flux used in the calculations for the saturated zone flow fields was the same 0.6 meters (2 feet) per
year used in the Viability Assessment (DOE 1998a, Volume 3, page 3-138). The actual repository
geometry was a rectangular source with an area equivalent to that of the repository configuration for the
appropriate thermal load. The larger dimension of the rectangular source was normal to the flow
direction and assumed equal in the unsaturated and saturated zones. The smaller dimension of the
rectangular source, parallel to the flow in the saturated zone, was modified in the saturated zone to fulfill
the continuity of flow requirement (that is, to reconcile large differences in the flow velocities in the
unsaturated and saturated zones).
The matrix of dilution factors (given in Table 1-36), calculated using the 3DADE computer code (Leij,
Scaggs, and van Genuchten 1991, all), was dependent on the major influences on the calculated dilution
factors, namely:
• The orientation of each repository configuration relative to the direction of groundwater flow
• The total area of each repository configuration
• The average percolation flux of each sector (or block) of the repository based on the Lawrence
Berkeley National Laboratory hydrologic model
Extension of the repository area in a direction orthogonal to that of groundwater flow had little effect on
the calculated dilution factor. However, for dilution factors calculated for the repository and enlarged in
the direction parallel to that of groundwater flow, there were changes on the order of factors of two or
three. Thus, the intermediate thermal load scenario had the same dilution factor as the high thermal load
Proposed Action scenario for the 20-kilometer (12-mile) distance, because the repository shape was
relatively similar with essentially no changes parallel to the flow direction. In contrast, the low thermal
1-46
Environmental Consequences of Long-Term Repository Performance
Table 1-36. Dilution factors for three thermal load scenarios and four exposure locations.'
Proposed Action
Inventory Modules 1 and 2
Thermal load High Intermediate Low High Intermediate Low
(MTHM per acre)" (85) (60) (25) (85) (60) (25)
Distance
Repository area (acres)
740
1,050
2,520
1,240
1,750
4,200
5 kilometers'^
Minimum
1.0
1.0
1.0
1.0
1.0
1.0
Median
5.15
5.15
2.9
5.15
3.8
2.5
Maximum
50.02
50.02
24.6
50.02
354
19.2
20 kilometers
Minimum
1.0
1.0
1.0
1.0
1.0
1.0
Median
10.0
10.0
5.1
10.0
7.2
4.1
Maximum
100.0
100.0
49.2
100.0
70.8
38.4
30 kilometers
Minimum
1.0
1.0
1.0
1.0
1.0
1.0
Median
12.2
12.2
6.2
12.2
8.8
4.9
Maximum
122.0
122.0
60.2
122
86.7
47
80 kilometers
Minimum
1.0
1.0
1.0
1.0
1.0
1.0
Median
19.894
19.84
9.9
19.84
14.2
7.8
Maximum
200.04
200.04
98.4
200.04
141.6
76.7
a. Source: TRW (1999a, Table 4.1-1, page 4-6).
b. To convert acres to square miles, multiply by 0.0015625.
c. To convert kilometers to miles, multiply by 0.62137.
load Proposed Action scenario has almost double the area of the intermediate thermal load Proposed
Action scenario. The repository is approximately twice the distance in the direction parallel to flow,
resulting in a dilution factor almost twice that of the intermediate thermal load Proposed Action scenario.
Thus, because of the repository geometry, the differences in the dilution factors between the low and
intermediate thermal load Proposed Action scenarios resulted in less dilution in the low thermal load
Proposed Action scenario.
1.4.5.5 Modifications to the RIP Model to Account for Unsaturated Zone and Saturated
Zone Particle Transport
Transport through the unsaturated zone is modeled in RIP using particles that are assigned a "start
location" at the level of the repository. The Viability Assessment analysis considered particle releases
only in the upper block of the repository. For the EIS analyses, the Lawrence Berkeley National
Laboratory model (Bodvarsson, Bandurraga, and Wu 1997, all) element centroids were mapped to the
outline of the upper block, and particles were released from these locations.
Because the EIS analysis considered expanded areas for the emplacement of waste, additional particle
coverage was needed to represent transport throughout the entire region of interest. This region included
the additional repository blocks for the expanded waste inventories considered in Inventory Modules 1
and 2. An orthogonal grid was mapped for each of the emplacement zones within the area covered by the
Lawrence Berkeley National Laboratory model, and this grid was use to determine the coordinates of
particle start points at the repository horizon. These coordinates were then converted to the centroid of
the nearest Lawrence Berkeley National Laboratory model elements. In this way, a file containing
Lawrence Berkeley National Laboratory element numbers was created for each waste emplacement zone
for the particle-start coordinates. From this functional area of the RIP model, both the EIS and Viability
Assessment performance assessment analyses used the FEHM model (Zyvoloski et al. 1995, all) to model
particle transport through the unsaturated zone.
At the base of the unsaturated zone, a corresponding change of coordinates was used to collect and
distribute the mass transported through the unsaturated zone to the saturated zone convolution stream
tubes that carried dissolved radionuclides to the various exposure locations. The unsaturated and
saturated zone capture regions for the EIS analysis were scaled-up modifications of the six regions used
1-47
Environmental Consequences of Long-Term Repository Performance
by the Total System Performance Assessment - Viability Assessment analysis, as extended to the edge of
the Lawrence Berkeley National Laboratory model area. The nodes at the bottom of the unsaturated zone
were calculated to ensure complete capture of the mass coming out of the unsaturated zone and to
appropriately distribute that mass among the six stream tubes, based on those six repository regions being
modified and applied to the expanded areas addressed by the EIS analysis.
Table 1-37 lists the ranges of stochastic parameters that were included in the analysis of saturated zone
flow and transport.
Table 1-37. Stochastic parameters for saturated zone flow and transport."
Parameter
Effective porosity, alluvium
Effective porosity, upper volcanic aquifer
Effective porosity, middle volcanic aquifer
Effective porosity, middle volcanic confining unit
Effective porosity [plutonium], volcanic units
Distribution coefficient Kj (milliliters per gram) for:
Neptunium (alluvium)
Neptunium (volcanic units)
Protactinium (alluvium)
Protactinium (volcanic units)
Selenium (alluvium)
Selenium (volcanic units)
Uranium (alluvium)
Uranium (volcanic units)
Plutonium (all units)
Longitudinal dispersivity, all units (meters)
Distribution type
Distribution statistics [bounds]
Truncated normal
Log triangular
Log triangular
Log triangular
Log uniform
Uniform
Beta (approx. exp.)
Uniform
Uniform
Uniform
Beta (approx. exp.)
Uniform
Uniform
Log uniform
Log-normal
Mean = 0.25, SD" = 0.075 [0, 1.0]
[1x10^0.02,0.16]
[1x10^0.02,0.23]
[lxlO"\ 0.02, 0.30]
[lxlO"^ IxlO"']
[5, 15]
Mean= 1.5, SD= 1.3 [0, 15]
[0, 550]
[0, 100]
[0, 150]
Mean = 2.0, SD= 1.7, [0, 15]
[5, 15]
[0,4.]
[1 X 10 "^ 10]
Log(mean) = 2.0, log(SD) = 0.753
Fraction of flow path in alluvium
Discrete CDF^
a. Source: DOE (1998a, Volume 3, Table 3-20, page 3-140).
b. SD = standard deviation.
c. CDF = cumulative distribution function.
[0, 0.3] (see text)
1.4.5.6 Biosphere Dose Conversion Factors for Waterborne Radionuclides
A biosphere dose conversion factor for groundwater is a number used to convert the annual average
concentration of a radionuclide in the groundwater to an annual radiological dose for humans. The
calculation of a biosphere dose conversion factor requires knowledge about the pathway the radionuclide
would follow from the well to humans and the lifestyle and eating habits of humans. Figure 1-36
illustrates the biosphere modeling components.
The approach used in this long-term performance assessment calculated the health consequences for a
reference person living in the Amargosa Valley. The reference person would be an adult who lived year-
round on a farm in the Amargosa Valley, grew a garden, raised livestock, and ate locally grown food.
Because future human technologies, lifestyles, and activities are inherently unpredictable, the analysis
assumed that the future inhabitants of the region would be similar to present-day inhabitants. This
assumption has been accepted in similar international efforts at biosphere modeling and is preferable to
developing a model for a future society (National Research Council 1995, all).
A lifestyle survey of people living in the area was completed in 1997 (TRW 19981, Section 9.4, pages
9-25 to 9-35). Among other functions, the survey was intended to give an accurate representation of
dietary patterns and lifestyle characteristics of residents within 80 kilometers (50 miles) of the Yucca
i
1-48
Environmental Consequences of Long-Term Repository Performance
Mountain site. Of special interest was the proportion of locally grown foodstuff consumed by local
residents and details about regularly consumed food types.
The Amargosa Valley region is primarily rural agrarian in nature and the local vegetation is primarily
desert scrub and grasses. Agriculture consists mainly of growing livestock feed (for example, alfalfa);
however, gardening and animal husbandry are common. Water for household uses, agriculture,
horticulture, and animal husbandry is primarily from local wells.
Another component of the dose to people would be the inadvertent ingestion of contaminated soil, usually
from vegetables. The inhalation pathways would include breathing small soil particles that became
airborne during outdoor activities, especially farming, mining, and construction activities that would
disturb the soil or bedrock. Proximity to a radiation source external to the body would result in an
external pathway. This pathway is called "groundshine" when the contaminants are on the ground,
"submersion" when they are in the atmosphere, and "immersion" when they are in water.
The analysis calculated biosphere dose conversion factors for the exposure pathways described above.
Although many of the input parameters were derived from site-specific data obtained from the Yucca
Mountain regional survey and weather data tabulations, some were from other published sources. The
input parameters used in the biosphere modeling are described in the Viability Assessment (DOE 1998a,
Volume 3, Section 3.8). The estimated consumption rates for vegetables, fruits, grains, beef, poultry,
milk, eggs, and water were from the results of the survey (TRW 19981, Tables 9-14 through 9-20, pages
T9-20 to T9-26). Generic food-transfer factors were from IAEA (1994, pages 5 to 58). The amount of
plant uptake of radionuclides used in the calculations was taken from LaPlante and Poor (1997, pages
2-12 to 2-14).
The analysis calculated the dose from each radionuclide that would reach the reference person by
multiplying the amount of radionuclide ingested, inhaled, or deposited near that person by the dose
conversion factor for that radionuclide. Dose conversion factors have important uncertainties associated
with them. However (as is customary for radiological compliance evaluations and EISs), this analysis
used only fixed values derived by methods from the International Commission on Radiological
Protection Publication 30 (ICRP 1979, all). These methods are similar to those specified by the
Environmental Protection Agency (Eckerman, Wolbarst, and Richardson 1988, all).
The long-term performance assessment calculations used the statistical distributions of biosphere dose
conversion factors. When the postulated climate change occurred during the model run, the biosphere
dose conversion factors changed to reflect the precipitation patterns associated with the new climate. The
major impact of a wetter climate would be to reduce the amount of well water required for irrigation. The
analysis did not consider other climate-related effects such as the appearance of springs, seeps, or other
surface water, because they would be unlikely to cause a large change in the consequences for a
maximally exposed individual. The result was the annual dose rate that the reference person would
receive from that radionuclide at a given time. The reference person (referred to in this EIS as a
maximally exposed individual) was developed from a series of lifestyle assumptions based on the surveys
of lifestyles in the region. Details on the reference person development are in the Viability Assessment
(DOE 1998a, Volume 3, pages 3-150 to 3-155).
In the analyses for this EIS, the same biosphere dose conversion factors were used for the four locations
considered [5, 20, 30, and 80 kilometers (3, 12, 19, and 50 miles)]. The biosphere dose conversion
factors are appropriate for the 30-kilometer location due to its similarity to the 20-kilometer location.
However, using the same factors for the other locations resulted in a systematic dose overestimation at
5 and 80 kilometers. This overestimate resulted because not all of the exposure pathways considered in
the calculation of biosphere dose conversion factors for the 20-kilometer location were appropriate for the
5-and 80-kilometer locations. The 5-kilometer location would be a drinking-water-only pathway
(ingestion dose only) because this location is not suitable to irrigation or farming. The 80-kilometer
1-49
Environmental Consequences of Long-Term Repository Performance
location is a lake playa, wliere evaporating contaminated water would result in deposits of contaminated
dust. Resuspension of the contaminated dust present the only exposure pathway for this location (that is,
drinking water and irrigation water pathways would not be relevant). However, development and use of
location-specific biosphere dose conversion factors for 5 and 80 kilometers would only serve to reduce
the calculated impacts reported in this EIS. Therefore, using the biosphere dose conversion factors
developed for the Viability Assessment (DOE 1998a, Volume 3, pages 3-158 to 3-161) for the
20-kilometer location at all other locations evaluated in this EIS is considered conservative.
1.5 Waterborne Radioactive l\/laterial Impacts
This section presents the total radiological dose to maximally exposed individuals, as calculated by the
RIP model, at the following four groundwater withdrawal or discharge locations downgradient from the
Yucca Mountain site where contaminated water could reach the accessible environment:
• A potential well 5 kilometers (3 miles) from the repository
• A potential well 20 kilometers (12 miles) from the repository
• A potential well 30 kilometers (19 miles) from the repository
• Franklin Lake Playa, the closest potential groundwater discharge point downstream from the
repository [80 kilometers (50 miles)]
The total radiological dose was calculated from repository closure to 10,000 years following closure and
at a time when the peak radiological dose would be observable. RIP model simulations carried out to 1
million years after repository closure also will include the peak radiological dose. These results are
provided in Section 1.5.1.
Apparent anomalous behavior of total radiological dose results predicted by the RIP model for the low
and intermediate thermal load scenario under the Proposed Action inventory is explained in Section 1.5.2.
The sensitivity of the estimates of waterborne radioactive material impacts to the fuel cladding model is
examined in Section 1.5.3.
1.5.1 TOTAL RELEASES DURING 10,000 YEARS AND 1 MILLION YEARS
The RIP model calculated radionuclide releases and radiological doses from individual nuclides and th(
total radiological dose due to all nine modeled radionuclides released from the repository from failed
waste packages. The model calculated total radiological dose in either of two ways: as a single run usin]
expected values of variable parameters, or in multiple realizations (runs) using randomly selected values
for distributed parameters. The model can calculate the total radiological dose as the expected value of
individual nuclides or the sum of all nuclides, for which sum the model chooses the mean value of all
distributed parameters. In addition, the model can use the Monte Carlo code to stochastically, or
randomly, perform any number of realizations or runs to select values of the distributed parameters. The
stochastic nature of the predictions is shown by the complementary cumulative distribution function of
the total radiological dose rate (that is, the sum of doses over all radionuclides) for 10,000 or 1 million
years. The total radiological dose represents the radiological dose to a maximally exposed individual at
the accessible environment using potentially affected groundwater for drinking water. The
complementary cumulative distribution functions discussed in this section represent the result of 1(X)
realizations of the RIP model.
4
1-50
Environmental Consequences of Long-Term Repository Performance
The number of realizations used for a Monte Carlo simulation is an important issue with respect to the
reliability of analysis results and proper allocation of resources. The number of runs required to reliably
predict peak dose rates was examined (DOE 1998a, Volume 3, page 4-71). To verify that 100 realizations
would be sufficient, 10,000-year and 100,000-year simulations for the high thermal load scenario with
Proposed Action inventory were carried out with 1,000 and 300 realizations, respectively. The resulting
distributions of peak individual radiological dose rates were compared with the 100-realization base case
results for both periods. The complementary cumulative distribution functions for each time period were
found to nearly match. The 100-realization complementary cumulative distribution functions did not go
below a probability of 0.01 because each predicted dose rate has a probability of occurrence of one one-
hundredth, or 0.01. Similarly, the 1,000- and 300-realization distributions display minimum probabilities
of 0.001 and 0.003, respectively. Peak dose rates did continue to increase as probability decreased.
Increased dose rates at these low probabilities were caused by combinations of extremely uncertain
parameter values sampled from the tails of the parameter probability distributions. However, 100
realizations appear to be sufficient for a good compromise between cost and precision.
Figures 1-37 through 1-39 show the 10,000-year and 1 -million-year complementary cumulative
distribution functions of total peak radiological dose for the Proposed Action inventory (see Section
1.3.1.2) at 5, 20, 30, and 80 kilometers (3, 12, 19, and 50 miles). In sequence, these figures show the total
radiological dose at human exposure locations for the high, intermediate, and low thermal load scenarios
and show that the maximum peak radiological dose (total for all nuclides) would occur well after 10,0(X)
years. Further, the 10,000-year complementary cumulative distribution functions show that the distance
(of the four distances analyzed) at which the highest total radiological dose would occur is 5 kilometers
from the repository. As groundwater moves downgradient from the Yucca Mountain site, it flows from
tuffaceous rocks to an alluvial aquifer. The pattern of the complementary cumulative distribution reflects
the fact that there would be greater natural retardation in the alluvium than in the tuff portions of the
hydrostratigraphic units.
Figures 1-40 through 1-42 show the 10,(X)0-year and 1 -million-year complementary cumulative
distribution functions of total peak radiological doses for the Inventory Module 1 inventory at 5, 20, 30,
and 80 kilometers (3, 12, 19, and 50 miles). In sequence, these figures show the total radiological doses
at human exposure locations for the high, intermediate, and low thermal load scenarios. As for the
Proposed Action inventory, these figures show that the maximum peak radiological dose (total, all
nuclides) would occur well after 10,(X)0 years. Again, the 10,(X)0-year complementary cumulative
distribution functions show that the distance (of the four distances analyzed) at which the highest total
radiological dose would occur is 5 kilometers from the repository.
For the Viability Assessment and this EIS, the mean peak dose is the average peak dose of the 100
realizations of radiological dose to a maximally exposed individual (that is, the peak for each realization
is determined and all peaks are averaged). The 95th-percentile peak dose is the average of the 95th- and
96th-highest ranked peak doses of the 100 realizations of radiological dose to a maximally exposed
individual (that is, the peak for each realization is determined, those peaks are ordered from lowest to
highest, and the average of the 95th- and 96th-highest is computed).
1.5.2 APPARENT ANOMALOUS BEHAVIOR BETWEEN LOW AND INTERMEDIATE
THERMAL LOAD RESULTS FOR PROPOSED ACTION INVENTORY
Comparison of the expected-value simulations for the different thermal load scenarios at the same
distance from the repository reveals apparent anomalous behavior. The differences between the scenarios
involving low and intermediate thermal loads under the Proposed Action inventory, which show that the
low thermal load curve crosses over the intermediate thermal load curve, require further explanation.
The analysis of three thermal load scenarios revealed some differences in performance as measured by the
calculation of total radiological dose to maximally exposed individuals at various distances from the
1-51
Environmental Consequences of Long-Term Repository Performance
repository. In particular, there is an apparent inconsistent relationship between the total dose-rate history
curves for the low and intermediate thermal load scenarios at 20 kilometers (12 miles) from the
repository. The apparent differences can be explained by the following factors:
• The effect of repository-area shape on the calculation of the dilution factor using the 3DADE
analytical solution (Leij, Scaggs, and van Genuchten 1991, all)
• Waste package degradation differences resulting in the solubility-limited transport, among the
different repository blocks being considered for disposal, of neptunium-237 from waste-form
degradation
• The correlative differences in the percolation flux
1.5.2.1 Effect of the Dilution Factor
The saturated zone dilution factors were presented and discussed in Section 1.4.5.4. As noted in that
section, the major influences on the calculated dilution factors were the geometry of the total repository,
the orientation of the repository relative to the direction of groundwater flow, and the average estimated
infiltration for each repository block. The important finding was that for each repository configuration,
extension of the repository area in a direction orthogonal to that of groundwater flow had little effect on
the calculated dilution factor. However, when calculated for an enlargement parallel to groundwater
flow, there were changes in the range of two to three times the dilution factors.
Thus, the intermediate thermal load Proposed Action scenario for the 20-kilometer (12-mile) distance had
the same dilution factor as the high thermal load Proposed Action scenario, because the repository shape
was relatively similar with essentially no change orthogonally to the flow direction. In contrast, the low
thermal load Proposed Action scenario for the 20-kilometer distance has almost double the area of the
intermediate thermal load Proposed Action scenario. Moreover, the repository is approximately twice as
long in the direction parallel to groundwater flow, resulting in a dilution factor almost two times less than
that of the intermediate thermal load Proposed Action scenario. Thus, because of the repository
geometry, the dilution factors between the low and intermediate thermal load Proposed Action scenarios
would result in less dilution under the low thermal load scenario.
1.5.2.2 Effect of Waste Paclcage Degradation
Figure 1-43 shows the total-radiological-dose-history curve for the Proposed Action inventory for the
intermediate and low thermal load scenarios. The peak radiological dose from the low thermal load
scenario is slightly delayed compared to the intermediate thermal load scenario, due to the delay in
package failure initiation for the low thermal load scenario. An examination of the waste package failure
distribution between these two scenarios (Figure 1-44) shows that after the initial juvenile package failure
(one package fails early for every case) stipulated by the Viability Assessment analysis, the first failure of
the intermediate thermal load scenario is about 9,000 years after repository closure, whereas the first
failure of the low thermal load scenario is about 27,000 years after repository closure. Thus, the amount
of neptunium-237 available for removal from the repository is less for the low thermal load scenario than
for the intermediate thermal load scenario.
The disparity in amount of neptunium-237 available for removal persists until the time of the super-
pluvial climate. Figure 1-43 shows that until the super-pluvial climate cycle (about 300,000 years after
repository closure) the low thermal load total radiological dose history curve lies below and later than the
intermediate thermal load total radiological dose history curve. Essentially, the peak radiological doses
occur at different times by that same amount of material removed. At this time, the number of waste
package failures has increased to allow differences in removal rates from the repository due to the
solubility limitations of neptunium-237. A larger proportion of the neptunium-237 is removed under the
1-52
Environmental Consequences of Long-Term Repository Performance
intermediate thermal load conditions because of the relatively higher amount of percolation flux and
larger number of waste packages for the upper block for this scenario. However, more of the neptunium-
237 remains in the repository under the low thermal load case because it can not all be removed from the
larger repository area due to the reduced amount of water. The total-radiological-dose-to-receptor curve
then crosses over the intermediate thermal load curve at about 300,000 years after closure. Thereafter, the
two curves slowly approach one another during the remainder of the simulation but never recross during
the simulated period.
1.5.2.3 Effect of Percolation Flux Distribution
The percolation flux differs across Yucca Mountain, especially in relation to the proposed areas. Figure
1-45 shows the average percolation flux for the different repository areas. Note that Block 5 has the
lowest percolation flux and Block 8 has the largest percolation flux. The intermediate thermal load
Proposed Action scenario includes only the upper block (Block 1) and the capture areas are similar to the
high thermal load Proposed Action scenario. The average infiltration flux for the upper block is larger
than that for Block 8.
A sensitivity analysis using only the long-term average climate shows that the release rate of neptunium-
237 at the top of the water table has two peaks. One is influenced by percolation flux in capture regions
1, 2, and 4, and the other is influenced by percolation flux in capture regions 3, 5, and 6. The reason for
the two-peak aspect of the total release-rate curve is that neptunium-237 is solubility limited, and the
lower percolation flux in the lower block and Block 8 does not completely remove all of the available
neptunium-237 from these blocks at the same rate as in areas with greater percolation flux. The
comparable curve for the intermediate thermal load Proposed Action scenario shows that all neptunium-
237 is released at approximately the same time. Figures 1-46 through 1-49 show a comparison of the
neptunium-237 radiological dose-rate histories for the low and intermediate thermal load scenarios for
only the average long-term climate at the engineered barrier system and at the exposure location
[20 kilometers (12 miles)]. These figures show that the difference in percolation flux is apparent at the
engineered barrier system and accentuated in the saturated zone because of the retarded release of
neptunium-237 under lower percolation flux. Because neptunium-237 is the dominant radionuclide
contributing to the total radiological dose at times greater than 100,0(X) years, the curves indicating the
low and intermediate thermal load total radiological-dose rate history cross. After crossing, the curves do
not maintain their separation but tend to approach one another without recrossing for the remainder of the
1 -million-year simulation period. It appears that they would likely cross again between 1 million and 1.5
million years at the observed rate of closure if the simulation were extended.
1.5.2.4 Conclusion
The analysis of the three thermal loads proposed for the planned repository configuration revealed
anomalous differences in performance as measured by the calculation of total radiological dose to
maximally exposed individuals at various distances from the repository. The apparent differences can be
explained by three factors:
• The effect of repository area shape on the calculation of the saturated zone dilution factor using the
3DADE numerical code, based on an analytical solution to flow and transport from the repository
• Differences in waste package failure under the different thermal loads
• Differences in the percolation flux and the correlative neptunium-237 solubility-limited transport
among the different repository blocks being considered for disposal
t:
1-53
Environmental Consequences of Long-Term Repository Performance
1.5.3 SENSITIVITY TO FUEL CLADDING MODEL
Section 5.4.4 of this EIS describes a sensitivity analysis DOE conducted to assess the importance of fuel
pin cladding protection on radiological dose. This section contains additional details for the sensitivity
analysis.
The average radionuclide inventory listed in Table I-l for each commercial spent nuclear fuel waste
package was used in the sensitivity analysis. Under the Proposed Action, approximately 1.2 percent of
the spent nuclear fuel would have stainless-steel cladding rather than zirconium-alloy cladding. The
stainless steel would degrade much faster than zirconium alloy, so the sensitivity analysis neglected
stainless-steel cladding as a protective barrier. In addition, approximately 0. 1 percent of the fuel pins are
proposed to fail in the reactor environment. Thus, under the Proposed Action, 1.3 percent of the
radionuclides in every spent nuclear fuel waste package would be available for degradation and transport
as soon as the waste package failed.
For the purposes of comparison, the analysis performed additional stochastic runs for 10,000 and
1 million years after repository closure assuming the zirconium-alloy cladding would provide no
resistance to water or radionuclide movement after the waste package failed. Table 1-38 compares the
peak radiological dose rate from groundwater transport of radionuclides for the base case and this case,
which assures zirconium-alloy cladding would not be present. The analysis used data representing the
high thermal load scenario to calculate individual exposures for a 20-kilometer (12-mile) distance only for
purposes of comparison.
Table 1-38. Comparison of consequences for a maximally exposed individual from groundwater releases
of radionuclides using different fuel rod cladding models under the high thermal load scenario.
Mean consequence" 95th-percentile consequence
Dose rate Probability Dose rate Probability
Maximally exposed individual (millirem/year) ofanLCF (millirem/year) ofanLCF
Peak at 20 kilometers'' within 10,000 years after 0.22 7.6x10"'' 0.58 2.0x10"^
repository closure with cladding credit
Peak at 20 kilometers within 10,000 years after 5.4 1.9x10"" 15 5.3x10"'*
repository closure without cladding credit
Peak at 20 kilometers within 1 million years 260 9.0x10"^ 1,400 5.0x10"^
after repository closure with cladding credit
Peak at 20 kilometers within 1 million years after 3,000 1.1x10"' 10,800 3.8x10"'
b
repository closure without cladding credit
a. Based on sets of 100 simulations of total system performance, each using random samples of uncertain parameters.
b. Represents a value for which 95 out of the 100 simulations yielded a smaller value.
c. LCF = latent cancer fatality; incremental lifetime (70 years) risk of contracting a fatal cancer for individuals, assuming a risk
of 0.0005 latent cancer per rem for members of the public (NCRP 1993a, page 31).
d. To convert kilometers to miles, multiply by 0.62137.
Figure 1-50 shows complementary cumulative distribution functions of the peak radiological dose rates
for the four suites of model runs. Approximately 25 percent of the 10,000-year runs did not show any
releases to the locations at a distance of 20 kilometers (12 miles). The zero releases are the reason the
10,000-year curves in Figure 1-50 start at an exceedance probability of 0.73 and decrease with increasing
radiological dose rate. All of the 1 -million-year runs show releases at 20 kilometers.
The analysis assumed that the zirconium-alloy cladding would provide no barrier to water movement and
radionuclide mobilization after the failure of the waste package. However, DOE expects that the
zirconium alloy would provide some impediment to radionuclide mobilization when the waste package is
breached. Therefore, the results for no cladding listed in Table 1-38 should be viewed as an upper
boundary.
1-54
Environmental Consequences of Long-Term Repository Performance
1.6 Waterborne Chemically Toxic Material Impacts
Further transport analysis is wan^anted because the screening analysis (Section 1.3.2.3.3) indicated that the
repository could release chromium into groundwater in substantial quantities and thus could represent a
human-health impact. Surrogate calculations were performed using the RIP model and inputs based on
the radiological materials transport simulations. This approach selected a long-lived unretarded isotope
(iodine-129) to serve as a surrogate for chromium. Iodine is highly soluble and exhibits little or no
sorption so when corrected for radioactive decay, its movement represents scalar transport. This method
avoided the extensive inputs necessary to define a new species for the RIP model and revision of the
associated external function modules that the analysis had carefully constructed for the nine modeled
radionuclides.
1.6.1 CHROMIUM
The screening analysis for chemically toxic materials (Section 1.3.2.3) identified chromium from the
waste packaging as a potential impact of concern. This section describes a chromium inventory for use in
the RIP model and evaluates chromium impacts.
1.6.1.1 RIP Model Adaptations for Chromium Modeling
The following assumptions were applied to the chromium surrogate calculation approach:
1. Iodine-129 will serve adequately as a surrogate for chromium because it has a long radioactive
half-life, lacks decay ingrowth by predecessors in a decay chain in the RIP model calculations, and is
not retarded in groundwater (chromate is also unretarded). A small error introduced by the slight
radioactive decay of iodine-129 during the model simulations can be corrected by an analytical
expression as a postprocessing step.
2. Alloy-22 degradation and release is modeled using general corrosion depth of the corrosion-resistant
material taken from WAPDEG modeling results (Mon 1999, all) for both dripping and nondripping
conditions. The WAPDEG modeled the general corrosion depth (in millimeters per year) of
corrosion-resistant material for 4(X) waste packages were averaged to produce a general degradation
rate for dripping and nondripping conditions and converted to a fraction of corrosion-resistant
material per year rate for use in the RIP model. The fractional degradation rate curves are show in
Figure 1-51.
3. Chromium associated with stainless-steel components used in many commercial spent nuclear fuel
waste packages would be released proportionately with Alloy-22 chromium. This conservative
assumption effectively assumes no credit for the delay of the onset of interior stainless-steel
degradation or for the degradation rate of the interior stainless steel itself.
The treatment of Alloy-22 corrosion-resistant material degradation and chromium mobilization required
the redefinition of the RIP container model. This calculation used the "Primary Container" in the RIP
model to represent only the corrosion-allowance material (outer layer) of the waste package. The
"Secondary Container" in the RIP model (used to represent cladding in the radiological material transport
simulations) was not used. The waste matrix was used to represent the corrosion-resistant inner layer
made of Alloy-22. These steps, with the proper material inventory and degradation coefficients, enabled
the use of the current RIP model structure for this calculation.
1-55
Environmental Consequences of Long-Term Repository Performance
The following additional changes were made to the radiological RIP model input files to conduct the
surrogate chromium mobilization and migration calculation:
1 . Iodine- 1 29 solubility was specified as 1 ,976 grams per cubic meter (0. 1 2 pounds per cubic foot),
based on the near-field geochemistry screening study results for chromium (iodine-129 serving as a
surrogate for chromium). Section 1.3.2.3.1 contains details on determining this solubility limit.
2. For each source term, the inventory of all radionuclides (except iodine-129) received a value of zero.
3. The inventory of iodine-129 in each source term were specified in units of grams (rather than the
original units of curies) per waste package using the values in Tables 1-18 through 1-22 in Section 1.3.
All inventory was assigned to the RIP model Waste Matrix Fraction (and none to the Primary or
Secondary Container Fractions) in each source term.
4. The analysis assumed that mobilized chromium from the corrosion-resistant material would advect
directly from the exposed corrosion-resistant material surface onto the invert (drift floor).
5. All secondary container definitions were all changed to a "degenerate" distribution at time zero, to
eliminate the effects of any cladding protection from the calculation. A degenerate distribution
simply results in all secondary containers failing at the specified time. The Alloy-22
corrosion-resistant material layer would be outside the cladding and, hence, not a barrier from this
perspective.
6. The primary container definitions were changed to a "Degenerate" distribution at time zero, to
eliminate the effects of corrosion-allowance material protection. This step is necessary because the
protective benefits of the corrosion-allowance material are implicit in the WAPDEG results used to
directly incorporate corrosion-resistant material degradation into the RIP model.
7. The waste-form-degradation rate for each source term was replaced with new variables representing
weight-averaged Alloy-22 degradation. The definition of these degradation rates is detailed below.
8. RIP model output was requested in grams (mass) rather than curies (radioactivity).
To arrive at a weight-averaged fractional corrosion rate to apply to all waste packages of a given category
(spent nuclear fuel, high-level radioactive waste, or DOE spent nuclear fuel) in a given repository region,
the following steps were taken. The Alloy-22 generalized corrosion depth for dripping and nondripping
conditions was converted to a fractional degradation rate, as described above. The Alloy-22 fractional
corrosion rate was computed from a weighted average (with respect to the fraction of packages subject to
dripping and nondripping conditions in the current climate) of dripping and nondripping generalized
corrosion rates. This weight-averaged fractional degradation rate was then used to model the release of
chromium from the waste package to the near-field environment.
For the Proposed Action, 30 percent of the chromium inventory would originate from interior
stainless-steel components used in some commercial spent nuclear fuel waste packages (see Table 1-16).
Because the waste package would have to fail before degradation and transport of interior components
could begin, simply adding the two chromium inventories together would yield artificially high results.
A two-stage scoping analysis, following the steps outlined above for using the RIP model to calculate
chromate migration, was performed for the Proposed Action inventory under the high thermal load
scenario to predict chromate concentrations at the 5-kilometer (3-mile) distance. In the first stage, the
model was run with only the chromium inventory from the Alloy-22 corrosion-resistant material [904,(XX)
grams (about 2,0(X) pounds) of chromium per commercial spent nuclear fuel waste package] following the
steps outlined above for chromium modeling. In the second stage, the model was run again with only the
1-56
Environmental Consequences of Long-Term Repository Performance
interior stainless-steel inventory [514,000 grams (about 1,100 pounds) of chromium per commercial spent
nuclear fuel waste package] but used the complete WAPDEG waste package model (as used in the
Viability Assessment) to represent complete waste package containment. Only the commercial spent
nuclear fuel packages would differ; no interior stainless-steel internal components would be used in high-
level radioactive waste or DOE spent nuclear fuel containers. Each RIP model run was held to the same
random number seed (used to "seed" the random number generator that is used to select random values of
stochastic parameters) so the realizations would be replicated. The results of each simulation were
summed, with respect to realization and time step, to calculate the total chromium concentration at 5
kilometers (3 miles). The results are listed in Table 1-39.
Table 1-39. Chromium groundwater concentrations (milligrams per liter)' at 5 kilometers (3 miles) under
Proposed Action inventory using the high thermal load scenario and a two-stage RIP model.
Peak chromium concentration
Model Mean 95th-percentile
RIP Stage 1 : Corrosion-resistant material ( Alloy-22) chromium inventory 0.0085 0.037
RIP Stage 2: Interior-to-waste package (SS/B" alloy) chromium inventory 0.000000086 0.00000048
Totals (Stage 1 + Stage 2, by realization; time step) 0.0085 0.037
a. To convert milligrams per liter to pounds per cubic foot, multiply by 0.00000624.
b. SS/B = stainless-steel boron.
The chromium concentrations obtained in this scoping analysis demonstrated that the inventory of
chromium associated with interior stainless-steel components, although it would represent 30 percent of
the total chromium inventory, would be small with respect to the peak chromium concentration in
groundwater at the closest downgradient location considered. Including the interior stainless-steel
chromium inventory increased the estimate of the mean peak chromium concentration by 0.00088 percent
over modeling the corrosion-resistant material chromium alone. The 95th-percentile peak chromium
concentration was increased by 0.000072 percent over modeling the corrosion-resistant material inventory
of chromium alone. Therefore, an additional step to model the interior stainless-steel corrosion and
transport was unnecessary to predict peak chromate concentrations.
Two factors would contribute to the inconsequential impact of the chromium inventory from the waste
package interior. First, the Alloy-22 in the waste package would have to be breached before interior
stainless steel was exposed to water and began to degrade. Thus, much of the chromium in the Alloy-22
would already have migrated before the interior stainless-steel chromium began to degrade and migrate.
Second, the Alloy-22 degradation would depend strongly on the RIP model parameters controlling the
fraction of packages exposed to dripping conditions. Packages that experienced dripping conditions
would degrade much faster; only those that experienced dripping conditions would fail within 10,000
years and permit exposure of interior stainless steel. The vast majority of waste packages would not fail,
so the interior chromium inventory would never be exposed for degradation and transport.
Based on this demonstration of the relative unimportance of the interior stainless-steel chromate inventory
in calculating peak chromium concentrations within 10,000 years, only the corrosion-resistant material
(Alloy-22) in the chromium inventory was simulated for analysis of chromium impacts as a waterbome
chemically toxic material.
1.6.1 .2 Results for the Proposed Action
The chromium-migration calculation was conducted for the Proposed Action inventory under the high,
intermediate, and low thermal load scenarios using the same stochastic approach as that used for the
waterbome radioactive material assessment. The 100 independent realizations, using randomly selected
input parameter values chosen from assigned probability distributions of values, were simulated with the
RIP model. Simulations were performed to estimate chromium concentrations at 5, 20, 30, and 80
kilometers (3, 12, 19, and 50 miles) for I0,0(X) years following closure. The resulting concentrations
1-57
Environmental Consequences of Long-Term Repository Performance
were decay-corrected to remove the slight radioactive decay calculated by the REP model for the surrogate
constituent, iodine-129.
The mean peak concentrations and 95th-percentile peak concentrations computed with the RIP model,
using the surrogate chromium-migration calculation described above, are listed in Table 1-40 for all
thermal load scenarios under the Proposed Action. Figures 1-52 through 1-54 show the complementary
cumulative distribution function for the 100 realizations of chromium concentration under the Proposed
Action at each of the four locations for the low, intermediate, and high thermal load scenarios,
respectively.
Table 1-40. Peak chromium groundwater concentration (milligrams per
liter) under the Proposed Action inventory.
Thermal load
Maximally exposed individual
Mean
95th-percentile
High
At 5 kilometers"
0.0085
0.037
At 20 kilometers
0.0028
0.012
At 30 kilometers
0.0018
0.0063
At 80 kilometers
0.00022
0.00061
Intermediate
At 5 kilometers
0.0029
0.0096
At 20 kilometers
0.0023
0.010
At 30 kilometers
0.00080
0.0038
At 80 kilometers
0.000031
0.00015
Low
At 5 kilometers
0.0046
0.016
At 20 kilometers
0.0018
0.0083
At 30 kilometers
0.00067
0.0033
At 80 kilometers
0.000053
0.00034
a. To convert milligrams per liter to pounds per cubic foot, multiply by 0.0000624.
b. Based on 100 repeated simulations of total system performance, each using randomly
sampled values of uncertain parameters.
c. To convert kilometers to miles, multiply by 0.62137.
A simple sensitivity run, reducing the solubility limit of the iodine-129 surrogate by one order of
magnitude (from 1,976 to 197.6 milligrams per liter), demonstrated that the imposed value of the
solubility limit did not affect the resulting concentration at the accessible environment. This
demonstration suggests that the chromium degradation rate is a major controlling factor over the release
of chromium.
There are two measures for comparing human health effects for chromium. When the Environmental
Protection Agency established its Maximum Contaminant Level Goals, it considered safe levels of
contaminants in drinking water and the ability to achieve these levels with the best available technology.
The Maximum Contaminant Level Goal for chromium is 0. 1 milligram per liter (0.0000062 pound per
cubic foot) (40 CFR 141.51). The other measure for comparison is the reference dose factor for
chromium, which is 0.005 milligram per kilogram (0.0004 ounce per pound) of body mass per day (EPA
1999, all). The reference dose factor represents a level of intake that has no adverse effect on humans. It
can be converted to a threshold concentration level for drinking water. The conversion yields essentially
the same concentration for the reference dose factor as the Maximum Contaminant Level Goal.
No attempt can be made at present to estimate the groundwater concentrations of hexavalent chromate in
Table 1-40, in terms of human health effects (for example, latent cancer fatalities). The carcinogenicity of
hexavalent chromium by the oral route of exposure cannot be determined because of a lack of sufficient
epidemiological or toxicological data (EPA 1999, all; EPA 1998, page 48).
1-58
Environmental Consequences of Long-Term Repository Performance
1.6.1.3 Results for Inventory Modules 1 and 2
Chromium impacts were calculated for Inventory Modules 1 and 2 using the same approach as for the
Proposed Action. Peak mean and 95th-percentile chromium concentrations for Inventory Modules 1 and
2 are listed in Tables 1-41 and Table 1-42, respectively. Figures 1-55 through 1-57 show the
complementary cumulative distribution function for the 100 realizations of chromium concentration for
Inventory Module 1 at each of the four locations for the low, intermediate, and high thermal load
scenarios, respectively.
Table 1-41. Peak chromium groundwater concentration (milligrams per liter)*
for 10,000 years after closure under Inventory Module 1.
Thermal load Maximally exposed individual Mean 95th-percentile
High
Intermediate
Low
At 5 kilometers
At 20 kilometers
At 30 kilometers
At 80 kilometers
At 5 kilometers
At 20 kilometers
At 30 kilometers
At 80 kilometers
At 5 kilometers
At 20 kilometers
At 30 kilometers
At 80 kilometers
0.032
0.14
0.018
0.10
0.0057
0.027
0.00029
0.00070
0.023
0.083
0.0089
0.042
0.0032
0.017
0.00019
0.00057
0.0093
0.0353
0.0050
0.022
0.0020
0.0084
0.000074
0.00026
a. To convert milligrams per liter to pounds per cubic foot, multiply by 0.0000624.
b. Based on 100 ref)eated simulations of total system performance, each using randomly
sampled values of uncertain parameters.
c. To convert kilometers to miles, multiply by 0.62137.
Table 1-42. Peak chromium groundwater concentration (milligrams per
liter)" due only to Greater-Than-Class-C and Special-Performance-
Assessment-Required wastes for 10,000 years after closure under Inventory
Module 2."
Thermal load Maximally exposed individual
Expected Value
High
Intermediate
Low
At 5 kilometers'^
At 20 kilometers
At 30 kilometers
At 80 kilometers
At 5 kilometers
At 20 kilometers
At 30 kilometers
At 80 kilometers
At 5 kilometers
At 20 kilometers
At 30 kilometers
At 80 kilometers
0.0014
0.00058
0.00021
0.000000012
0.00080
0.00033
0.00012
0.0000000094
0.00060
0.00025
0.000086
0.000000010
a. To convert milligrams per liter to pounds per cubic foot, multiply by 0.0000624.
b. Based on an expected value simulation using the mean of all stochastic parameters for the
additional inventory of Inventory Module 2 over Inventory Module 1 .
c. To convert kilometers to miles, multiply by 0.62137.
There are two measures for comparing human health effects for chromium. When the Environmental
Protection Agency established its Maximum Contaminant Level Goals, it considered safe levels of
contaminants in drinking water and the ability to achieve these levels with the best available technology.
The Maximum Contaminant Level Goal for chromium is 0.1 milligram per liter (0.0(X)0062 pound per
1-59
Environmental Consequences of Long-Term Repository Performance
4
cubic foot) (40 CFR 141.51). The other measure for comparison is the reference dose factor for
ciiromium, which is 0.005 milligram per kilogram (0.0004 ounce per pound) of body mass per day (EPA
1999, all). The reference dose factor represents a level of intake that has no adverse effect on humans. It
can be converted to a threshold concentration level for drinking water. The conversion yields essentially
the same concentration for the reference dose factor as the Maximum Contaminant Level Goal.
No attempt can be made at present to express the estimated groundwater concentrations of hexavalent
chromate in Table 1-42 in terms of human health effects (for example, latent cancer fatalities). The
carcinogenicity of hexavalent chromium by the oral route of exposure cannot be determined because of a
lack of sufficient epidemiological or toxicological data (EPA 1999, all; EPA 1998, page 48).
1.6.2 MOLYBDENUM
AlIoy-22 used as a waste package inner barrier also contains 13.5 percent molybdenum (ASTM 1994,
page 2). During the corrosion of Alloy-22, molybdenum behaves almost the same as the chromium. Due
to the corrosion conditions, molybdenum also dissolves in a highly soluble hexavalent form. Therefore,
the source term for molybdenum will be exactly 13.5/22 times (61.4 percent) the source term for
chromium. All the mechanisms and parameters are the same as those used for chromium so modeling is
unnecessary. It is reasonable to assume that molybdenum would be present in the water at concentrations
61.4 percent of those reported above for chromium.
There is currently no established toxicity standard for molybdenum (in particular, the Environmental
Protection Agency has not established a Maximum Contaminant Level Goal for molybdenum), although
this does not mean that molybdenum is not toxic. The concentrations of molybdenum would be very
small, so no effect would be likely to result from the molybdenum released to the groundwater.
1.6.3 URANIUM
While the screening analysis indicated that elemental uranium would not pose a health risk as a
waterbome chemically toxic material (see Section 1.3.2.3.3), it was retained for consideration for other
reasons. The total uranium inventory (all uranium isotopes) is listed for the inventory modules in
Table 1-23.
The reference dose for elemental uranium is 0.003 milligram per kilogram of body mass per day (EPA
1999, all). Assuming that a child would experience the maximum individual exposure from the drinking-
water pathway, the analysis used a 1-liter (0.26-gallon) daily intake rate and a 16-kilogram (35-pound)
body weight to convert the reference dose to a threshold concentration of 4.8 x 10'^ milligram per liter
(2.9 X 10"^ pound per cubic foot).
1.6.3.1 RIP Model Adaptations for Elemental Uranium Modeling
To evaluate the consequences of total uranium migration, the mobilization and transport of the total
uranium inventory for the Proposed Action listed in Table 1-23 were simulated using the RIP model. The
following steps were taken in the RIP model adaptation for the total uranium simulations:
1 . The inventory of all radionuclides except uranium was set to zero (as a precaution and to prevent
confusion with radiological runs).
2. The inventory of uranium (all isotopes) was changed to 8,1 19 kilograms (17,900 pounds) for
commercial spent nuclear fuel packages, 786 kilograms (1,730 pounds) for DOE spent nuclear fuel
packages, and 2,826 kilograms (6,220 pounds) for high-level radioactive waste packages.
3. Output from the RIP model was requested in grams rather than curies.
1-60
Environmental Consequences of Long-Term Repository Performance
4. The radiological decay rate of uranium-234 was left to represent all uranium isotopes in the waste
packages, although the resulting concentrations obtained from RIP model simulations were decay-
corrected to provide undecayed concentrations. Various uranium isotopes have different half-lives,
so the analysis ignored decay benefits in reducing impacts.
5. Because the chemical properties (such as sorption rate) are functions of the element and not the
isotope, the other transport properties of uranium were left the same as those used for the radiological
consequences simulations.
6. Use of the parameter FCSOLU, which is used in the RIP model to partition the solubility coefficient
to account for the fact that radionuclide simulations model only one isotope of uranium, was omitted
for full uranium elemental simulations.
DOE ran 100 simulations to model the release and transport of uranium. The Proposed Action inventory
is approximately 70,000 MTHM (77,000 tons). Although a small percentage of the heavy metal in the
spent fuel is not uranium, it was reasonable to assume all of it was because doing so had a very small
effect on the result and would make the analysis more conservative. This assumption introduced an
approximate 7-percent increase into the result. The runs are based on the high thermal load scenario, and
the consequences are computed for 5 kilometers (3 miles) from the repository. In addition, the analysis
neglected radioactive decay. Most of the uranium present has a very long half-life compared to the
analysis period, so decay would have a very small conservative effect on the result.
1.6.3.2 Results for the Proposed Action
The Proposed Action inventory of elemental uranium would be approximately 65 million kilograms
(72,000 tons) (see Table 1-23). Total elemental uranium migration calculations were made using the RIP
model code for the Proposed Action inventory under the high thermal load scenario for 10,000 years
following closure for the 5-kilometer (3-niile) distance. The resulting concentrations of elemental
uranium in groundwater at the 5-kilometer (3-mile) discharge location were obtained from the simulation
results.
The reference dose for elemental uranium is 3.0 x 10"' milligram per kilogram (4.8 x 10^ ounce per
pound) of food intake per day (EPA 1999, all). Assuming that a child would experience the maximum
individual exposure for the drinking water scenario, the analysis used a 1 -liter (0.26-gallon) daily intake
rate and a 16-kilogram (35-pound) body weight to convert the reference dose to a threshold concentration.
The threshold concentration would be 0.048 milligram per liter (3.0 x 10* pound per cubic foot).
The maximum uranium concentration over 10,000 years was extracted for each of the 100 sets of
simulation results. The mean peak concentration of uranium would be 6.7 x 10"* milligram per liter
(5.2 X 10"' pound per cubic foot), and the 95th-percentile peak concentration would be 2.2 x 10 *
milligram per liter (1.7 x 10"' pound per cubic foot). These concentrations would be six orders of
magnitude lower than the threshold concentration for the oral reference dose, so DOE expects no human
health effects from the chemical effects of waterbome uranium under the high thermal load scenario.
Figure 1-58 shows the complementary cumulative distribution function for elemental uranium
concentrations at the 5-kilometer (3-mile) discharge location for 10,000 years following closure under the
high thermal load scenario. The groundwater concentration information in this figure shows that
uranium, as a chemically toxic material, would be far below the reference dose at any probability level.
Based on trends in waterbome radioactive material results, the concentrations of elemental uranium at
locations that were more distant [20, 30, and 80 kilometers (12, 19, and 50 miles)] and for the
intermediate and low thermal load scenarios at all distance would be even lower. Because of the
extremely low concentrations from these simulations, further simulations were unnecessary to evaluate
1-61
Environmental Consequences of Long-Term Repository Performance
Other thermal loads under the Proposed Action. Elemental uranium would not present a health risk as a
chemically toxic material under the Proposed Action for any thermal load scenario.
1.6.4 RESULTS FOR INVENTORY MODULES 1 AND 2
Under Inventory Modules 1 and 2, the total uranium inventory would increase from the Proposed Action
total of 70,000 MTHM to 120,000 MTHM (Table 1-18). The 70-percent increase in elemental uranium
inventory would be likely to increase the groundwater concentration at the discharge location (1) at most,
if the percentage of the inventory was increased, or (2) by less, if solubility limits were exceeded along
the transport paths in groundwater in any case. Even doubling the groundwater concentrations calculated
for the Proposed Action inventory would result in concentration levels that would be several orders of
magnitude below the reference dose concentration level. Therefore, elemental uranium would not present
a substantial health risk as a chemically toxic material under Inventory Module 1 or 2 for any thermal
load scenario.
1.7 Atmospheric Radioactive Material Impacts
After DOE closed the Yucca Mountain Repository, there would be limited potential for releases to the
atmosphere because the waste would be isolated far below the ground surface. Still, the rock is porous
and does allow gas to flow, so the analysis must consider possible airborne releases. The only
radionuclide that would have a relatively large inventory and a potential for gas transport is carbon- 14.
Iodine-129 can exist in a gas phase, but it is highly soluble and therefore would be more likely to dissolve
in groundwater rather than migrate as a gas. Other gas-phase isotopes were eliminated in the screening
analysis (Section 1.3), usually because of short half-lives and because they are not decay products of long-
lived isotopes. After carbon- 14 escaped from the waste package, it could flow through the rock in the
form of carbon dioxide. Atmospheric pathway models were used to estimate human health impacts to the
local population in the 84-kilometer (52-mile) region surrounding the repository.
About 2 percent of the carbon- 14 in commercial spent nuclear fuel exists as a gas in the space (or gap)
between the fuel and the cladding around the fuel (Oversby 1987, page 92). The average carbon- 14
inventory in a commercial spent nuclear fuel waste package is approximately 12 curies (see Table I-l), so
the analysis used a gas-phase inventory of 0.23 curie of carbon-14 per commercial spent nuclear fuel
waste package to calculate impacts from the atmospheric release pathway. The analysis described in
Section 5.4 included the entire inventory of the carbon-14 in the repository in the groundwater release
models. Thus, the groundwater-based impacts would be overestimated slightly (by 2 percent) by this
modeling approach.
Carbon is the second-most abundant element (by mass) in the human body, constituting 23 percent of
Reference Man (ICRP 1975, page 377). Ninety-nine percent of the carbon comes from food ingestion
(Killough and Rohwer 1978, page 141). Daily carbon intakes are approximately 300 grams (0.7 pound)
and losses include 270 grams (0.6 pound) exhaled, 7 grams (0.02 pound) in feces, and 5 grams (O.OI
pound) in urine (ICRP 1975, page 377).
Carbon-14 dosimetry can be performed assuming specific-activity equivalence. The primary human-
intake pathway of carbon is food ingestion. The carbon-14 in food results from photosynthetic processing
of atmospheric carbon dioxide, whether the food is the plant itself or an animal that feeds on the plant.
Biotic systems, in general, do not differentiate between carbon isotopes. Therefore, the carbon-14 activity
concentration in the atmosphere will be equivalent to the carbon-14 activity concentration in the plant,
which in turn will result in an equivalent carbon-14 specific activity in human tissues.
1-62
Environmental Consequences of Long-Term Repository Performance
1.7.1 CARBON-14 RELEASES TO THE ATMOSPHERE
The calculation of regional radiological doses requires estimation of the annual release rate of carbon-I4.
The analysis based the carbon- 14 release rate on the predicted timeline of container failures for the high
thermal load scenario, using average values for the stochastic parameters that were entered. The expected
number of spent nuclear fuel waste package failures in 100-year intervals was used to estimate the carbon-
14 release rate after repository closure. The estimated amount of material released from each package as
a function of time was reduced to account for radiological decay.
As for the waterbome releases described in Section 5.4, some credit was taken for the intact zirconium-
alloy cladding (on approximately 99 percent by volume of the spent nuclear fuel) delaying the release of
gas-phase carbon- 14. The remaining 1 percent by volume of the spent nuclear fuel either would have
stainless-steel cladding (which degrades much more quickly than zirconium alloy) or would already have
failed in the reactor. The RIP model uses a waste package failure model that conceptually divides the
surface area of the waste packages into many patches. A corrosion future for each patch is then
calculated. The zirconium-alloy cladding failure model is implemented in the same fashion, with the
cladding corrosion rate set to a fraction of the corrosion rate of the Alloy-22 in the inner shell of the waste
package. This analysis set the cladding corrosion rate for the zirconium alloy to the same value used in
the Viability Assessment (DOE 1998a, Volume 3, page 3-101). A plot of the patch-area fraction of the
zirconium-alloy cladding that has failed as a function of time after repository closure is shown in
Figure 1-59. Although difficult to see on the plot scale, no zirconium-alloy cladding would fail during the
first 5,(XX) years after repository closure.
The amount (in curies) of carbon- 14 that would be available for transport from a failed
waste package, Aj, is calculated as:
At = (FiF -I- Ffc) X 0.23 curies per package
where:
FiF = fraction immediately failed (fuel with stainless-steel cladding or previously failed fuel
pins)
Ffc = fraction of failed cladding (if the value shown in Figure 1-59 is less than 0.01, then that
value is used; if the value shown in Figure 1-59 exceeds 0.01, then a value of 0.9875 is
used)
The model uses the patch failure rate on the zirconium alloy as the fraction of the failed pins until the
patch failure rate reaches 1 percent. After the patch failure rate reaches 1 percent, the release rate is reset
to not take further credit for zirconium-alloy cladding reducing the transport rate of gas-phase carbon- 14.
Rather than conducting a detailed gas-flow model of the mountain, the analysis assumed that the
carbon- 14 from the failed waste package would be released to the ground surface uniformly over a
l(X)-year interval. Thus, the release rate to the ground surface for a waste package would be At divided
by 100 (curies per year).
Figure 1-60 shows the estimated release rate of carbon-14 from the repository for 50,(XX) years after
repository closure, assuming that the spent nuclear fuel with stainless-steel cladding had failed and
released its gas-phase carbon-14 prior to being placed in a waste package. This assumption is represented
by FiF=0 in the calculation for At. The results in Figure 1-60 are based on the Proposed Action inventory.
Each symbol in the figure represents the carbon-14 release rate to the ground surface for a period of 100
years. The general downward slope of the symbols is due to radioactive decay (carbon-14 has a half-life
of 5,730 years). The symbols marking zero releases (curies per year) indicate that no waste packages
failed during some l(X)-year periods. The jagged nature of the plot indicates a different number of waste
packages failing in different l(X)-year intervals. Only 97 of 7,760 spent nuclear fuel waste packages
would have failed during the first 10,000 years after repository closure. By 40,000 years after repository
1-63
Environmental Consequences of Long-Term Repository Performance
closure, 676 of the 7,760 spent nuclear fuel waste packages would have failed. Using this expected-value
representation of waste package lifetime, no more than three waste packages would have failed in any
single 100-year interval before 30,000 years after repository closure. Between 30,000 and 50,000 years
after repository closure, as many as five waste packages would fail in a single 100-year interval. The
maximum release rate would occur about 19,000 years after repository closure. The estimated maximum
release rate would be about 0.098 microcurie per year.
1.7.2 ATMOSPHERE CONSEQUENCES TO THE LOCAL POPULATION
DOE used the GENII-S code (Leigh et al. 1993, all) to model the atmospheric transport and human
uptake of released carbon-14 for the 84-kilometer (52-mile) population radiological dose calculation.
This calculation used 84 kilometers rather than the typical 80 kilometers (50 miles) used in an EIS to
include the population of Pahrump, Nevada, in the impact estimate. Radiological doses to the regional
population near Yucca Mountain from carbon-14 releases were estimated using the population
distribution compiled from DOE (1998a, Volume 3, Figure 3-76), which indicates approximately 28,000
people would live in the region surrounding Yucca Mountain in the year 2000. The population by
distance and sector used in the calculations are listed in Table 1-43. The computation also used current
(1993 to 1996) annual average meteorology. The joint frequency data are listed in Table 1-44.
Table 1-43. Population by sector and distance from Yucca Mountain used to calculate regional airborne
consequences."
Distance from the
repository (kilometers)''
Direction
6^
16
24
32
40
48
56
64
72
84
Totals"
S
0
0
16
238
430
123
0
10
0
0
817
SSW
0
0
0
315
38
0
0
7
0
0
360
SW
0
0
0
0
0
0
868
0
0
0
868
WSW
0
0
0
0
0
0
0
0
87
0
87
W
0
0
0
638
17
0
0
0
0
0
655
WNW
0
0
0
936
0
0
0
0
0
20
956
NW
0
0
0
28
2
0
0
0
33
0
63
NNW
0
0
0
0
0
0
0
0
0
0
0
N
0
0
0
0
0
0
0
0
0
0
0
NNE
0
0
0
0
0
0
0
0
0
0
0
NE
0
0
0
0
0
0
0
0
0
0
0
ENE
0
0
0
0
0
0
0
0
0
0
0
E
0
0
0
0
0
0
0
0
0
0
0
ESE
0
0
0
0
0
0
0
0
1,055
0
1,055
SE
0
0
0
0
3
0
13
0
0
206
222
SSE
0
0
0
0
23
172
6
17
6,117
16,399
22,734
Totals
0
0
16
2,155
513
295
887
34
7,292
16,625
27,817
a. Source: Compiled from DOE (1998a, Volume 3, Figure 3-76).
b. To convert kilometers to miles, multiply by 0.62137.
c. The 80-kilometer (50-mile) distance typically used in an EIS analysis was increased to 84 kilometers (52 miles) in order to
include the population of Pahrump in the SSE sector in the calculations.
d. Population figures are estimates for 20(K}.
A population radiological dose factor of 2.2 x 10"' person-rem per microcurie per year of release was
calculated by the GENII code. For a 0.098-microcurie-per-year release, this corresponds to a
7.8 X lO'^-rem-per-year average radiological dose to individuals in the population. Thus, a maximum
84-kilometer (52-mile) population radiological dose rate would be 2.2 x 10''° person-rem per year. This
radiological dose rate represents 1.1 x 10''^ latent cancer fatalities in the regional population of 28,000
1-64
Environmental Consequences of Long-Term Repository Performance
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d d d o o
m 00 m cs rt On
Tt r~- (^ ■* vo <N
m o o — o o
d d d d d d
fS o o
— r- (N
(N o o
d d d o o o d
On
o
<N ON m
o d d d o
O On 00 On CN
n Q m r-) —
O O O — O
d d d d d o
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OOOOO cnOOi
t^ m — NO
00 — ■^ cs
U-) — o o
d d d d o
m (N u^
oo r-1 ^^
O (N O
o — d d o o o
CN CN CS On rO (^
t^ I/O m Q Q p
ro O O O O O
o d o d o o d d d d d d d d d d d d
S8
1^ NO ON
ro ON Q
1/0 O O
d d d o o o
[^ 1/-1 CN
ON O —
— — o
ON On r- "/o m
(N Q IT) m p
o o — o o
oooooo oooooo
ONNOONNDm ONOot^tn m
oodoo (Nooo o
d d d d d o d d d d o d
r^ O oo ON r-- cN O
NO 1^ r- (N «-, cN r^
nooo m — o
ddddoo dddooo
§rNi o
ON (N
<N — o
dddooo
r^ >r) fN vn ro
— — o >/-> p
O O m o o
d d d d d o
r<-| r^ 0^ — NO
r<^ — (N OO m
rj — — OOP ^mONCNj —
--— OO . O
O O
d d d d d o
OOOOOO
<N — On (N
m f*o fN m
(N — fS O
ddddoo
m On ^- NO ro
<N -a r~ o p
O O NO (N o
dddooo d d d d d o
oo o »/~i
oo ON tri
— CNl o
o ON r-~
Tt ON NO
m Tt o
dddooo
r^ ^ (N
NO 00 (N
Tt Tt —
d — d o o o
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- . - o ON - -
O O On n
d d d d d d
NO NO o m >n
fS 1^ ON 00 —
O O <N (N O
d d d d d o
OOOOOO
— t^l^OOOfS PO-^OOTtf-l
NONOmNOmro OiriTt^p
tsrNir^ — OO CM — — OO
dddddd d d d d d o
ON — ON
On 1^ <N
(N — O
dddooo
rj o (N
O t^ m
ro (S O
dddooo
(N w-i NO O fN) >r^
Tt <N fN r-1 On 1/0
— ' m in Tt — O
dddddd
<N t^ On T* m
f*-l NO On 1/0 p
CS CS fS — o
d d d d d o
^^OTtTtmro ooiriooinNO
SnOO^+PP mOrioi^p
r'OTt'q-OO OrONOmo
ddfsddd dd — cNJoo
^•oOTtooio— mONCMr^NO
OOTtOONO — -^ NO — 0<N0
OmONf*oiriN^ — ^rniriNOO
dddrorj— d d d d d o
O Tt t-~ m
OO O O p
ro Tt — O
ddddoo
1^ Tt — fS
o t- — —
T* NO r»0 O
ddddoo
On On 1^
UO t^ I/O
NO fS —
— 1^ r-1 IT) c»-i
NO r^ On m p
O ro — O O
0(N0000 OO — OOO
f<0 — O CS ro
rj NO Tt On p
o CNi r- m o
d d d d d o
NOOOTtNOOCM OnOnOnO
fSUOTtNOOO-^ t^TtI-~:*
oorNiint<o— oooo
dddddd ddddoo
ro On O r<^
Tt On ro O
r<0 Tt — O
ddddoo
OO Tt ^
NO tS O
d — d o o o
rn»rioONOr*o irioo^-o
CNfScspp — r~mcNi
O — roOO OO — O
dddddo ddddoo
m f*o O vo
<N n ON ro
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NO O ON CO
r) fN p p
OOOO
OOOOOO oooooo
NO ^ NO
NO NO n
ro (N O
dddooo
Tt O O r^ ^^ iri ro
O Tt (N — NO O p
NO m O O O — O
dddooo ddddoo
CS (S 1/0
NO r~ —
Tt ro O
dddooo
o lo r^
<N fS —
in m o
dddooo
m CNl T*
— On NO
NO m O
dddooo
m ON in f*o
fo r^ o p
NO Tt — ©
ddddoo
NO CM o ro
cs m — p
o o — o
ddddoo
On NO NO o
o o o
ddddoo
CN NO On O
— p c) n
OOOO
ddddoo
0^ NO 00 NO
SgS8
oooooo
tri O Cv O O (^
S88S88
— r- O ro
o o o S
ddddoo
Tt NO (N
Tt — —
(N — O
dddooo
ON m o rn ro
ri p (N p p
OOOOO
Tt m p
cs — O
OOOOOO oooooo oooooo
O NO NO r«0 m
_ _ in m O ON
(NCN- pp — rJr^-3
OO — OO OOOO
OO m cs O cs
fO p — CN —
OOOOO
OO ON CN ro
t- o — p
OOOO
OOOOOO OOOOOO OOOOOO oooooo
NO OO NO OO
CN r<0 m c<0
O O <N O
ddddoo ddddoo
—^ — 00 CO
33:2:2
inr^inmr-c*o 00 — 000
— noooonOO — 'OomcN
CNO — — 00 (NOOO
dddddd ddddoo
r-ONNOro Nor-^-"^r^
NO-^PP QNOOTtNO^-
cnOOO Oocncno
d d d d dddddo
cncnocnnono inr-- — cnnDco
T^inNO^-NOON mcNNO — pp
NO»ninONr*oO onnocncnOO
dddddd dddddd
CN r^ r-~
(N o —
ro — o
dddooo
00 CO r-
— in 00
Tt CN O
dddooo
NO 00 r- NO
O NO NO p
r~ NO (N o
ddddoo
[^ ON CO CO
O 1^ — p
00 CN — O
ddddoo
On 00 ON in ro
Sm CN o p
O (N — O
dddddo
NO CN CN CO NO
Sin Tt m p
O — (N O
dddddo
oocNcO'^moo TtmomON
NOOOOnOnCOCO rONOCN — CN
inNOONUot^cN ^-ooi^Tto
— NO CN NO
NO CO CN p
NO 00 CO O
— 00 — 00 CNOOOOO — OOOOO
<cQUQwu- <DauQMa. < cau Oum. < ca u Oiau. <DauCitQtt. <a3UQwii.
o^
00
On
CN
00
o
00
(N
V CO
CO ^
a.
^1
Si
< c
. o
■a u
ON "
ON >"
<^ S
■a
n.
b
<
i
>
^
u
0
mH
1-65
Environmental Consequences of Long-Term Repository Performance
persons each year at the maximum release rate. This annual population radiological dose rate corresponds
to a lifetime radiological dose of 1.5 x 10'^ rem over a 70-year lifetime, which corresponds to 7.6 x 10"'"
latent cancer fatalities during the 70-year period of the maximum release.
1.7.3 SENSITIVITY TO THE FRACTION OF EARLY-FAILED CLADDING
DOE performed a sensitivity analysis in which all of the cladding on commercial spent nuclear fuel that
had stainless-steel cladding (about 1.3 percent of the fuel by volume) was assumed to fail immediately as
the waste package failed. The commercial spent nuclear fuel with zirconium-alloy cladding was assumed
to fail as shown in Figure 1-57. The number of latent cancer fatalities per year in the local population at
the time of maximum release would increase from 1. 1 x 10"'^ to 4.0 x 10"" under the sensitivity analysis
assumptions. The time of maximum release would be 2,000 years after repository closure rather than
19,000 years after repository closure.
1-66
Environmental Consequences of Long-Term Repository Performance
T0UGH2
mountain-scale
thermal
hydrology
Xaqg
EQ3/6
near-field
geochemical
environment
NUFT
drift-scale
thermal
hydrology
PH
T,RH
WAPDEG
waste-package
degradation
T0UGH2
unsaturated
zone-flow
calibration
hydro
props
T0UGH2
drift-scale
unsaturated
zone flow
fsQs
h'
T0UGH2
mountain-scale
unsaturated
zone flow
qpSi
qi
FEHM
saturated
zone flow,
transport
'szi
GENII-S
biosphere
BDCFi
pH,IC03-2,l
ipit'patch*perf
Repository
integration program
waste-form
degradation,
EBS transport
~^
FEHM
unsaturated
zone transport
J
M,
SZ_CONVOLUTE
saturated
zone transport
1
Repository
integration program
dose
calculation
Afuel
CLAD_DEG
cladding
degradation
Final
Performance
Measure
Time
Run within repository
integration program
Biosphere
OUTPUT Parameters
T
Temperature
Qs
Seep flow rate
RH
Relative humidity
pH
pH
S|
Liquid saturation
IC03-2
Carbonate concentration
Xa
Air mass fraction
1
Ionic strength
%
Gas flux
"pit
Initial-pit-penetration time
qi
Liquid flux
'patch
Initial-patch-penetration time
qi
Infiltration flux
*perf
Perforated container area
Mi
Radionuclide mass flux
Afuel
Exposed fuel area
Ci
Radionuclide concentration
'szi
Saturated zone transport time
fs
Fraction of WPs with seeps
BDCFi
Biosphere dose conversion factor
EBS
Engineered Barrier System
Legend
3 External Process Model
H Repository Integration Program Cells
' 1 Repository Integration Program Calls
External Code
Response Surface Between
Process Models
■*■ Response Surface from
Process Model to RIP
► Between RIP Cells and
External Code
Source: Modified Irom DOE {1998a, Volumes, Figure 2-13),
Figure I-l. Total system performance assessment model.
1-67
Environmental Consequences of Long-Term Repository Performance
Legend
i to.' "»■*>-■ High thermal load
^— — Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Scale in feet
500 1,000
Scale: 1 :40,000
A
ri
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Modified from TRW (1999a. Figure 2.3-1, page 2-13).
Figure 1-2. Layout for Proposed Action inventory for high thermal load (85 MTHM per acre) scenario.
1-68
Environmental Consequences of Long-Term Repository Performance
N238000
N236000
N234000
N232000
Legend
High thermal load
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Scale in feet
500 1 ,000
Scale: 1 :40,000
$
Source: Modified from TRW (1999a. Figure 2.3-2. page 2-14).
Figure 1-3, Layout for Inventory Modules 1 and 2 for high thermal load (85 MTHM per acre) scenario.
1-69
Environmental Consequences of Long-Term Repository Performance
N234000
N230000
Legend
Intermediate thermal load
— — Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Note; The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Scale in feet
600 1.000
2,000
Scale: 1:40,000
$
Source: Modified from TRW (19998. Figure 2.3-3, page 2-15).
Figure 1-4. Layout for Proposed Action inventory for intermediate thermal load (60 MTHM per acre)
scenario.
1-70
Environmental Consequences of Long-Term Repository Performance
N238000
N236O00
N234000
N232000
N230000
Legend
Intermediate thermal load
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Note: The grid system Is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Scale in feet
500 1,000
Scale: 1:40,000
t
Source; Modified from TFWV (1999a. Frgure2.3^, page 2-16).
Figure 1-5. Layout for Inventory Modules 1 and 2 for intermediate thermal load (60 MTHM per are)
scenario.
1-71
Environmental Consequences of Long-Term Repository Performance
Legend
Low thermal load
— Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Scale in feet
500 1,000
Scale: 1 :40.000
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Modified (rom TRW (1999a, Figure 2,3-5, page 2-17),
Figure 1-6. Layout for Proposed Action inventory for low thermal load (25 MTHM per acre) scenario.
1-72
Environmental Consequences of Long-Term Repository Performance
N238000
N236000
N234000
N230000
Legend
Low thermal load
— — Fiscal Year 1997 Lawrence
Berkeley National Laboratory
model domain
Faults
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Scale in feet
500 1,000
A
II
Scale: 1:40.000
Source: Modrfied from TRW (1999a. Figure 2.3-6. page 2-18).
Figure 1-7. Layout for Inventory Modules 1 and 2 for low thermal load (25 MTHM per acre) scenario.
1-73
Environmental Consequences of Long-Term Repository Performance
N238000
BLOCK 7
BLOCK 6
N232000
Legend
EIS-PA design
Fiscal Year 1 997 Lawrence
Berlceley National Laboratory
model domain
Faults
Early design repository block
boundaries
©
Stratigraphic columns for
thermal hydrology
Scale in leet
500 1 ,000
Scale: 1 :40,000
I
Note; The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Modified from TRW (1999a. Figure 2,3-1 to 2.3-6. pages 2-13 10 2-18).
Figure 1-8. Relationship between the early performance assessment design and emplacement block layout
considered in this EIS performance assessment analysis.
1-74
Environmental Consequences of Long-Term Repository Performance
jo
«
o
u>
(D
o>
2
2
0)
Q.
E
|2
c
o
3
(/>
■g
'3
20 ^
Scaling with temperature
00-
80-
60-
40-
A — N.
^^.,,.^\ ,*''-*1 ■"'■^^^\.
■*\,
\
; : ' '
" ft
: :
20-
■ 2D
■ f = 0.8
f = 0.6
: ;
i i
1
0-
i —
i i
10
100
1,000
10,000
100,000
Time (years)
0.9
0.8
0.7
0.6
0.5
0.4
0.3
0.2
0.1
0.0
Scaling with
liquid saturation
-■* * -j-
- J - . ■
■
1
K
-■ T
; >
\ ■■^'
■1 /
:::i:.\ i : / 1
.\.B....;. . .
i I
\ ♦
\ I
A ;
\ 1
■
2D
-f = 0.8
.f = 0.6
^ - -
\j
f
,
10 100 1,000
Time (years)
10,000
100,000
Scaling with mass fraction air in gas
1,000
Time (years)
10,000 100,000
Soufce: Modified from TRW (1999a.
Figure 3,2-1. page 3-23).
Figure 1-9. Development of thermal load scale factors on the basis of two-dimensional and one-
dimensional model comparisons using time history of temperature, liquid saturation,
and air mass fraction.
1-75
Environmental Consequences of Long-Term Repository Performance
100
80
03
J2
Q)
o
to
0)
O
U>
o
E
60
40
20
0
T 1 1 1 1 1 r
] 1 1 1 1 1 1 1 1 1 r
' ' I
H >C X X X X »« X X»
V.
-^ — 500 years ^
Eqv. step function
at 500 years
—X — 10,000 years
Eqv. step function
at 10,000 years J
J_
_L
_L
_L
_L
El 70000
El 70200
El 70400
E1 70600
El 70800 El 71 000
El 71 200
Repository extent in x-direction (meters)
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Modified from TRW (1999a, Figure 3.2-2. page 3-24),
Figure I-IO. Partition of repository area between center and edge regions.
1-76
Environmental Consequences of Long-Term Repository Performance
240 I 1 — I — 1 I I I ni r — I — I I I I III 1 — 1 — I I I I III 1 — I — I I I Nil 1 — I — I I I I III 1 — I — I I I 1 1 1
0 I I I I I M I il I I I I I 1 1 II I I I Mini I I I I I I
Time (years)
BWR = boiling-water reactor; PWR = presurized-water reactor.
1E+5
Source: Modified (rom TRW (1999a. Figure 3.2-3, page 3-25).
Figure I-ll. Temperature and relative humidity histories for all waste packages for high thermal load
scenario, Proposed Action inventory, and long-term average climate.
1-77
Environmental Consequences of Long-Term Repository Performance
240
0 I I 1 I I ] I 111 I I I I 1 1 1 il I I I I I 1 1 il I I il I I I I 1 1 III
I I I I I 11 — ■ I I — I I I I I 11 1 1 — I Mill
Waste packages
H 12 PWR
-0— 21 PWR
-B — 21 PWR Design
-0 — 21 PWR Average
-r^s — 44 BWR #1
-^s^z— 44 BWR #2
-X — Codisposal
-^ — Direct disposal
_i I I I I I 1 1 II
1E-1
1E+0
1E+1
1E+2
Time (year)
1E+3
1E+4
1E+5
BWR = boiling-water reactor; PWR = presurized-water reactor.
Source: Modified Irom TRW (1999a, Figure 3.2-4. page 3-26).
Figure 1-12. Temperature and relative humidity histories for all waste packages, low thermal load
scenario, Proposed Action inventory, and long-term average climate.
1-78
Environmental Consequences of Long-Term Repository Performance
240
0 I I I r Mini I J I I I I III I I I
CD
>
a:
<X>ini^ r-)ie
■^ — 85 MTU/ac, base inventory
■O — 85 MTU/ac, expanded inventory
60 MTU/ac, base inventory
■Q — 60 MTU/ac, expanded inventory
Q — 25 MTU/ac, base inventory
i^ — 25 MTU/ac, expanded inventory
J I I I I r I rl I
1E-1
1E+0
1E+1
1E+2
Time (years)
1E+3
1E+4
1E+5
MTU/ac = metric tons of uranium per acre.
Source: Modified from TRW (1999a. Figure 3.2-5, page 3-27).
Figure 1-13. Temperature and relative humidity histories for the 21 pressurized-water-reactor average
waste packages, long-term average climate scenario, showing sensitivity to waste inventory.
1-79
Environmental Consequences of Long-Term Repository Performance
240
200
(0
1 160
o
o>
S 120
2
0)
a.
E
80
40
1 1 1 1 r M 1 1 1 1 1 1 1 1 1 1 ] T r 1 1 1 1 ii| 1 1 1 1 1 1 ii| 1 — 1 1 1 — i 1 1 1 1 1
■ ! K \ 1 1
"^ J \ i^" r\\ i
. : : 1 \ \ ;
1 : : : ^ \ i
; ; In \.
- 1 i 1 i r^^^
qI 1 1 — I M I III 1 1 — I I I I III I I I I I I III I I I I I I III I Ill
^ I I I N III 1 1 I I I I III
Center
0.0
1E-1
J I I 1 1 III 1 I I I 1 1 III I I I I 1 I I I I I I 1 1 III I I
1E+0 1E+1 1E+2
Time (years)
1E+3
1E+4
1E+5
Source: Modified from TRW (1999a. Figure 3,2-6, page 3-28),
Figure 1-14. Temperature and relative humidity histories for the 21 pressurized-water-reactor average
waste packages, high thermal load scenario, Proposed Action inventory, long-term average
climate scenario, comparing the center and edge scenarios.
1-80
Environmental Consequences of Long-Term Repository Performance
1.0
0.9
0.8
0.7
0.6
E
3
.c
05
<i)
>
m
0)
0.4
cc
0.3
0.2
0.1
0.0
- 1 1 .-' ' M"yji
,^— ■ 1 .
V ' ,- -" ^-/'
i
>. 1 y' -' l'' 11
^ v-"> - m -
;
\ \ \ 1' ;.
;
L.. A \.4 .'Aa
;
25 base center
stratigraphic column 7 •
- - 25 base center
stratigraphic column 2 J
25 base center
stratigraphic column 1 ;
- - 60 base edge '.
85 base edge
; ;
85 base center
. ..1 1
.
1 1 1 111J.1 L 1 1 1 i 1 1 i
1E+0
1E+1
1E+2 1E+3
Time (years)
1E+4
1E+5
200
^ 150
(0
3
jo
(D
U
(0
(D
<D
k_
cn
(D
Q.
E
100
50
^
T '-- ' 1 ' '
1. ^ ' ^ \ J . --
■
' 1 1 1 1 1 . 1
85 base center
60 base center
85 base edge
- - 60 base edge
- ~ 25 base center
stratigraphic column 1
- - 25 base center
stratigraphic column 2 .
25 base center
stratigraphic column 7
.^"^ / y
' '' ^ ^ \ •
■
1 ' -N.^-
\
■^-"•^"N.
y \ \ '
...-:s,-^-->xX -
\ \
i • ... .1..
1E+0
1E+1
1 E+2 1 E+3
Time (years)
1E+4
1E+5
Source: P*Ddified from TRW (1999a, Figure 3.3-1. page 3-29).
Figure 1-15. WAPDEG input temperature and relative humidity histories for all thermal loads with
Proposed Action inventory.
1-81
Environmental Consequences of Long-Term Repository Performance
£
O)
IS
I
E
1E+0
1E+1
1E+2
1E+3
1E+4
1E+5
Time (years)
200
150
85 expanded
85 expanded
85 expanded
85 expanded
60 expanded
60 expanded
60 expanded
25 expanded
25 expanded
25 expanded
25 expanded
center stratigraphic column 1
edge stratigraphic column 1
center stratigraphic column 2
edge stratigraphic column 2
center stratigraphic column 1
center stratigraphic column 2
center stratigraphic column 3
center stratigraphic column 1
center stratigraphic column 2
center stratigraphic column 4
center stratigraphic column 7
100 -
^.Vl
.,
■ ■ ■ ' I ■ ' I I I 1 1 1 1 I I I I I 1 1 1
1E+0
1E+1
1E+2 1E+3
Time (years)
1E+4
1E+5
Source: Modtfied from TRW (1999a. Figure 3.3-2, page 3-30),
Figure 1-16. WAPDEG input temperature and relative humidity histories for all thermal loads with
Inventory Modules 1 and 2.
1-82
Environmental Consequences of Long-Term Repository Performance
T3
01
Cfl
«
cn
(0
o
(0
a.
c
g
o
(0
u.
1.0
0.8
0.6
0.4
0.2
0.0
'\ f
I
'
\ 1
1 •
I 1
I
: 1
'
\ 1
1
^ 1*
> 1
>
; i
l(
; ■
F
Stratigraphic column 1 '.
- - Stratigraphic column 2 J
Stratigraphic column 7 '•
J
,
' ' • '
1E+2
1E+3
1E+4
Time (years)
1E+5
1E+6
Source: Modified from TRW (1999a, Figure 3 3-3. page 3-31)-
Figure 1-17. Time to first breach of the corrosion-allowance material for low thermal load scenario.
Proposed Action inventory, all three stratigraphic columns, always-dripping waste packages.
•D
in
u>
CO
o
CO
CL
c
o
o
(0
Numbers on plot refer to uncertaintyA/ariability
splitting sets shown on Table 1-29.
1E+4
Time (years)
1E+6
Source: Modified from TRW (1999a. Figure 3.3-4. page 3-32).
Figure 1-18. Time to first breach of the corrosion-resistant material for low thermal load scenario.
Proposed Action inventory, all three stratigraphic columns, always-dripping waste
packages, and all nine uncertainty/variability splitting sets.
1-83
Environmental Consequences of Long-Term Repository Performance
1000
-D 800
is
to
B 600
Q.
0)
jQ
E
g 4001-
0)
2
I
200
■~T 1 1 — I — rill
I'.. -A
li
'TT'
^-;-/-.
I " I I
1E+2
Numbers on plot refer to uncertainty/variability
splitting sets shown on Table 1-29.
1E+3
1E+4
Time (years)
1E+5
1
23
1E+6
Source: Modified from TRW (1999a, Figure 3.3-5, page 3-33).
Figure 1-19. Average number of patches failed per waste package as a function of time for low thermal
load scenario, Proposed Action inventory, all three stratigraphic columns, always-dripping
waste packages, and all nine uncertainty/variability splitting sets.
to
a>
O)
to
^^
u
CO
Q.
c
o
1
1.0
0.8
0.6
0.4
0.2-
0.0
—I — I — I — I 1 1 1 1 — I — I — I—
■ Center stratigraphic column 1
■ Edge stratigraphic colunnn 1
Center stratigraphic column 2
Edge stratigraphic column 2
1E+2
1E+3
1E+4
Time (years)
1E+5
1E+6
Source; Modified from TRW (1999a, Figure 3,3-6, page 3-34)
Figure 1-20. Time to first breach of the corrosion-allowance material for high thermal load scenario,
Inventory Modules 1 and 2, center and edge regions for both stratigraphic columns,
always-dripping waste packages.
1-84
Environmental Consequences of Long-Term Repository Performance
T3
0)
V
D)
CO
o
«
Q.
o
o
<o
1.0
0.8
0.6
0.4
0.2
0.0
1E+2
Numbers on plot refer to uncertainty/variability
splitting sets shown on Table 1-29.
1E+6
SotJfce: Modified from TRW {1999a. Figure 3.3-7, page 3-35).
Figure 1-21. Time to first breach of the corrosion-resistant material for high thermal load scenario,
Inventory Modules 1 and 2, center and edge regions for both stratigraphic columns,
always-dripping waste packages, and all nine uncertainty/variability splitting sets.
1000
f I I I T T T I I
I I I I I I I I
1E+3
Numbers on plot refer to uncertaintyA/arlability
splitting sets shown on Table 1-29.
Source: Hilodified from TRW H999a, Figure 3.3.8, page 3-36).
Figure 1-22. Average number of patches failed per package as a function of time for high thermal load
scenario. Inventory Modules 1 and 2, center and edge regions for both stratigraphic columns,
always-dripping waste packages, and all nine uncertainty/variability splitting sets.
1-85
Environmental Consequences of Long-Term Repository Performance
0)
(0
(D
O)
CO
o
(0
Q.
1.0
0.8
0.6
I 0.4
o
(0
0.2
0.0
1E+2
MTHM/acre; Inventory;
Region; Stratigraphic
85 Base Center Column 1
85 Base Edge Column 1
60 Base Center Column 1
60 Base Edge Column 1
25 Base Center Column 1
25 Base Center Column 2
25 Base Center Column 7
85 Expanded Center Column 1
85 Expanded Edge Column 1
85 Expanded Center Column 2
85 Expanded Edge Column 2
60 Expanded Center Column 1
60 Expanded Center Column 2
60 Expanded Center Column 3
25 Expanded Center Column 1
25 Expanded Center Column 2
25 Expanded Center Column 4
25 Expanded Center Column 7
1E+3
1E+4
Time (years)
1E+5
1E+6
Source: Modified from TRW (1999a. Figure 3.3-9, page 3-37).
Figure 1-23. Time to first breach of the corrosion-allowance material for all thermal loads and inventories,
all regions, always-dripping waste packages, uncertainty/variability splitting set 5.
(/}
0)
O)
CO
o
CO
Q.
1.0
0.8
0.6
■g 0.4
o
CO
0.2-
MTHM/acre; Inventory;
Region; Stratigraphic
85 Base Center Column 1
85 Base Edge Column 1
60 Base Center Column 1
60 Base Edge Column 1
25 Base Center Column 1
25 Base Center Column 2
25 Base Center Column 7
85 Expanded Center Column 1
85 Expanded Edge Column 1
85 Expanded Center Column 2
85 Expanded Edge Column 2
60 Expanded Center Column 1
60 Expanded Center Column 2
60 Expanded Center Column 3
25 Expanded Center Column 1
25 Expanded Center Column 2
25 Expanded Center Column 4
25 Expanded Center Column 7
O.Ol—
1E-(-2
1E+3
1E-1-4
Time (years)
1E-I-5
1E+6
Source: Modified (rom TRW (1999a, Figure 3,3-10. page 3-38)
Figure 1-24, Time to first breach of the corrosion-resistant material for all thermal loads and inventories,
all regions, always-dripping waste packages, uncertainty/variability splitting set 5.
1-86
Environmental Consequences of Long-Term Repository Performance
-a
(O
a>
1000
800
B 600
CO
Q.
5
E
3
<D
o>
(D
400
200
■ ■ ' 1 1 1
MTHM/acre; Inventory;
Region; Stratigraphic
85 Base Center Column 1
_ 85 Base Edge Column 1
60 Base Center Column 1
60 Base Edge Column 1
25 Base Center Column 1
25 Base Center Column 2
25 Base Center Column 7
85 Expanded Center Column 1
' 85 Expanded Edge Column 1
• 85 Expanded Center Column 2
- - - 85 Expanded Edge Column 2
60 Expanded Center Column 1
60 Expanded Center Column 2
60 Expanded Center Column 3
25 Expanded Center Column 1
25 Expanded Center Column 2
25 Expanded Center Column 4
25 Expanded Center Column 7
-1 — 1 — . — 1
/
/
I
J
1 1 — 1 ^ ' ' ' '
—r^T — r I" , , ','rr, 1 , —
1E+2
1E+3
1E+4
Time (years)
1E+5
1E+6
Sotjrce: Modified trom TRW (1999a, Figure 3.3-1 1. page 3-39).
! Figure 1-25. Average number of patches failed per waste package as a function of time for all thermal
loads and inventories, all regions, always-dripping waste packages, uncertainty/variability
splitting set 9.
160
140
■n
(U
120
m
(0
Q)
x:
100
(0
u.
.^_
o
80
<D
a
E
3
C
60
<D
TO
ffl
%
40
<
20
I I I I I r
MTHM/acre; Inventory;
Region; Stratigraphic
85 Base Center Column 1
85 Base Edge Column 1
60 Base Center Column 1
60 Base Edge Column 1
25 Base Center Column 1
25 Base Center Column 2
25 Base Center Column 7
85 Expanded Center Column 1
85 Expanded Edge Column 1
85 Expanded Center Column 2
85 Expanded Edge Column 2
60 Expanded Center Column 1
60 Expanded Center Column 2
60 Expanded Center Column 3
25 Expanded Center Column 1
25 Expanded Center Column 2
25 Expanded Center Column 4
25 Expanded Center Column 7
1E+2
1E+3
Time (years)
1E+6
Source: Modified from TRW (1999a. Figure 3.3-12, page 3-40).
Figure 1-26. Average number of patches failed per waste package as a function of time for all thermal
loads and inventories, all regions, always-dripping waste packages, uncertainty/variability
splitting set 5.
1-87
Environmental Consequences of Long-Term Repository Performance
N238000
N236000
N2340CI0
N2320CX)
Legend
High thermal load
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Note; The grid system Is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Scale In feet
500 1,000
n
Scale: 1 :40.000
Source: Modified from TRW (1999a, Figure 3.5-1. page 3-41)
Figure 1-27. Regions for performance assessment modeling, Option 1, high thermal load scenario,
Proposed Action inventory.
1-88
Envimnmental Consequences of Long-Term Repository Performance
Legend
Intermediate thermal load
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Scale in feet
0 SOO 1.000 2.000
Scale: 1:40,000
t
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Moditie<a from TRW (1999a. F'^ure 3.5-2. page 3-42),
Figure 1-28. Regions for performance assessment modeling, Option 2, intermediate thermal load scenario.
Proposed Action inventory.
1-89
Environmental Consequences of Long-Term Repository Performance
Figure 1-29. Repository block areas for performance assessment modeling, Option 3, low thermal load
scenario with Inventory Module 1 , and intermediate thermal load scenario with Inventory
Module 1 cases.
1-90
J
Environmental Consequences of Long-Term Repository Performance
Legend
High thermal load
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Scale in feet
500 1,000
Scale: 1:40.000
I
Note: The grid system is the Nevada Stale Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Modified from JVW (1999a. Figure 3.5-4, page 3-44).
Figure 1-30. Regions for performance assessment modeling. Option 4, high thermal load scenario,
Proposed Action inventory.
1-91
Environmental Consequences of Long-Term Repository Performance
Legend
Intermediate thermal load
— Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Scale in feet
500 1,000
Scale: 1 :40,000
A
ri
Note: The grid system Is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Modified from TRW (1999a. Figure 3.S-5, page 3-45),
Figure 1-31. Regions for performance assessment modeling, Option 5, intermediate thermal load
scenario, Inventory Module 1 .
1-92
Environmental Consequences of Long-Term Repository Performance
Figure 1-32. Repository block areas for performance assessment modeling, Option 6, low thermal load
scenario. Inventory Module 1 .
1-93
Environmental Consequences of Long-Term Repository Performance
Legend
in Capture region
Repository block
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Scale in feet
500 1,000
Scale; 1:40,000
2,000
A
ri
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Modified from TRW {1999a, Figure 3.5.7, page 3-47)
Figure 1-33. Capture regions for high and intermediate thermal load scenarios with Proposed Action
inventory.
1-94
Environmental Consequences of Long-Term Repository Performance
Legend
[2] Capture region
I 1 Repository block
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Scale in feet
500 1.000
Scale: 1 :40,000
I
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Source: Utodified from TRW (1999a. Figure 3.5-7. page 3-47).
Figure 1-34. Capture regions for low thermal load scenario with Proposed Action Inventory and low and
intermediate thermal load scenarios with Inventory Modules 1 and 2.
1-95
Environmental Consequences of Long-Term Repository Performance
N238000
N236000
N234000
N232000
Legend
[21 Capture region
I I Repository blocl<
Fiscal Year 1997 Lawrence
Berl<eley National Laboratory
model domain
Faults
Note: The grid system Is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing.
Scale in leet
500 1.000
A
n
Scale: 1 :40,000
Source: Modified (rom TRW (1999a. Figure 3.5-8. page 3-46)
Figure 1-35. Capture regions for high thermal load scenario with Inventory Modules 1 and 2.
1-96
Envimnmental Consequences of Long-Term Repository Performance
1-97
Environmental Consequences of Long-Term Repository Performance
km =
3
2.
8 Complementary cumulative probability
10°-
10-1 -
10-2-
rert kilom
■"■.?\ ^ N, '^'■■^, 1,000,000 years
■^Ol 0,000 years ^ "\\ |^
1
5 km
20 km
30 km
80 km
V "■ \ \ ^ ■ 1
^ ^ ^ ' M ' '^
\^'\ \ 1 ;
1 ' 1 ' 1 ' 1 ' 1
10-" 10-2 10° 102 104
Peak maximally exposed individual dose rate (millirem per year)
eters to miles, multiply by 0.62137.
Figure 1-37. Complementary cumulative distribution function of peak maximally exposed individual
radiological dose rates during 1 0,000 and 1 million years following closure for high thermal
load scenario with Proposed Action inventory (100 realizations, all pathways, all distances).
CO
2
Q.
.>
E
CJ
c
CD
E
10°-
10-1-
• '-.Y \ 1 '■. 1 ,000,000 years
V '•. 1 0,000 years \ \ '•. |
-
\
\
5 km
20 km
E
-
30 km
- \H %
o
O
10-2
80 km
\ • \ \ \\\
1U , 1 1 1 1 1 1 1 1 1 1
10-4 10-2 100 102 104
Peak maximally exposed individual dose rate (millirem per year)
km =
= kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-38. Complementary cumulative distribution function of peak maximally exposed individual
radiological dose rates during 10,000 and 1 million years following closure for intermediate
thermal load scenario with Proposed Action inventory (100 realizations, all pathways, all
distances).
1-98
Environmental Consequences of Long-Term Repository Performance
10" -If
(0
£1
P
.1
E
U
s
c
o
E
®
Q.
E
o
o
10-^
10-
10-* 10-2 10° 10^ 10"
Peak maximally exposed individual dose rate (millirem per year)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-39. Complementary cumulative distribution function of peak maximally exposed individual
radiological dose rates during 10,000 and 1 million years following closure for low thermal
load scenario with Proposed Action inventory (100 realizations, all pathways, all distances).
|U"_
-^_ ^ vr^^r?^^~^-v
^
---.•:rTT.^:Tj:
2
-
^^^-^
a>
-
~~ \ ^- <\ \
v"-. 1 ,000,000 years
>
-
\ \-A \
■ -* V
to
3
^ ^. -10,000 years
\ '■'•: \
3
10""-
\ '^ 'A
\ '■^'' \
O
,
£>
-
V. \ -A
\ > '-. \
c
■
\ '■ M
\ ^ ■■•. \
E
(D
Q.
■
5 km
20 km
-
30 km
\ N •■
\ '-\ \
80 km
\\\ \
10-2
- . \-.
\ \ \
1 '
10-"
1 ' 1 '
10-2 10°
102 -lo"
Peak maximally exposed individual dose rate (millirem per year)
km = kilometer. To convert kilometers to miles, multiply by 0.621 37.
Figure 1-40. Complementary cumulative distribution function of peak maximally exposed individual
radiological dose rates during 1 0,000 and 1 million years following closure for high thermal
load scenario with Inventory Module 1 (100 realizations, all pathways, all distances).
1-99
Environmental Consequences of Long-Term Repository Performance
lulative probability
""^^, V "" '~^".'\ 1 ,000,000 years
-,■-,10,000 years s ., •, \
Complementary curt
P
V \ ■. V • •. \
5 km
20 km
30 km
\ ^. ■••\
\ :
in2 —
80 km
\ *.\
\ ■.
10 . 1 1 1 , 1 1 1 , 1 1
10-" 102 10° 10^ 10"
Peak maximally exposed Individual dose rate (millirem per year)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-41. Complementary cumulative distribution function of peak maximally exposed individual
radiological dose rates during 10,000 and 1 million years following closure for intermediate
thermal load scenario with Inventory Module 1 (100 realizations, all pathways, all distances).
10° —
=<~-^^3rr:— -^^
!5
vrv ' ""^^^
^*
2
a.
V, \ ^ '"^■."•■.1,600,000 years
^^ ■ '. 10,000 years ^ "- \
ementary cumulat
p
1 1 1 1 1 1 1 1
\
5 km
\ --vA \ \'.
Q.
E
20 km
» \ \ W
o
O
30 km
\ '\ . ^^
in-2
80 km
\\ ' u
\ \"\ I !• 1
10 , 1 , 1 , 1 1 1 1 1 ,
10" 10^ 10° 102 10"
Peak maximally exposed individual dose rate (millirem per year)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-42. Complementary cumulative distribution function of peak maximally exposed individual
radiological dose rates during 10,000 and 1 million years following closure for low thermal
load scenario with Inventory Module 1 (100 realizations, all pathways, all distances).
I- 100
Environmental Consequences of Long-Term Repository Performance
200,000 400,000 600,000 800,000 1,000,000
Time (years)
RIP Version 5.19.01
1/7/99
Figure 1-43. Comparison of low and intermediate thermal load scenarios total radiological dose histories
for the Proposed Action inventory 20 kilometers (12 miles) from the repository.
10° -F
0)
is lO"'--
(0
<s>
CO
u
<a
Q.
T5
10-=
10-=
— I 1 r-
"T 1 1 1 1 1 r 1 1 1 1 T"
/
^
^;;;_;:::s::::::::^
;;/"
25 Low
_1 I L.
200,000 400,000 600,000
Time (years)
800,000
1,000,000
RIP Version 5.19.01
1/8/99
Figure 1-44. Waste package failure curves for low and intermediate thermal load scenarios.
I-lOl
Environmental Consequences of Long-Term Repository Performance
N238000
N236000
N234000
N230000
Legend
Low thermal load
Fiscal Year 1 997 Lawrence
Berkeley National Laboratory
model domain
Faults
Note: The grid system is the Nevada State Plane Coordinate System
converted to metric units. E = Easting; N = Northing,
Scale In feet
500 1.000
2,000
Scale: 1 :40,000
Source: Modified from TRW (1999a. Figure 4.1-8, page 4.17).
Figure 1-45. Average percolation flux for repository blocks.
1-102
Environmental Consequences of Long-Term Repository Performance
10-'
10-:
10-3-^
ca
>»
Q.
(0
.2
3
U
« 10''--
2
(D
(0
■s 10-5 -^
ir
io-«.
—I 1 1 1 1 1 r-
"/"^^^^^
Unsaturated zone total
Capture region 1
Capture region 2
Capture region 3
Capture region 4
Capture region 5
Capture region 6
*'*^>*^^*^^^n^>^vv^+v^^^,
•V.wj"'-**'»-'<-sv
200,000 400,000 600,000
Time (years)
800,000
1 ,000,000
Figure 1-46. Neptunium-237 release rate at the water table for fixed long-term average climate for low
thermal load scenario during the first 1 million years following repository closure.
10-^
r 10-2
a.
S 10-^
3
2
i_
0)
u>
IS
«
tr
lO-'l-r
10-5--
io-«
Unsaturated zone total
Capture region 1
Capture region 2
Capture region 3
Capture region 4
Capture region 5
Capture region 6
200,000 400,000 600,000 800,000 1,000,000
Time (years)
Figure 1-47, Neptunium-237 release rate at the water table for fixed long-term average climate for
intermediate thermal load scenario during the first 1 million years following repository
closure.
1-103
Environmental Consequences of Long-Term Repository Performance
Dose total
Capture region 1
Capture region 2
Capture region 3
Capture region 4
Capture region 5
Capture region 6
200,000 400,000 600,000
Time (years)
800,000 1 ,000,000
Figure 1-48, Neptunium-237 release rate at the end of the saturated zone for fixed long-term average
climate for low thermal load scenario during the first 1 million years following repository
closiu-e.
103 -F
? 10^ +
(1)
>.
S. 10^-^
E
<u
i 10°
E.
2 10- 4-
</)
S 10-2
10-=
Dose total
Capture region 1
Capture region 2
Capture region 3
Capture region 4
Capture region 5
Capture region 6
+
200,000
400,000
600,000
800,000
1 ,000,000
Time (years)
Figure 1-49. Neptunium-237 release rate at the end of the saturated zone for fixed long-term average
climate for intermediate thermal load scenario during the first 1 million years following
repository closure.
1-104
i
Environmental Consequences of Long-Term Repository Performance
10°-
^■'v,,^^^' ' .^
ity
' \ '" ^"^^*x^*
2
^X '" ^~\^''
■§
\ '"- ^t* * *
itivepr
\ No cladding \ *
^ lO.CXX) years \ ' , ^^.
A 1 \ No cladding
3
^ > \ 1 million years
1 10-1-
Cladding ' 1 «
3 :
10,000 years ' \ •
^
\ ^ \ '.
0
\ \ Cladding ,
■§
N. ' 1 million ,
1
\ \ years -^
o
\ \ \ "
\ « \ '
E
1
5
"^
10-2-
:
' 1 ' 1 ' 1 ' 1 ' 1 ' 1
10-* 10-2 10° 102 10" 106
Peak maximally exposed individual dose rate (miliirem per year)
Figure 1-50. Complementary cumulative distribution function of radiological doses with and without
cladding for a maximally exposed individual at 20 kilometers (12 miles) under the
Proposed Action 10,000 and 1 million years after repository closure.
1 0F-06-,
Always dripping ^^^^^
1.0E-07-
Si,
^^ ^ '
ai
Q.
c
g
1.0E-08-
/ ^ ^ . - • ' Nondripping
/ *
M t
J *
2
(D
2
1.0E-09-
1^
•
<8
1
»
CO
0)
(U
GC
1.0E-10
1.0E-11
1 nP-19
1 *
/ ^
/ «
/ «
0
3 1 ,000
2,000
3,000 4,000 5,000 6,000 7,000 8,000 9,000 10,000
Years after closure
Source Fof WAPDEG modeltng results: Moo (1999,
all).
Figure 1-51. Average fractional release rate of corrosion-resistant material (Alloy-22) for continually
dripping and nondripping conditions computed from WAPDEG modeling results for 400
simulated waste packages.
1-105
Environmental Consequences of Long-Term Repository Performance
10"-
— ■---._. i. i-^^-- ^^^^^
^ :
"""-vX
robabi
\ 5 km
a.
* ^^^
s
N \ 20 km Ni
CO
\ \ ' V
3
\ '. \
1 10-1-
\ 30 km-. \
o
£« :
\ ' '■ \
a
\ '. 1
'■ .
c
1 • •
erne
80 km \ ■.
\ 1 1 \
a.
■ \
\ ■ \
8
\ 1 \
10-2 J
10-5 10-4 10-3 10-2 10-1
10°
Concentration (milligram per liter)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-52. Complementary cumulative distribution function of mean peak groundwater concentrations
of chromium during 10,000 years following closure under high thermal load scenario with
Proposed Action inventory.
10° ^
— • — -^.^ — - .-: -^..^
^«
^" ~ - ^^\
\ * ^v.
■g
\ ■". 5 km
n
s
Q
'. s.^
Q.
^, 20'km\
<D
V *' \
>
^ \ v
«
\ \ V
3
1 10-1-
\ 30 km '.l
\ \ A
o
\ ', \
£>
"^
a
\ 'v \
E
\ \ i
«
, 1 1
Q.
\ ' V
E
1 I
5
\ \ \
80 km ' \-,
10-2 J
\ ) \'.
10-5 10-4 10-3 10-2 10-1
10°
Concentration (milligram per liter)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-53. Complementary cumulative distribution ftinction of mean peak groundwater concentrations
of chromium during 10,000 years following closure under intermediate thermal load
scenario with Proposed Action inventory.
1-106
Environmental Consequences of Long-Term Repository Performance
10"-
."•••••-. >.
.>
"^•^"■■■■■■••- \
2
^ ■•-- \
S
"". •. 5 km
1^
\ . -. V
<D
\ "^. •-. \
.^
» N 20 km \
cd
* '* \
"3
"^ \ •• \
1 10-^ -
\ 30km '•. \
o
^•s \ 1
2
\ 1 • I
1
80km ; '
1 \ ■••
(D
» ■
Q.
\ \ •
-in-2
\ Ml
10-5 10-^ 10-3 10-2 10-1
10°
Concentration (milligram per liter)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-54. Complementary cumulative distribution function of mean peak groundwater concentration
of chromium during 1 0,000 years following closure under low thermal load scenario with
Proposed Action inventory.
10" ■
8
Q.
i
I 10-1
I
o
E
o
o.
E
3
10-=
N
V.
80 km
-I -1 — I I I iiiii 1 — I I M iii| 1 — I I I r I iij 1 — I 1 M M!| f — I — r I I III
10-5 10-* 10-3 10-2 10-1 IQO
Concentration (milligram per liter)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-55. Complementary cumulative distribution function of mean peak groundwater concentration
of chromium during 10,000 years following closure under high thermal load scenario with
Inventory Module 1 .
1-107
Environmental Consequences of Long-Term Repository Performance
* r.n
10"-
v_ -., A.
' *■ ^"'^ ^^^v.
ty
' \
\
r^
■V *
\
\
(0
V
5 km
^ \ '•
s
Q.
^ . 20 km
N
O
V
1
^ \ ■••
\
ns
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10-5 10-" 10-3 .|o-2
10-'
10°
Concentration (milligram per liter)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-56. Complementary cumulative distribution function of mean peak groundwater concentration
of chromium during 10,000 years following closure under intermediate thermal load
scenario with Inventory Module 1 .
10°-
^ - -. "v;-^,^^^
2>
^•v '-.A.
bj
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o
\ \
\ \
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10-5 10" 10-3 10-2 10'
10°
Concentration (milligram per liter)
km = kilometer. To convert kilometers to miles, multiply by 0.62137.
Figure 1-57. Complementary cumulative distribution function of mean peak groundwater concentration
of chromium during 10,000 years following closure under low thermal load scenario with
Inventory Module 1 .
1-108
Environmental Consequences of Long-Term Repository Performance
10"-
:
^-^^^
-
^^\ 10,000 years
2
V Reference dose level
a.
X. 0.048 milligram per liter
Q)
^^
>
-
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^
3
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E
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o
-
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g-20 10-15 10-10 10-s loo
Concentration (milligram per liter)
Figure 1-58. Complementary cumulative distribution function of mean peak groundwater concentration
of elemental uranium in water at 5 kilometers (3 miles) during 10,000 years following
closure under high thermal load scenario with Proposed Action inventory.
i
Fraction of failed cladding
-
^
1"'1
-
1-6
1
1 1
1
0
50,000
100,000 150,000
200,000
250
,000
Years after repository closure
Figure 1-59. Fraction (patch area) of cladding that would fail using a zirconium-alloy corrosion rate
equal to 1.0 percent of that of Alloy-22.
1-109
Environmental Consequences of Long-Term Repository Performance
1-^
q-8
i-
CO
^
8-8
k_
ffi
Q.
(0
7-8
0)
3
o
6-8
o
o
F
5-8
i
4-8
0)
s
3-8
<B
2-8
10,000 20,000 30,000 40,000
Years after repository closure
50,000
Figure 1-60. Release rate of carbon- 14 from the repository to the ground surface.
I-llO
Environmental Consequences of Long-Term Repository Performance
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Appendix J
Transportation
Transportation
TABLE OF CONTENTS
Section Page
J.l Methods Used To Estimate Potential Impacts of National Transportation J-1
J.1.1 Analysis Approach and Methods J-1
J.l. 1.1 CALVIN J-5
J.1.1.2 HIGHWAY J-5
J.l. 1.3 INTERLINE J-6
J.l. 1.4 RADTRAN4 J-7
J.l. 1.5 RISKIND J-9
J.l. 2 Number and Routing of Shipments J-9
J.l. 2.1 Number of Shipments J-9
J.1.2.1.1 Commercial Spent Nuclear Fuel J-12
J.1.2.1.2 DOE Spent Nuclear Fuel and High-Level Radioactive Waste J-15
J. 1 .2. 1 .3 Greater-Than-Class-C and Special-Performance- Assessment-Required
Waste Shipments J-20
J. 1.2. 1.4 Sensitivity of Transportation Impacts to Number of Shipments J-21
J.l. 2.2 Transportation Routes J-23
J.1.2.2.1 Routes Used in the Analysis J-23
J.1.2.2.2 Routes for Shipping Rail Casks from Sites Not Served by a Railroad J-33
J.1.2.2.3 Sensitivity of Analysis Results to Routing Assumptions J-33
J.l. 3 Analysis of Impacts from Incident-Free Transportation J-34
J. 1.3.1 Methods and Approach for Analysis of Impacts for Loading Operations J-34
J.1.3.1.1 Radiological Impacts of Loading Operations at Commercial Sites J-35
J. 1 .3 . 1 .2 Radiological Impacts of DOE Spent Nuclear Fuel and High-Level
Radioactive Waste Loading Operations J-38
J. 1.3 .2 Methods and Approach for Analysis of Impacts from Incident-Free
Transportation J-39
J.1.3.2.1 Incident-Free Radiation Dose to Populations J-39
J. 1.3. 2.2 Methods Used To Evaluate Incident-Free Impacts to Maximally Exposed
Individuals J-43
J.1.3.2.2.1 Incident-Free Radiation Doses to Inspectors J-44
J.l. 3.2.2.2 Incident-Free Radiation Doses to Escorts J-46
J. 1.3 .2. 3 Vehicle Emission Impacts J-47
J. 1.3.2.4 Sensitivity of Dose Rate to Characteristics of Spent Nuclear Fuel J-48
J.1.4 Methods and Approach to Analysis of Accident Scenarios J-48
J. 1.4.1 Accidents in Loading Operations J-48
J.1.4. 1.1 Radiological Impacts of Loading Accidents J-48
J.1.4. 1.2 Industrial Safety Impacts of Loading Operations at Commercial Facilities J-50
J.1.4. 1.3 Industrial Safety Impacts of DOE Loading Operations J-51
J. 1.4.2 Transportation Accident Scenarios J-52
J.1.4.2.1 Radiological Impacts of Transportation Accidents J-52
J. 1 .4.2.2 Methods and Approach for Analysis of Nonradiological Impacts of
Transportation Accidents J-62
J.1.4.2.3 Data Used To Estimate Incident Rates for Rail and Motor Carrier Accidents J-64
J.1.4.2.3.1 Transportation Accident Reporting and Definitions J-64
J.l .4.2.3.2 Accident Rates for Transportation by Heavy-Combination Truck, Railcar,
and Barge in the United States J-69
J. 1 .4.2.3.3 Accident Data Provided by the States of Nevada, California, South
Carolina, Illinois, and Nebraska J-70
J.1.4. 2.4 Transportation Accidents Involving Nonradioactive Hazardous Materials J-72
J-iii
Transportation
i
J.2 Evaluation of Rail and Intermodal Transportation Options J-72
J.2. 1 Impacts of the Shipment of Commercial Spent Nuclear Fuel by Barge and
Heavy-Haul Truck from 19 Sites Not Served by a Railroad J-73
J.2. 1.1 Routes for Barges and Heavy-Haul Trucks J-73
J.2. 1 .2 Analysis of Incident-Free Impacts for Barge and Heavy-Haul Truck
Transportation J-73
J.2.1.2.1 Radiological Impacts of Incident-Free Transportation J-73
J.2. 1 .2.2 Nonradiological Impacts of Incident-Free Transportation (Vehicle
Emissions) J-80
J.2. 1 .3 Analysis of Impacts of Accidents for Barge and Heavy-Haul Truck
Transportation J-80
J.2. 1.3.1 Radiological Impacts of Accidents J-80
J.2. 1.3.2 Nonradiological Accident Risks J-82
J.2. 1.3.3 Maximum Reasonably Foreseeable Accidents J-82
J.2.2 Effects of Using Dedicated Trains or General Freight Service J-82
J.3 Nevada Transportation J-82
J.3.1 Transportation Modes, Routes, and Number of Shipments J-83
J.3. 1.1 Routes in Nevada for Legal-Weight Trucks J-83
J.3. 1.2 Routes in Nevada for Transporting Rail Casks J-84
J.3. 1.3 Sensitivity of Analysis Results to Routing Assumptions J-92
J.3.2 Analysis of Incident-Free Transportation in Nevada J-95
J.3. 3 Analysis of Transportation Accident Scenarios in Nevada J-95
J.3. 3.1 Intermodal Transfer Station Accident Methodology J-95
J.3. 4 Impacts in Nevada from Incident-Free Transportation for Inventory
Modules 1 and 2 J-98
J.3.4.1 Mostly Legal- Weight Truck Scenario J-98
J.3.4.2 Nevada Rail Implementing Alternatives J-98
J.3.4.3 Nevada Heavy-Haul Truck Implementing Alternatives J-98
J.3. 5 Impacts in Nevada from Transportation Accidents for Inventory Modules 1
and 2 J-101
J.3. 5.1 Mostly Legal- Weight Truck Scenario J-101
J.3.5.2 Nevada Rail Implementing Alternatives J-101
J.3. 5. 3 Nevada Heavy-Haul Truck Implementing Alternatives J-103
J.3.6 Impacts from Transportation of Other Materials J-105
J. 3. 6.1 Transportation of Personnel and Materials to Repository J-105
J.3. 6.2 Impacts of Transporting Wastes from the Repository J-108
J.3. 6.3 Impacts from Transporting Other Materials and People in Nevada for
Inventory Modules 1 and 2 J-109
J.3.6.4 Environmental Justice J-110
J.3. 6. 5 Summary of Impacts of Transporting Other Materials J-110
References J- 112
J-iv
Transportation
LIST OF TABLES
Table Paee
J-1 Summary of estimated numbers of shipments for the various inventory and
national transportation analysis scenario combinations J-10
J-2 Analysis basis — national and Nevada transportation scenarios J-11
J-3 Shipping cask configurations J-14
J-4 Anticipated receipt rate for spent nuclear fiiel and high-level radioactive waste at
the Yucca Mountain Repository J-1 5
J-5 Shipments of commercial spent nuclear fuel, mostly legal-weight truck scenario J-16
J-6 Shipments of commercial spent nuclear fuel, mostly rail scenario J-18
J-7 DOE spent nuclear fuel shipments by site J-20
J-8 Number of canisters of high-level radioactive waste and shipments from DOE
sites J-20
J-9 Commercial Greater-Than-Class-C waste shipments J-21
J-10 DOE Special-Performance-Assessment-Required waste shipments J-22
J-1 1 Highway distances for legal-weight truck shipments from commercial and DOE
sites to Yucca Mountain, mostly legal-weight truck transportation J-26
J- 12 Rail transportation distances from commercial and DOE sites to Nevada ending
rail nodes J-28
J-13 Barge transportation distances from sites to intermodal rail nodes J-34
J-14 Typical cesium- 137, actinide isotope, and total radioactive material content in a
rail shipping cask J-36
J- 1 5 Principal logistics bases and results for the reference at-reactor loading
operations J-37
J- 1 6 At-reactor reference loading operations — collective impacts to involved workers J-3 8
J- 1 7 Input parameters and parameter values used for the incident- free national truck
and rail transportation analysis J-40
J-18 Population within 800 meters of routes for incident- free transportation using
1990 census data J-40
J-19 Information used for analysis of incident- free transportation impacts J-41
J-20 Unit dose factors for incident- free national truck and rail transportation of spent
nuclear fuel and high-level radioactive waste J-42
J-21 Fractions of selected radionuclides in commercial spent nuclear fuel projected to
be released from casks in transportation accidents for cask response regions J-57
J-22 Fractions of selected radionuclides in aluminum and metallic spent nuclear fuel
projected to be released from casks in transportation accidents for cask response
regions J-58
J-23 Frequency of atmospheric and wind speed conditions - U.S. averages J-59
J-24 Annual probability of severe accidents in urbanized and rural areas - category 5
and 6 accidents, national fransportation J-61
J-25 Consequences of maximum reasonably foreseeable accidents in national
transportation J-63
J-26 National transportation distances from commercial sites to Nevada ending rail
nodes J-77
J-27 Barge shipments and ports J-78
J-28 Risk factors for incident- free heavy-haul truck and barge fransportation of spent
nuclear fuel and high-level radioactive waste J-79
J-29 Comparison of population doses and impacts from incident- free national
transportation for heavy-haul-to-rail, barge-to-rail, and legal-weight truck options J-79
J-v
Transportation
J-30 Population health impacts from vehicle emissions during incident-free national
transportation for mostly legal-weight truck scenario J-80
J-31 Conditional probabilities for barge transportation J-80
J-32 Food transfer factors used in the barge analysis J-81
J-33 Accident risks for shipping spent nuclear fuel from Turkey Point J-81
J-34 Comparison of general freight and dedicated train service J-83
J-3 5 Route characteristics for rail and heavy-haul truck implementing alternatives J-88
J-36 Populations in Nevada within 800 meters of routes J-88
J-37 Potential road upgrades for Cahente route J-89
J-38 Potential road upgrades for Caliente-Chalk Mountain route J-89
J-39 Potential road upgrades for Caliente-Las Vegas route J-90
J-40 Potential road upgrades for Apex/Dry Lake route J-90
J-41 Potential road upgrades for Sloan/Jean route J-90
J-42 Possible alignment variations of the Carlin corridor J-91
J-43 Possible alignment variations of the Cahente corridor J-91
J-44 Possible alignment variations of the CaHente-Chalk Mountain corridor J-92
J-45 Possible alignment variations of the Jean corridor J-92
J-46 Possible ahgnment variations of the Valley Modified corridor J-92
J-47 Nevada routing sensitivity cases analyzed for a legal-weight truck J-93
J-48 Comparison of impacts from the sensitivity analyses (national and Nevada) J-94
J-49 Screening analysis of external events considered potential accident initiators at
intermodal transfer station J-96
J-50 Projectile characteristics J-97
J-51 Results of aircraft projectile penetration analysis J-98
J-52 Population doses and radiological impacts from incident-free Nevada
fransportation for mostly legal-weight truck scenario - Modules 1 and 2 J-99
J-53 Population health impacts from vehicle emissions during incident-free Nevada
transportation for the mostly legal-weight truck scenario - Modules 1 and 2 J-99
J-54 Radiological and nonradiological impacts from incident-free Nevada
transportation for the mostly rail scenario - Modules 1 and 2 J-99
J-55 Collective worker doses from transportation of a single cask J-100
J-56 Doses and radiological health impacts to involved workers from intermodal
transfer station operations - Modules 1 and 2 J-100
J-57 Radiological and nonradiological health impacts from incident- free
fransportation for the heavy-haul truck implementing alternatives - Modules 1
and 2 J-101
J-58 Accident radiological health impacts for Modules 1 and 2 - Nevada
fransportation J-102
J-59 Rail corridor operation worker physical trauma impacts (Modules 1 and 2) J-102
J-60 Industrial health impacts from heavy-haul truck route operations (Modules 1
and 2 J-104
J-61 Annual physical trauma impacts to workers from intermodal fransfer station
operations (Module 1 or 2) J-104
J-62 Human health and safety impacts from shipments of material to the repository J- 1 06
J-63 Health impacts from transportation of construction and operations workers J- 1 07
J-64 Impacts of disposal container shipments for Proposed Action J-107
J-65 Annual amount of carbon monoxide emitted to Las Vegas Valley airshed from
transport of personnel and material to repository for the Proposed Action J-108
J-66 Shipments of waste from the Yucca Mountain Repository J-109
J-67 Impacts from transportation of materials, consumables, personnel, and waste for
Modules 1 and 2 J-110
J-vi
Transportation
J-68 Health impacts from transportation of materials, consumables, personnel, and
waste for the Proposed Action J-11 1
LIST OF FIGURES
Figure Page
J-1 Methods and approach for analyzing transportation radiological health risk J-3
J-2 Methods and approach for analyzing transportation nonradiological health risk J-4
J-3 Artist's conception of a truck cask on a legal-weight tractor- trailer truck J-13
J-4 Artist's concept of a large rail cask on a railcar J-13
J-5 Commercial and DOE sites and Yucca Mountain in relation to the U.S. Interstate
Highway System J-24
J-6 Commercial and DOE sites and Yucca Mountain in relation to the U.S. railroad
system J-25
J-7 Comparison of GA-4 cask dose rate and spent nuclear fuel bumup and cooling
time J-49
J-8 Probability matrix for mechanical forces and heat in transportation accidents J-56
J-9 Routes for barges from sites to nearby railheads J-74
J- 10 Potential Nevada routes for legal-weight truck shipments of spent nuclear fuel
and high-level radioactive waste to Yucca Mountain J-85
J-1 1 Potential Nevada rail routes to Yucca Mountain and approximate number of
shipments for each route J-86
J- 1 2 Nevada routes for heavy-haul truck shipments of spent nuclear fuel and high-
level radioactive waste to Yucca Mountain J-87
J-vii
Transportation
APPENDIX J. TRANSPORTATION
This appendix provides additional information for readers who wish to gain a better understanding of the
methods and analyses the U.S. Department of Energy (DOE) used to determine the human health impacts
of transportation for the Proposed Action and Inventory Modules 1 and 2 discussed in this environmental
impact statement (EIS). The materials included in Module 1 are the 70,000 metric tons of heavy metal
(MTHM) for the Proposed Action and additional quantities of spent nuclear fuel and high-level
radioactive waste that DOE could dispose of in the repository as part of a reasonably foreseeable future
action. The materials included in Module 2 include the materials in Module 1 and other highly
radioactive materials. Appendix A describes materials included in Modules 1 and 2. This appendix also
provides the information DOE used to estimate traffic fatalities that would be associated with the long-
term maintenance of storage facilities at 72 commercial sites and 5 DOE sites.
The appendix describes the key data and assumptions DOE used in the analyses and the analysis tools and
methods the Department used to estimate impacts of loading operations at 72 commercial and 5 DOE
sites; incident-free transportation by highway, rail and barge; intermodal transfer; and transportation
accidents. The references listed at the end of this appendix contain additional information.
This appendix presents information on analyses of the impacts of national transportation and on analyses
of the impacts that could occur in Nevada. Section J.l presents information on the analysis of
occupational and public health and safety impacts for the transportation of spent nuclear fuel and high-
level radioactive waste from the 77 sites to the repository. Section J.2 presents information on the
analysis of rail and intermodal transportation options. Section J.3 presents information on the analysis of
transportation in Nevada. Section J.4 presents a summary assessment of the Nevada transportation
implementing alternatives.
J.1 Methods Used To Estimate Potential Impacts of
National Transportation
This section provides information on the methods and data DOE used to estimate impacts from shipping
spent nuclear fuel and high-level radioactive waste from 72 commercial sites and 5 DOE sites throughout
the United States to the Yucca Mountain Repository.
MOSTLY LEGAL-WEIGHT TRUCK AND MOSTLY RAIL SCENARIOS
The Department does not anticipate that either the mostly legal-weight truck or the mostly rail
scenario represents the actual mix of truck or rail transportation modes it would use. Nonetheless,
DOE used these scenarios as a basis for the analysis of potential impacts to ensure the analysis
addressed the range of possible transportation impacts. Thus, the estimated numbers of shipments
for the mostly legal-weight truck and mostly rail scenarios represent only the two extremes in the
possible mix of transportation modes. Therefore, the analysis provides estimates that cover the
range of potential impacts to human health and safety and to the environment for the transportation
modes DOE could use for the Proposed Action.
J.1.1 ANALYSIS APPROACH AND METHODS
Three types of impacts could occur to the public and workers from transportation activities associated
with the Proposed Action. These would be a result of the transportation of spent nuclear fuel and high-
J-1
Transportation
level radioactive waste and of the personnel, equipment, materials, and supplies needed to construct,
operate and monitor, and close the proposed Yucca Mountain Repository. The first type, radiological
impacts, would be measured by radiological dose to populations and individuals and the resulting
estimated number of latent cancer fatalities that would be caused by radiation from shipments of spent
nuclear fuel and high-level radioactive waste from the 77 sites under normal and accident transport
conditions. The second and third types would be nonradiological impacts — fatalities caused by vehicle
emissions and fatalities caused by vehicle accidents. The analysis also estimated impacts due to the
characteristics of hazardous cargoes from accidents during the transportation of nonradioactive hazardous
materials to support repository construction, operation and monitoring, and closure. For perspective,
about 10 fatalities resulting from hazardous material occur each year during the transportation of more
than 300 million shipments of hazardous materials in the United States (DOT 1998a, Table 1). Therefore,
DOE expects that the risks from exposure to hazardous materials that could be released during shipments
to and from the repository sites would be very small (see Section J. 1.4.2.4). The analysis evaluated the
impacts of traffic accidents and vehicle emissions arising from these shipments.
The analysis used a step-wise process to estimate impacts to the public and workers. The process used
the best available information from various sources and computer programs and associated data to
accomplish the steps. Figures J-1 and J-2 show the steps followed in using data and computer programs.
DOE has determined that the computer programs identified in the figure are suitable, and provide results
in the appropriate measures, for the analysis of impacts performed for this EIS.
The CALVIN computer program (TRW 1998, all) is used to estimate the numbers of shipments of spent
nuclear fuel from commercial sites. This program uses information on spent nuclear fuel stored at each
site and an assumed scenario for picking up the spent fuel from each site. The program also uses
information on the capacity of shipping casks that could be used.
The HIGHWAY computer program (Johnson et al. 1993a, all) is a routing tool used to select existing
highway routes that would satisfy Department of Transportation route selection regulations and that DOE
could use to ship spent nuclear fuel and high-level radioactive waste from the 77 sites to the repository.
The I>rrERLINE computer program (Johnson et al. 1993b, all) is a routing tool used to select existing rail
routes that railroads would be likely to use to ship spent nuclear fuel and high-level radioactive waste
from the 77 sites to the repository.
The RADTRAN4 computer program (Neuhauser and Kanipe 1992, all) is used to estimate the
radiological dose risks to populations and transportation workers of incident-free transportation and to the
general population from accident scenarios. For the analysis of incident-free transportation risks, the code
uses scenarios for persons who would share transportation routes with shipments — called onlink
populations, persons who live along the route of travel — offlink populations, and persons exposed at
stops. For accident risks, the code evaluates the range of possible accident scenarios from high
probability and low consequence to low probability and high consequence.
The RISKIND computer program (Yuan et al. 1995, all) is used to estimate radiological doses to
maximally exposed individuals for incident-free transportation and to populations and maximally exposed
individuals for accident scenarios. To estimate incident-free doses to maximally exposed individuals,
RISKIND uses geometry to calculate the dose rate at specified locations that would arise from a source of
radiation. RISKIND is also used to calculate the radiation dose to a population and hypothetical
maximally exposed individuals from releases of radioactive materials that are postulated to occur in
maximum reasonably foreseeable accident scenarios.
The following sections describe these programs in detail.
J-2
Transportation
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Transportation
DOSE RISK
Dose risk is a measure of radiological impacts to populations - public or workers - from the potential
for exposure to radioactive materials. Thus, a potential of 1 chance in 1,000 of a population
receiving a collective dose of 1 rem (1 person-rem) from an accident would result in a dose risk of
0.001 person-rem (0.001 is the product of 1 person-rem and the quotient of 1 over 1 ,000). Dose risk
is often expressed in units of latent cancer fatalities.
The use of dose risk to measure radiological impacts allows a comparison of alternatives with
differing characteristics in terms of radiological consequences that could result and the likelihood
that the consequences would actually occur.
J.1.1.1 CALVIN
The Civilian Radioactive Waste Management System Analysis and Logistics Visually Interactive
(CALVIN) model (TRW 1998, all) was developed to be a planning tool to estimate the logistic and cost
impacts of various operational assumptions for accepting radioactive wastes. CALVIN is used in
transportation modeling to determine the number of shipments of commercial spent nuclear fuel from
each reactor site. The parameters that the CALVIN model used to determine commercial spent nuclear
fuel movement include the shipping cask specifications including heat limits, kinfimty (measure of
criticality) limits for the contents of the casks, capacity (assemblies or canisters/cask), bumup/enrichment
curves, and cooling time for the fuel being shipped.
The source data used by CALVIN for commercial spent nuclear fuel projections include the RW-859
historic data collected by the Energy Information Administration, and the corresponding projection
produced based on current industry trends for commercial fuel (see Appendix A). This EIS used
CALVIN to estimate commercial spent nuclear fuel shipment numbers based on the cask capacity (see
Section J. 1.2) and the shipping cask handling capabilities at each site. For the mostly rail national
transportation scenario, CALVIN assumed that shipments would use the largest cask a site would be
capable of handling. In some cases, CALVIN estimated that the characteristics of the spent nuclear fuel
that would be picked up at a site would exceed the capabilities of the largest cask if the cask was fully
loaded. In such cases, to provide a realistic estimate of the number of shipments that would be made, the
program derated (reduced the capacity of) the casks. The reduction in capacity was sufficient to
accommodate the characteristics of the spent nuclear fuel the program estimated for pickup at the site.
J.I. 1.2 HIGHWAY
The HIGHWAY computer program (Johnson et al. 1993a, all) was used to select highway routes for the
analysis of impacts presented in this EIS. HIGHWAY calculates routes by minimizing the total
impedance between the origin and the destination. The impedance is determined by distance and driving
time along a particular segment of highway. Using Rand McNally route data and rules that apply to
carriers of Highway Route-Controlled Quantities of Radioactive Materials (49 CFR 397.101),
HIGHWAY selected highway routes for legal-weight truck shipments from each commercial and DOE
site to the Yucca Mountain site. In addition, DOE used this program to estimate the populations within
800 meters (0.5 mile) of the routes it selected. These population densities were used in calculating
incident-free radiological risks to the public along the routes.
One of the features of the HIGHWAY model is its ability to estimate routes for the transport of Highway
Route-Controlled Quantities of Radioactive Materials. The Department of Transportation has established
a set of routing regulations for the transport of these materials (49 CFR 397. 101). Routes following these
J-5
Transportation
regulations are frequently called HM-164 routes. The regulations require the transportation of these
shipments on preferred highways, which include:
• Interstate highways
• An Interstate System bypass or beltway around a city
• State-designated preferred routes
State routing agencies can designate preferred routes as an alternative to, or in addition to, one or more
Interstate highways. In making this determination, the state must consider the safety of the alternative
preferred route in relation to the Interstate route it is replacing, and must register all such designated
preferred routes with the Department of Transportation.
Frequently, the origins and destinations of Highway Route-Controlled Quantities of Radioactive
Materials are not near Interstate highways. In general, the Department of Transportation routing
regulations require the use of the shortest route between the pickup location to the nearest preferred route
entry location and the shortest route to the destination from the nearest preferred route exit location. In
general, HM-164 routes tend to be somewhat longer than other routes; however, the increased safety
associated with Interstate highway travel is the primary purpose of the routing regulations.
Because many factors can influence the time in transit over a preferred route, a carrier of Highway Route-
Controlled Quantities of Radioactive Materials must select a route for each shipment. Seasonal weather
conditions, highway repair or construction, highways that are closed because of natural events (for
example, a landslide in North Carolina closed Interstate 40 near the border with Tennessee from June
until November 1997), and other events (for example, the 1996 Olympic Games in Atlanta, Georgia) are
all factors that must be considered in selecting preferred route segments to reduce time in transit. For this
analysis, the highway routes were selected by the HIGHWAY program using an assumption of normal
travel and without consideration for factors such as seasons of the year or road construction delays.
Although these shipments could use other routes, DOE considers the impacts determined in the analyses
to be representative of other possible routings that would also comply with Department of Transportation
regulations. Specific route mileages for truck transportation are presented in Section J. 1.2. 1. 1.
In selecting existing routes for use in the analysis, the HIGHWAY program determined the length of
travel in each type of population zone — rural, suburban, and urban. The program characterized rural,
suburban, and urban population areas according to the following breakdown: rural population densities
range from 0 to 54 persons per square kilometer (0 to 140 persons per square mile); the suburban range is
55 to 1,300 persons per square kilometer (140 to 3,300 persons per square mile); and urban is all
population densities greater than 1,300 persons per square kilometer (3,3(X) persons per square mile). The
population densities along a route used by the HIGHWAY program are derived from 1990 data from the
Bureau of the Census.
J.I .1.3 INTERLINE
Shipments of radioactive materials by rail are not subject to route restrictions imposed by regulations.
For general freight rail service, DOE anticipates that railroads would route shipments of spent nuclear fuel
and high-level radioactive waste to provide expeditious travel and the minimum practical number of
interchanges between railroads. The selection of a route determines the potentially exposed population
along the route as well as the expected frequency of transportation-related accidents. The analysis used
the INTERLINE computer program (Johnson et al. 1993b, all) to project the railroad routes that DOE
would use to ship spent nuclear fuel and high-level radioactive waste from the sites to the Yucca
Mountain site. Specific routes were projected for each originating generator with the exception of 9 that
do not have capability to handle or load a rail transportation cask (see Section J.1.2.1.1, Table J-6).
J-6
Transportation
INTERLINE computes rail routes based on rules that simulate historic routing practices of U.S. railroads.
The INTERLINE data base consists of 94 separate subnetworks and represents various competing rail
companies in the United States. The data base, which was originally based on data from the Federal
Railroad Administration and reflected the U.S. railroad system in 1974, has been expanded and modified
extensively over the past two decades. The program is updated periodically to reflect current track
conditions and has been benchmarked against reported mileages and observations of commercial rail
firms. The program also provides an estimate of the population within 800 meters (0.5 mile) of the routes
it selected. This population estimate was used to calculate incident-free radiological risk to the public
along the routes selected for analysis.
In general, rail routes are calculated by minimizing the value of a factor called impedance between the
origin and the destination. The impedance is determined by considering trip distance along a route, the
mainline classification of the rail lines that would be used, and the number of interchanges that would
occur between different railroad companies involved. In general, impedance determined by the
INTERLINE program:
• Decreases as the distance traveled decreases
• Is reduced by use of mainline track that has the highest traffic volume (see below)
• Is reduced for shipments that involve the fewest number of railroad companies
Thus, routes that are the most direct, that use high-traffic volume mainline track, and that involve only
one railroad company would have the lowest impedance. The most important of these characteristics
from a routing standpoint is the mainline classification, which is the measure of traffic volume on a
particular link. The mainline classifications used in the INTERLINE routing model are as follows:
• A - mainline - more than 20 million gross ton miles per year
• B - mainline - between 5 and 20 million gross ton miles per year
• A - branch line - between 1 and 5 million gross ton miles per year
• B - branch line - less than I million gross ton miles per year
The INTERLINE routing algorithm is designed to route a shipment preferentially on the rail lines having
the highest traffic volume. Frequently traveled routes are preferred because they are generally well
maintained because the railroad depends on these lines for a major portion of its revenue. In addition,
routing along the high-traffic lines usually replicates railroad operational practices.
The population densities along a route were derived from 1990 data from the Bureau of the Census, as
described above for the HIGHWAY computer program.
DOE anticipates that routing of rail shipments in dedicated (special) train service, if used, would be
similar to routing of general freight shipments for the same origin and destination pairs. However,
because cask cars would not be switched between trains at classification yards, dedicated train service
would be likely to result in less time in transit.
J.1.1.4 RADTRAN4
The RADTRAN4 computer program (Neuhauser and Kanipe 1992, all) was used for the routine and
accident cargo-related risk assessment to estimate the radiological impacts to collective populations.
RADTRAN4 was developed by Sandia National Laboratories to calculate population risks associated
with the transportation of radioactive materials by a variety of modes, including truck, rail, air, ship, and
barge. The code has been used extensively for transportation risk assessment since it was issued in the
late 1970s and has been reviewed and updated periodically. In 1995, a validation of the RADTRAN4
J-7
Transportation
code demonstrated that it yielded acceptable results (Maheras and Pippen 1995, page iii). In the context
of the validation analysis, acceptable results means that the difference between the estimates generated by
the RADTRAN4 code and hand calculations were small, that is, less than 5 percent (Maheras and Pippen
1995, page 3-1).
The RADTRAN4 calculations for routine (or incident-free) dose are based on expressing the dose rate as
a function of distance from a point source. Associated with the calculation of routine doses for each
exposed population group are parameters such as the radiation field strength, the source-receptor distance,
the duration of the exposure, vehicular speed, stopping time, traffic density, and route characteristics such
as population density. In calculating population doses from incident-free transportation, the RADTRAN4
program used population density data provided by the HIGHWAY and INTERLINE computer programs.
These data are based on the 1990 Census.
In addition to routine doses, RADTRAN4 was used to estimate dose risk from a spectrum of accident
scenarios. The spectrum of accident scenarios encompass the range of possible accidents, including low-
probability accident scenarios that have high consequences, and high-probability accident scenarios that
have low consequences (fender benders). The RADTRAN4 calculation of collective accident risk for
populations along routes employed models that quantified the range of potential accident severities and
the responses of the shipping casks to the accident scenarios. The spectrum of accident severity was
divided into categories. Each category of severity received a conditional probability of occurrence; that
is, the probability that an accident will be of a particular severity if an accident occurs — the more severe
the accident, the more remote the chance of such an accident. A release fraction, which is the fraction of
the material in a shipping cask that could be released in an accident, is assigned to each accident scenario
severity category on the basis of the physical and chemical form of the material being transported. The
model also takes into account the mode of transportation, the state-specific accident rates, and population
densities for rural suburban, and urban population zones through which shipments would pass to estimate
accident risks for this analysis. The RADTRAN4 program used actual population densities within
800 meters (0.5 mile) of transportation routes based on 1990 census data as the basis for estimating
populations within 80 kilometers (50 miles).
For accident scenarios involving the release of radioactive material, RADTRAN4 assumes that the
material is dispersed in the environment as described by a Gaussian dispersion model. The dispersion
analysis assumes that meteorological conditions are national averages for wind speed and atmospheric
stability. For the risk assessment, the analysis used these meteorological conditions and assumed an
instantaneous ground-level release and a small diameter source cloud (Neuhauser and Kanipe 1993,
page 5-6). The calculation of the collective population dose following the release and the dispersal of
radioactive material includes the following exposure pathways:
External exposure to the passing radioactive cloud
External exposure to contaminated ground
Internal exposure from inhalation of airborne contaminants
Internal exposure from ingestion of contaminated food
For the ingestion pathway, the analysis used state-specific food transfer factors (TRW 1999a, page 35),
which relate the amount of radioactive material ingested to the amount deposited on the ground, as input
to the RADTRAN4 code. Radiation doses from the ingestion or inhalation of radionuclides were
calculated by using standard dose conversion factors from Federal Guidance Reports No. 1 1 and 12
(TRW 1999a, page 36).
J-8
Transportation
J.1.1.5 RISKIND
The RISKIND computer program (Yuan et al. 1995, all) was used as a complement to the RADTRAN4
calculations to estimate scenario-specific doses to maximally exposed individuals for both routine
operations and accident conditions and to estimate population impacts for the assessment of accident
scenario consequences. The RISKIND code was originally developed for the DOE Office of Civilian
Radioactive Waste Management specifically to analyze radiological consequences to individuals and
population subgroups from the transportation of spent nuclear fuel and is used now to analyze the
transport of other radioactive materials, as well as spent nuclear fuel.
The RISKIND external dose model considers direct external exposure and exposure from radiation
scattered from the ground and air. RISKIND was used to calculate the dose as a function of distance from
a shipment on the basis of the dimensions of the shipment (millirem per hour for stationary exposures and
millirem per event for moving shipments). The code approximates the shipment as a cylindrical volume
source, and the calculated dose includes contributions from secondary radiation scatter from buildup
(scattering by material contents), cloudshine (scattering by air), and groundshine (scattering by the
ground). Credit for potential shielding between the shipment and the receptor was not considered.
The RISKIND code was also used to provide a scenario-specific assessment of radiological consequences
of severe transportation-related accidents. Whereas the RADTRAN4 risk assessment considers the entire
range of accident severities and their related probabilities, the RISKIND consequence assessment focuses
on accident scenarios that result in the largest releases of radioactive material to the environment. The
consequence assessment was intended to provide an estimate of the potential impacts posed by a severe,
but highly unlikely, transportation-related accident scenario.
The dose to each maximally exposed individual considered was calculated with RISKIND for an
exposure scenario defined by a given distance, duration, and frequency of exposure specific to that
receptor. The distances and durations were similar to those given in previous transportation risk
assessments. The scenarios were not meant to be exhaustive but were selected to provide a range of
potential exposure situations.
J.1.2 NUMBER AND ROUTING OF SHIPMENTS
This section discusses the number of shipments and routing information used to analyze potential impacts
that would result from preparation for and conduct of transportation operations to ship spent nuclear fuel
and high-level radioactive waste to the Yucca Mountain site. Table J-1 summarizes the estimated
numbers of shipments for the various inventory and national shipment scenario combinations.
J.1 .2.1 Number of Shipments
DOE used two analysis scenarios — mostly legal-weight truck and mostly train (rail) — as bases for
estimating the number of shipments of spent nuclear fuel and high-level radioactive waste from 72
commercial and 5 DOE sites. The number of shipments for the scenarios was used in analyzing
transportation impacts for the Proposed Action and Inventory Modules 1 and 2. DOE selected the
scenarios because, more than 10 years before the projected start of operations at the repository, it cannot
accurately predict the actual mix of rail and legal-weight truck transportation that would occur from the
77 sites to the repository. Therefore, the selected scenarios enable the analysis to bound (or bracket) the
ranges of legal-weight truck and rail shipments that could occur.
J-9
Transportation
Table J-1. Summary of estimated numbers of shipments for the various inventory and national
transportation analysis scenario combinations.
Mosdy
truck
Mostly
rail
Truck
Rail
Truck
Rail
Proposed Action
Commercial spent nuclear fuel
37,738
0
2,601
8,386
High-level radioactive waste
8,315
0
0
1,663
Spent nuclear fuel
3,470
300
0
766
Greater-Than-Class-C waste
0
0
0
0
Special-Performance-Assessment-Required waste
0
0
0
0
Proposed Action totals
49,523
300
2,601
10,815
Module 1"
Commercial spent nuclear fuel
66,850
0
3,701
13,906
High-level radioactive waste
22,280
0
0
4,456
Spent nuclear fuel
3,721
300
0
797
Greater-Than-Class-C waste
0
0
0
0
Special-Performance- Assessment-Required waste
0
0
0
0
Module 1 totals
92,851
300
3,701
19,159
Module 2"
Commercial spent nuclear fuel
66,850
0
3,701
13,906
High-level radioactive waste
22,280
0
0
4,456
Spent nuclear fuel
3,721
300
0
797
Greater-Than-Class-C waste
1,096
0
0
282
Special-Performance- Assessment-Required waste
2,010
0
0
404
Module 2 totals
95,957
300
3,701
19,845
a. The number of shipments for Module 1 includes all shipments of spent nuclear fuel and high-level radioactive waste
included in the Propxjsed Action and shipments of additional sf)ent nuclear fuel and high-level radioactive waste as described
in Appendix A. The number of shipments for Module 2 includes all the shipments in Module 1 and additional shipments of
highly radioactive materials described in Appendix A.
The analysis estimated the number of shipments from commercial sites where spent nuclear fuel would be
loaded and shipped and from DOE sites where spent nuclear fuel, naval spent nuclear fuel, and high-level
radioactive waste would be loaded and shipped.
For the mostly legal-weight truck scenario, with one exception, shipments were assumed to use legal-
weight trucks. Overweight, overdimensional trucks weighing between about 36,300 and 52,300
kilograms (80,000 and 1 15,000 pounds) but otherwise similar to legal-weight trucks could be used for
some spent nuclear fuel and high-level radioactive waste (for example, spent nuclear fuel from the South
Texas reactors). The exception that gives the scenario its name — mostly legal-weight truck — was for
shipments of naval spent nuclear fuel. Under this scenario, naval spent nuclear fuel would have to be
shipped by rail because of the size and weight of the shipping container (cask) that would be used.
For the mostly rail scenario, the analysis assumed that all sites would ship by rail, with the exception of
those with physical limitations that would make rail shipment impractical. The exception would be for
shipments by legal-weight trucks from 9 commercial sites that do not have the capability to load rail
casks. The analysis assumed that 19 commercial sites that do not have direct rail service but that could
handle large casks would ship by barge or heavy-haul truck to nearby railheads with intermodal
capability.
J-10
Transportation
For commercial spent nuclear fuel, the CALVIN code was used to compute the number of shipments.
The number of shipments of DOE spent nuclear fuel and high-level radioactive waste was estimated
based on the data in Appendix A and information provided by the DOE sites. The numbers of shipments
were estimated based on the characteristics of the materials shipped, mode interface capability (for
example, the lift capacity of the cask-handling crane) of each shipping facility, and the modal-mix case
analyzed. Table J-2 summarizes the basis for the national and Nevada transportation impact analysis.
Table J-2. Analysis basis — national and Nevada transportation scenarios.
a,b
Material
Mostly legal-weight truck
scenario national and
Nevada
National mostly rail scenario
Nevada rail scenario
Nevada heavy-haul truck
scenario
Casks
Commercial SNF
Truck casks - about 1.8
MTHM per cask
DOE HLW and DOE
SNF, except naval
SNF
Naval SNF
Transportation modes
Commercial SNF
Truck casks - 1 SNF or
HLW canister per cask
Disposal canisters in large
rail casks for shipment from
INEEL
Legal-weight trucks
DOE HLW and DOE
SNF, except naval
SNF
Naval SNF
Legal-weight trucks
Rail from INEEL to
intermodal transfer station in
Nevada, then heavy-haul
trucks to repository
Rail casks - 6 to 12 MTHM
per cask for shipments from
63 sites
Truck casks - about 1.8
MTHM per cask for
shipments from 9 sites
Rail casks - four to nine
SNF or HLW canisters per
cask
Disposable canisters in large
rail casks for shipments from
INEEL
Direct rail from 44 sites
served by railroads to
repository
Heavy-haul trucks from 5
sites to railhead, then rail to
repository
Heavy-haul trucks or barges"
from 14 sites to railhead,
then rail to repository
Legal-weight trucks from
9 sites to ref)ository
Rail from DOE sites'* to
repository
Rail from INEEL to
repository
Rail casks - 6 to 12 MTHM per
cask for shipments from 63 sites
Truck casks - about 1.8 MTHM
per cask for shipments from 9 sites
Rail casks - four to nine SNF or
HLW canisters per cask
Disposable canisters in large rail
casks for shipments from INEEL
Rail from 44 sites served by
railroads to intermodal transfer
station in Nevada, then heavy-haul
trucks to repository
Heavy-haul trucks from 5 sites to
railheads, then rail to intermodal
transfer station in Nevada, then
heavy-haul trucks to repository
Heavy-haul trucks or barges from
14 sites to railheads, then rail to
intermodal transfer station in
Nevada, then heavy-haul trucks to
repository'
Legal-weight trucks from 9 sites to
repository
Rail from DOE sites to intermodal
transfer station in Nevada, then
heavy-haul trucks to repository
Rail from INEEL to intermodal
transfer station in Nevada, then
heavy-haul trucks to repository
a. Abbreviations: SNF = spent nuclear fuel; MTHM = metric tons of heavy metal; HLW = high-level radioactive waste;
INEEL = Idaho National Engineering and Environmental Laboratory.
b. G. E. Morris facility is included with the Dresden reactor facilities in the 72 commercial sites.
c. Fourteen of 19 commercial sites not served by a railroad are on or near a navigable waterway. Some of these 14 sites could
ship by barge rather than by heavy-haul truck to a nearby railhead.
d. Hanford Site, Savannah River Site, Idaho National Engineering and Environmental Laboratory, West Valley Demonstration
Project, and Ft. St. Vrain.
J-11
Transportation
Detailed descriptions of spent nuclear fuel and high-level radioactive waste that would be shipped to the
Yucca Mountain site are presented in Appendix A.
J.I .2.1 .1 Commercial Spent Nuclear Fuel
For the analysis, the CALVIN model used 32 shipping cask configurations: 15 for legal-weight truck
casks (Figure J-3) and 17 for rail casks (Figure J-4). Table J-3 lists the legal-weight truck and rail cask
configurations used in the analysis and their capacities. The analysis assumed that all shipments would
use one of the 32 configurations. If the characteristics of the spent nuclear fuel projected for shipment
exceeded the capabilities of one of the casks, the model reduced the cask's capacity for the affected
shipments. The reduction, which is sometimes referred to as cask derating, was needed to satisfy nuclear
criticality, shielding, and thermal constraints. For shipments that DOE would make using specific casks,
derating would be accomplished by partially filling the assigned casks in compliance with provisions of
applicable Nuclear Regulatory Commission certificates of compliance. An example of derating is
discussed in Section 5 of the GA-4 legal-weight truck shipping cask design report (General Atomics
1993, page 5.5-1). The analysis addresses transport of two high-bumup or short cooling time pressurized-
water reactor assemblies rather than four design basis assemblies.
RAIL SHIPMENTS
This appendix assumes that rail shipments of spent nuclear fuel would use large rail shipping casks,
one per railcar. DOE anticipates that as many as five railcars with casks containing spent nuclear
fuel or high-level radioactive waste would move together in individual trains with buffer cars and
escort cars. For general freight service, a train would include other railcars with other materials. In
dedicated (or special) service, trains would move only railcars containing spent nuclear fuel or high-
level radioactive waste and the buffer and escort cars.
For the mostly rail scenario, 9 sites without sufficient crane capacity to lift a rail cask or without other
factors such as sufficient floor loading capacity or ceiling height were assumed to ship by legal-weight
truck. The 19 sites with sufficient crane capacity but without direct rail access were assumed to ship by
heavy -haul truck to the nearest railhead. Of these 19 sites, 14 with access to navigable waterways were
analyzed for shipping by barge to a railhead (see Section J.2.1). The number of rail shipments (direct or
indirect) was estimated based on each site using the largest cask size feasible based on the load capacity
of its cask handling crane. In calculating the number of shipments from the sites, the model used the
DOE allocation of delivery rights (10 CFR Part 961) to the sites and the anticipated receipt rate at the
repository listed in Table J-4. Using CALVIN, the number of shipments of legal-weight truck casks
(Figure J-3) of commercial spent nuclear fuel estimated for the Proposed Action (63,000 MTU of
commercial spent nuclear fuel) for the mostly legal-weight truck scenario, would be about 14,000
containing boiling-water reactor assemblies and 24,000 containing pressurized-water reactor assemblies.
Under Inventory Modules 1 and 2, for which approximately 105,000 MTU of commercial spent nuclear
fuel would be shipped to the repository (see Appendix A), the estimated number of shipments for the
mostly legal-weight truck scenario would be 24,000 for boiling-water reactor spent nuclear fuel and
43,000 for pressurized-water reactor spent nuclear fuel. Table J-5 lists the number of shipments of
commercial spent nuclear fuel for the mostly legal-weight truck scenario. Specifically, it lists the site,
plant, and state where shipments would originate, the total number of shipments from each site, and the
type of spent nuclear fuel that would be shipped. A total of 72 commercial sites with 104 plants (or
facilities) are listed in the table.
J-12
Transportation
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J-13
Transportation
Table J-3. Shipping cask configurations.
Shipping casks
Capacity (number of spent
nuclear fuel assemblies)
Description"'
Rail
B-RAIL-LGSP
B-RAIL-SMSP
BP-TRAN-OVLG74
B-TRAN-OVLG
B-TRAN-OVMED
B-TRAN OVSM
B-High Heat Rail
P-RAIL-LGSP
P-RAIL-SMSP
P-RAIL-MOX
P-RL-LGSP-ST
P-TRAN-OVLG-YR
P-TRAN-OVLG
P-TRAN-OVMED
P-TRAN-OVSM
P-TRNST-OVLG
P-High Heat-Rail
Truck
B-LWT-GA9I
B-LWT-GA9II
B-LWT-GA9III
B-LWT-GA9IV
B-LWT-GAV
BP-LWT-GA4I
B-NLI-1/2
P-LWT-GA4I
P-LWT-GA4II
P-LWT-GA4III
P-LWT-GA4I-ST
P-LWT-GA4II-ST
P-LWT-GA4III-ST
P-NLI-1/2
P-LWT-MOX
61
24
74
61
44
24
17
26
12
9
12
36
24
21
12
12
7
9
7
5
4
2
4
2
4
3
2
4
3
2
1
4
Large BWR single-purpose shipping container
Small BWR single-purpose shipping container
Big Rock Point dual-purpose shipping container
Large BWR dual-purpose shipping container
Medium BWR dual-purpose shipping container
Small BWR dual-purpose shipping container
BWR high heat shipping container
Large PWR single-purpose shipping container
Small PWR single-purpose shipping container
Mixed-oxide SNF shipping container
South Texas single-purpose shipping container
Yankee Rowe dual-purpose shipping container
Large PWR dual-purpose shipping container
Medium PWR dual-purpose shipping container
Small PWR dual-purpose shipping container
South Texas dual-purpose shipping container
PWR high heat shipping container
Primary BWR shipping container
Derated BWR shipping container
Derated BWR shipping container
Derated BWR shipping container
Derated BWR shipping container
Big Rock Point shipping container
Secondary BWR shipping container
Primary PWR shipping container
Derated PWR shipping container
Derated PWR shipping container
South Texas shipping container
Derated South Texas shipping container
Derated South Texas shipping container
Secondary PWR shipping container
Mixed-oxide SNF shipping container
a. Source: TRW (1999a, page 3).
b. BWR = boiling-water reactor; PWR = pressurized-water reactor; SNF = Sfwnt nuclear fuel.
The number of shipments of truck and rail casks (Figure J-4) of commercial spent nuclear fuel estimated
for the Proposed Action for the mostly rail scenario would be 4,200 for boiling-water reactor spent
nuclear fuel and 6,800 for pressurized-water reactor spent nuclear fuel. Under Modules 1 and 2, the
estimated number of shipments for the mostly rail scenario would be 6,5(X) containing boiling-water
reactor spent nuclear fuel and 1 1,100 containing pressurized-water reactor spent nuclear fuel. Table J-6
lists the number of shipments for the mostly rail scenario. It also lists the site and state where shipments
would originate, the total number of shipments from each site, the size of rail cask assumed for each site,
and the type of spent nuclear fuel that would be shipped. In addition, it lists the 19 sites not served by a
railroad that would ship rail casks by barge or heavy-haul trucks to a nearby railhead and the 9
commercial sites without capability to load a rail cask.
J- 14
Transportation
Table J-4. Anticipated receipt rate for spent nuclear fuel and high-level radioactive waste at the Yucca
Mountain Repository^
High-level radioactive waste and DOE spent
Commercial
mthm"
spent nuclear fuel annual receipt''
Shipments
nuclear fuef annual i
receipts
MTHM
Shipments
Year
Mostly LWr
Mostly rail
Mostly LWT
Mostly rail
2010
300
267
100
0
0
0
2011
600
413
184
0
0
0
2012
1,200
757
294
0
0
0
2013
2,000
1,246
478
0
0
0
2014
3,000
1,805
663
0
0
0
2015
3,000
1,792
638
400
650
140
2016
3,000
1,797
600
400
650
140
2017
3,000
1,803
555
400
650
140
2018
3,000
1,787
497
400
650
140
2019
3,000
1,782
508
400
650
140
2020
3,000
1,773
501
400
650
140
2021
3,000
1,780
514
400
650
140
2022
3,000
1,771
513
400
650
140
2023
3,000
1,772
484
400
650
140
2024
3,000
1,796
496
400
650
140
2025
3,000
1,779
472
400
650
140
2026
3,000
1,777
437
400
650
140
2027
3,000
1,793
488
400
650
140
2028
3,000
1,772
469
400
650
140
2029
3,000
1,794
460
400
650
140
2030
3,000
1,768
419
400
675
140
2031
3,000
1,808
451
400
685
140
2032
3,000
1,781
458
200
675
49
2033
1,900
1,125
308
0
0
0
Totals
63,000
37,738
10,987
7,000
12,085
2,429
a. Receipt rates based on assumptions presented in the Analysis of the Total System Life-Cycle Cost of the Civilian Radioactive
Waste Management Program (DOE 1998a, all) and the results of the CALVIN analysis.
b. Projected spent nuclear fuel acceptance rates (until agreements are reached with purchasers/producers/custodians).
c. DOE spent nuclear fuel at the Idaho National Engineering and Enviroiunental Laboratory to be removed by 2035. Three
hundred rail shipments of Navy fuel will be among the early shipments to a DOE receiving facility.
d. MTHM = metric tons of heavy metal.
e. LWT = legal-weight truck.
J.1 .2.1 .2 DOE Spent Nuclear Fuel and High-Level Radioactive Waste
To estimate the number of DOE spent nuclear fuel and high-level radioactive waste shipments, the
analysis used the number of handling units or number of canisters and the number of canisters per
shipment reported by the DOE sites in 1998 (see Appendix A, page A-34; Jensen 1998, all). To
determine the number of shipments of DOE spent nuclear fuel and high-level radioactive waste, the
analysis assumed one canister would be shipped in a legal-weight truck cask. For rail shipments, the
analysis assumed that five 61 -centimeter (24-inch)-diameter high-level radioactive waste canisters would
be shipped in a rail cask. For rail shipments of DOE spent nuclear fuel, the analysis assumed that rail
casks would contain nine approximately 46-centimeter (18-inch) canisters or four approximately
61 -centimeter canisters. The number of DOE spent nuclear fuel canisters of each size is presented in
Appendix A.
J-15
Transportation
Table J-5. Shipments of commercial spent nuclear fuel, mostly legal-weight truck scenario" (page 1
of 2).
Proposed Action
Modules 1 and 2
Site
Reactor
State
Fuel type
(2010-2033)
(2010-2048)
Browns Ferry
Browns Ferry 1
AL
B"
856
1,465
Browns Ferry 3
AL
B
319
602
Joseph M. Farley
Joseph M. Farley 1
AL
F
336
544
Joseph M. Farley 2
AL
P
297
582
Arkansas Nuclear
One
Arkansas Nuclear One, Unit 1
AR
P
302
438
Arkansas Nuclear One, Unit 2
AR
P
332
525
Palo Verde
Palo Verde 1
AZ
P
345
797
Palo Verde 2
AZ
P
364
840
Palo Verde 3
AZ
P
309
861
Diablo Canyon
Diablo Canyon 1
CA
P
327
617
Diablo Canyon 2
CA
P
305
691
Humboldt Bay
Humboldt Bay
CA
B
44
44
Rancho Seco
Rancho Seco 1
CA
P
124
124
San Onofre
San Onofre 1
CA
P
52
52
San Onofre 2
CA
P
402
600
San Onofre 3
CA
P
413
632
Haddam Neck
Haddam Neck
CT
P
255
255
Millstone
Millstone 1
CT
B
463
543
Millstone 2
CT
P
358
551
Millstone 3
CT
P
245
575
Crystal River
Crystal River 3
FL
P
283
442
St. Lucie
St. Lucie 1
FL
P
389
571
St. Lucie 2
FL
P
292
515
Turkey Point
Turkey Point 3
FL
P
295
413
Turkey Point 4
FL
P
287
458
Edwin I. Hatch
Edwin L Hatch 1
GA
B
871
1,334
Vogtle
Vogtle 1
GA
P
593
1,462
Duane Arnold
Duane Arnold
lA
B
279
420
Braidwood
Braidwood 1
IL
P
615
1,494
Byron
Byron 1
IL
P
617
1,444
Clinton
Clinton 1
IL
B
296
690
Dresden/Morris
Dresden 1
IL
B
76
76
Dresden 2
DL
B
430
521
Dresden 3
IL
B
473
565
Morns'*
IL
B
319
319
Moms'"
IL
P
88
88
LaSalle
LaSalle 1
IL
B
596
1,261
Quad Cities
Quad Cities 1
IL
B
798
1,123
Zion
Zionl
IL
P
771
1,028
Wolf Creek
Wolf Creek 1
KS
P
349
708
River Bend
River Bend 1
LA
B
324
823
Waterford
Waterford 3
LA
P
313
675
Pilgrim
Pilgrim 1
MA
B
316
476
Yankee-Rowe
Yankee-Rowe 1
MA
P
134
134
Calvert Cliffs
Calvert Cliffs 1
MD
P
757
1,140
Maine Yankee
Maine Yankee
ME
P
356
356
Big Rock Point
Big Rock Point
MI
B
131
131
D. C. Cook
D. C. Cook 1
MI
P
824
1,235
Fermi
Fermi 2
MI
B
312
764
Palisades
Palisades
MI
P
367
454
Monticello
Monticello
MN
B
267
342
Prairie Island
Prairie Island 1
MN
P
572
805
Callaway
Callaway 1
MO
P
392
735
Grand Gulf
Grand Gulf 1
MS
B
516
1,016
Brunswick
Brunswick 1
NC
P
40
40
Brunswick 2
NC
P
36
36
J-16
Transportation
Table J-5. Shipments of commercial spent nuclear fuel, mostly legal-weight truck scenario^ (page 2
of 2).
Proposed Action
Modules 1 and 2
Site
Reactor
State
Fuel type
(2010-2033)
(2010-2048)
Brunswick (continued]
1
Brunswick 1
NC
B"
232
426
Brunswick 2
NC
B
232
401
Shearon Harris
Shearon Harris 1
NC
P'
298
769
Shearon Harris
NC
B
152
152
McGuire
McGuire 1
NC
P
387
690
McGuire 2
NC
P
436
774
Cooper Station
Cooper Station
NE
B
274
454
Fort Calhoun
Fort Calhoun
NE
P
258
362
Seabrook
Seabrook 1
NH
P
235
630
Oyster Creek
Oyster Creek 1
NJ
B
424
519
Salem/Hope Creek
Salem 1
NJ
P
330
545
Salem 2
NJ
P
298
571
Hope Creek
NJ
B
399
876
James A. FitzPatrick/
James A. FitzPatrick
NY
B
364
554
Nine Mile Point
Nine Mile Point 1
NY
B
401
499
Nine Mile Point 2
NY
B
329
918
Ginna
Ginna
NY
P
309
379
Indian Point
Indian Point 1
NY
P
40
40
Indian Point 2
NY
P
364
590
Indian Point 3
NY
P
297
525
Davis-Besse
Davis-Besse 1
OH
P
286
535
Perry
Perry 1
OH
B
288
631
Trojan
Trojan
OR
P
195
195
Beaver Valley
Beaver Valley 1
PA
P
330
534
Beaver Valley 2
PA
P
221
622
Limerick
Limerick 1
PA
B
693
1,722
Peach Bottom
Peach Bottom 2
PA
B
480
696
Peach Bottom 3
PA
B
444
712
Susquehanna
Susquehanna 1
PA
B
808
1,582
Three Mile Island
Three Mile Island 1
PA
P
287
435
Catawba
Catawba 1
SC
P
325
663
Catawba 2
SC
P
318
667
Oconee
Oconee I
SC
P
727
1,043
Oconee 3
SC
P
280
457
H. B. Robinson
H. B. Robinson 2
SC
P
231
306
Summer
Summer 1
SC
P
291
538
Sequoyah
Sequoyah
TN
P
560
1,179
Watts Bar
Watts Bar 1
TN
P
146
840
Comanche Peak
Comanche Peak 1
TX
P
559
1,558
South Texas
South Texas 1
TX
P
256
738
South Texas 2
TX
P
229
710
North Anna
North Anna 1
VA
P
634
1,079
Surry
Surry 1
VA
P
647
902
Vermont Yankee
Vermont Yankee 1
VT
B
369
484
WPPSS' 2
WPPSS 2
WA
B
353
736
Kewaunee
Kewaunee
WI
P
288
401
LaCrosse
LaCrosse
WI
B
37
37
Point Beach
Point Beach
WI
P
575
742
Total BWR''
13,965
234>14
Total PWR'
23,773
42,936
a. Source: TRW (1999a, Section 2).
b. B = boiling-water reactor (BWR).
c. P = pressurized-water reactor (PWR).
d. Morris is a storage facility located close to the three Dresden reactors.
e. WPPSS = Washington Public Power Supply System.
J-17
Transportation
Table J-6. Shipments of commercial spent nuclear fuel, mostly rail scenario" (page 1 of 2).
Site
Reactor
State Fuel type
Cask
Profwsed
Action
2010 - 2033
Modules
1 and 2
2010 - 2048
Browns Ferry
Browns Ferry 1
AL
B"
Medium
239
422
Browns Ferry 3
AL
B
Medium
88
168
Joseph M. Farley
Joseph M. Farley 1
AL
F
Large
54
78
Joseph M. Farley 2
AL
P
Large
49
79
Arkansas Nuclear One
Arkansas Nuclear One, Unit 1
AR
P
Medium
81
115
Arkansas Nuclear One, Unit 2
AR
P
Medium
89
137
Palo Verde
Palo Verde 1
AZ
P
Large
53
120
Palo Verde 2
AZ
P
Large
56
124
Palo Verde 3
AZ
P
Large
47
106
Diablo Canyon
Diablo Canyon 1
CA
P
Medium
103
169
Diablo Canyon 2
CA
P
Medium
97
174
Humboldt Bay
Humboldt Bay
CA
B
Truck
44
44
Rancho Seco
Rancho Seco 1
CA
P
Large
21
21
San Onofre
San Onofre 1
CA
P
Large
9
8
San Onofre 2
CA
P
Large
66
97
San Onofre 3
CA
P
Large
68
102
Haddam Neck
Haddam Neck
CT
P
Truck
255
255
Millstone
Millstone 1
CT
B
Small
174
204
Millstone 2
CT
P
Small
120
183
Millstone 3
CT
P
Medium
73
137
Crystal River
Crystal River 3
FL
P
Truck
283
442
St. Lucie
St. Lucie 1
FL
P
Truck
389
571
St. Lucie 2
FL
P
Medium
88
140
Turkey Point
Turkey Point 3
FL
P
Medium
73
111
Turkey Point 4
FL
P
Medium
72
117
Edwin I. Hatch
Edwin L Hatch 1
GA
B
Large
128
197
Vogtle
Vogtle 1
GA
P
Small
195
431
Duane Arnold
Duane Arnold
lA
B
Small
105
158
Braidwood
Braidwood 1
IL
P
Large
95
215
Byron
Byron 1
IL
P
Large
136
244
Clinton
Clinton 1
IL
B
Medium
103
200
Dresden/Morris
Dresden 1
IL
B
Small
29
29
Dresden 2
IL
B
Small
162
193
Dresden 3
IL
B
Small
177
208
Morris"
IL
B
Large
47
47
Morris''
IL
P
Large
14
14
LaSalle
laSallel
IL
B
Large
89
172
Quad Cities
Quad Cities 1
IL
B
Small
299
419
Zion
Zion 1
IL
P
Medium
147
250
Wolf Creek
Wolf Creek 1
KS
P
Large
52
106
River Bend
River Bend 1
LA
B
Large
48
101
Waterford
Waterford 3
LA
P
Large
49
91
Pilgrim
Pilgrim 1
MA
B
Truck
316
476
Yankee-Rowe
Yankee-Rowe 1
MA
P
Large
15
15
Calvert Cliffs
Calvert Cliffs 1
MD
P
Medium
198
303
Maine Yankee
Maine Yankee
ME
P
Large
60
60
Big Rock Point
Big Rock Point
MI
B
Large
8
8
D. C. Cook
D. C. Cook 1
MI
P
Medium
214
346
Fermi
Fermi 2
MI
B
Medium
100
199
Palisades
Palisades
MI
P
Medium
78
117
Monticello
Monticello
MN
B
Truck
267
342
Prairie Island
Prairie Island 1
MN
P
Medium
151
221
Callaway
Callaway 1
MO
P
Large
62
114
Grand Gulf
Grand Gulf 1
MS
B
Large
76
143
J-18
Transportation
Table J-6. Shipments
of commercial spent nuclear fuel, mostly rail
scenario* (page 2 of 2).
Proposed
Modules
Action
1 and 2
Site
Reactor
State
Fuel type
Cask
2010 - 2033
2010-2048
Brunswick
Brunswick 1
NC
P'
Small
14
14
Brunswick 2
NC
P
Small
12
12
Brunswick 1
NC
B'
Small
88
150
Brunswick 2
NC
B
Small
87
145
Shearon Harris
Shearon Harris 1
NC
P
Small
93
201
Shearon Harris
NC
B
Small
57
57
McGuire
McGuire 1
NC
P
Medium
115
199
McGuire 2
NC
P
Medium
138
228
Cooper Station
Cooper Station
NE
B
Small
103
166
Fort Calhoun
Fort Calhoun
NE
P
Small
87
121
Seabrook
Seabrook 1
NH
P
Large
37
83
Oyster Creek
Oyster Creek 1
NJ
B
Medium
108
151
Salem/Hope Creek
Salem 1
NJ
P
Medium
97
153
Salem 2
NJ
P
Medium
83
143
Hope Creek
NJ
B
Large
59
125
James A. FitzPatrick/
FitzPatrick
NY
B
Large
54
79
Nine Mile Point
Nine Mile Point 1
NY
B
Medium
135
167
Nine Mile Point 2
NY
B
Medium
101
206
Ginna
Ginna
NY
P
Truck
309
379
Indian Point
Indian Point 1
NY
P
Truck
40
40
Indian Point 2
NY
P
Truck
364
590
Indian Point 3
NY
P
Truck
297
525
Davis-Besse
Davis-Besse 1
OH
P
Large
44
71
Perry
Perry 1
OH
B
Large
42
82
Trojan
Trojan
OR
P
Large
33
33
Beaver Valley
Beaver Valley 1
PA
P
Large
52
81
Beaver Valley 2
PA
P
Large
34
79
Limerick
Limerick 1
PA
B
Medium
262
497
Peach Bottom
Peach Bottom 2
PA
B
Medium
138
206
Peach Bottom 3
PA
B
Medium
127
197
Susquehaima
Susquehanna 1
PA
B
Large
119
219
Three Mile Island
Three Mile Island 1
PA
P
Medium
71
113
Catawba
Catawba 1
SC
P
Large
72
123
Catawba 2
SC
P
Large
76
130
Oconee
Oconee 1
SC
P
Medium
187
266
Oconee 3
SC
P
Medium
67
107
H. B. Robinson
H. B. Robinson 2
SC
P
Small
75
97
Summer
Sunmier 1
SC
P
Large
46
82
Sequoyah
Sequoyah
TN
P
Large
90
161
Watts Bar
Watts Bar 1
TN
P
Large
21
121
Comanche Peak
Comanche Peak 1
TX
P
Large
90
246
South Texas
South Texas 1
TX
P
Large
79
180
South Texas 2
TX
P
Large
72
178
North Anna
North Anna 1
VA
P
Large
101
167
Surry
Surry 1
VA
P
Large
105
144
Vermont Yankee
Vermont Yankee 1
VT
B
Small
139
182
WPPSS' 2
WPPSS 2
WA
B
Large
53
107
Kewaunee
Kewaunee
WI
P
Medium
73
106
La Crosse
La Crosse
WI
B
Truck
37
37
Point Beach
Point Beach
WI
P
Large
93
118
Total BWR*
4,208
6,503
Total PWR'
6,779
11,104
b. Source: TRW (1999a, Section 2).
lb. B = boiling-water reactor (BWR).
[ c. P = pressurized-water reactor (PWR).
|d. Morris is a storage facility located close to the three Dresden reactors.
k. WPPSS = Washington Public Power Supply System.
J-19
Transportation
Under the mostly legal-weight truck scenario for the Proposed Action, a total of about 1 1,800 truck
shipments of DOE spent nuclear fuel and high-level radioactive waste would be shipped to the repository.
In addition, due to the size and weight of the shipping casks for canisters that would contain naval spent
fuel, DOE would transport 300 shipments of naval spent fuel by rail from the Idaho National Engineering
and Environmental Laboratory to the repository. For Modules 1 and 2, under the mostly legal-weight
truck scenario, the analysis estimated 3,740 DOE spent nuclear fuel and 22,300 high-level radioactive
waste truck shipments and 300 naval spent nuclear fuel shipments by rail.
Under the mostly rail scenario for the Proposed Action, the analysis estimated that 770 railcar shipments
of DOE spent nuclear fuel, including 300 railcar shipments of naval spent nuclear fuel (one naval spent
nuclear fuel canister per rail cask), and 1,660 railcar shipments of high-level waste would travel to the
repository. For Modules 1 and 2, under this scenario 800 railcar shipments of DOE spent nuclear fuel,
including 300 railcar shipments of naval spent nuclear fuel, and 4,460 railcar shipments of high-level
radioactive waste would be shipped. Table J-7 lists the estimated number of shipments of DOE spent
nuclear fuel from each of the four sites for both the Proposed Action and Modules 1 and 2. Table J-8 lists
the number of shipments of high-level radioactive waste for the Proposed Action and for Modules 1
and 2.
Table J-7. DOE spent nuclear fuel shipments by site.
Proposed
Action
Module 1
or 2
Site
Mostly truck
Mostly rail
Mostly truck
Mostly rail
INEEL^^"
1,388
434
1,467
443
Savannah River Site
1,316
149
1,411
159
Hanford
754
147
809
157
Fort St. Vrain
312
36
334
38
Totals
3,770
766
4,021
797
a. nVEEL = Idaho National Engineering and Environmental Laboratory.
b. Includes 300 railcar shipments of naval spent nuclear fuel.
Table J-8. Number of canisters of high-level radioactive waste and shipments from DOE sites.
Proposed Action Module 1 or 2
Site
Canisters
Mostly truck
Mostly rail
Mostly truck
Mostly rail
INEEL"
1,300
0
0
1,300
260
Hanford
14,500
1,960
400
14,500
2,900
Savannah River Site
6,200
6,055
1,200
6,200
1,240
West Valley"
300
300
60
300
60
Totals
22,300
8,315
1,660
22,300
4,460
a. INEEL = Idaho National Engineering and Environmental Laboratory.
b. High-level radioactive waste at West Valley is commercial rather than DOE waste.
J.I .2.1 .3 Greater-Than-Class-C and Special-Performance-Assessment-Required Waste
Stiipments
Reasonably foreseeable future actions could include shipment of Greater-Than-Class-C and Special-
Performance-Assessment-Required waste to the Yucca Mountain Repository (Appendix A describes
Greater-Than-Class-C and Special-Performance-Assessment-Required wastes). Commercial nuclear
powerplants, research reactors, radioisotope manufacturers, and other manufacturing and research
institutions generate low-level radioactive waste that exceeds the Nuclear Regulatory Commission Class
J-20
Transportation
C shallow-land-burial disposal limits. In addition to DOE-held material, there are three other sources or
categories of Greater-Than-Class-C low-level radioactive waste:
• Nuclear utilities
• Sealed sources
• Other generators
The activities of nuclear electric utilities and other radioactive waste generators to date have produced
relatively small quantities of Greater-Than-Class-C low-level radioactive waste. As the utilities take their
reactors out of service and decommission them, they could generate more waste of this type.
DOE Special-Performance-Assessment-Required low-level radioactive waste could include the following
materials:
• Production reactor operating wastes
• Production and research reactor decommissioning wastes
• Non-fuel-bearing components of naval reactors
• Sealed radioisotope sources that exceed Class C limits for waste classification
• DOE isotope production-related wastes
• Research reactor fuel assembly hardware
The analysis estimated the number of shipments of Greater-Than-Class-C and Special-Performance-
Assessment-Required waste by assuming that 10 cubic meters (about 350 cubic feet) would be shipped in
a rail cask and 2 cubic meters (about 71 cubic feet) would be shipped in a truck cask. Table J-9 lists the
resulting number of commercial Greater-Than-Class-C shipments in Inventory Module 2 for both truck
and rail shipments. The shipments of Greater-Than-Class-C waste from commercial utilities would
originate among the commercial reactor sites. Typically, boiling-water reactors would ship a total of
about 9 cubic meters (about 318 cubic feet) of Greater-Than-Class-C waste per site, while pressurized-
water reactors would ship about 20 cubic meters (about 710 cubic feet) per site (see Appendix A). The
impacts of transporting this waste were examined for each reactor site. The analysis assumed that sealed
sources and Greater-Than-Class-C waste identified as "other" would be shipped firom the DOE Savannah
River Site (see Table J- 10).
Table J-9. Commercial Greater-Than-Class-C waste shipments
Category
Volume (cubic meters)^*"
Truck
Rail
Commercial utilities
1,350
740
210
Sealed sources
240
120
25
Other
470
230
50
Total
2,060
1,090
285
a. Source: Appendix A.
b. To convert cubic meters to cubic feet, multiply by 35.314.
The analysis assumed DOE Special-Performance-Assessment-Required waste would be shipped from 4
DOE sites listed in Table J- 10. Naval reactor and Argonne East Special-Performance- Assessment-
Required waste is assumed to be shipped from the Idaho National Engineering and Environmental
Laboratory.
J.1 .2.1 .4 Sensitivity of Transportation Impacts to Number of Shipments
As discussed in Section J. 1.2.1, the number of shipments from commercial and DOE sites to the
repository would depend on the mix of legal-weight truck and rail shipments. Because DOE has decided
J-21
Transportation
Site"
Volume (cubic meters)''''
Rail
Hanford
20
2
INEEL
520
57"
SRS (ORNL)
2,900
290
West Valley
550
56
Total
3,990
405
Table J-10. DOE Special-Performance-Assessment-Required waste shipments.
Truck
10
260
1,470
280
2,020
a. Abbreviations: INEEL = Idaho National Engineering and Environmental Laboratory; SRS = Savannah River Site; ORNL =
Oak Ridge National Laboratory.
b. Source: Appendix A.
c. To convert cubic meters to cubic feet, multiply by 35.314.
d. Includes 55 shipments from naval reactors.
not to determine this mix at this time (10 years before the projected start of shipping operations), the
analysis used two scenarios to provide results that bound the range of anticipated impacts. Thus, for a
mix of legal-weight truck and rail shipments within the range of the mostly legal-weight truck and mostly
rail scenarios, the impacts would be likely to lie within the bounds of the impacts predicted by the
analysis. For example, a mix that is different from the scenarios analyzed could consist of 5,000 legal-
weight truck shipments and 9,000 rail shipments over 24 years (compared to 2,600 and 10,800,
respectively, for the mostly rail scenario), hi this example, the number of traffic fatalities would be
between 3.6 (estimated for the Proposed Action under the mostly rail scenario) and 3.9 (estimated for the
mostly legal-weight truck scenario). Other examples that have different mixes within the ranges bounded
by the scenarios would lead to results that would be within the range of the evaluated impacts.
In addition to mixes within the brackets, the number of shipments could fall outside the ranges used for
the mostly legal-weight truck and rail transportation scenarios. If, for example, the mostly rail scenario
used smaller rail casks than the analysis assumed, the number of shipments would be greater. If spent
nuclear fuel was placed in the canisters before they were shipped, the added weight and size of the
canisters would reduce the number of fuel assemblies that a given cask could accommodate; this would
increase the number of shipments. However, for the mostly rail scenario, even if the capacity of the casks
was half that used in the analysis, the impacts would remain below those forecast for the mostly legal-
weight truck scenario. Although impacts would be related to the number of shipments, because the
number of rail shipments would be very small in comparison to the total railcar traffic on the Nation's
railroads, increases or decreases would be small for impacts to biological resources, air quality,
hydrology, noise, and other environmental resource areas. Thus, the impacts of using smaller rail casks
would be covered by the values estimated in this EIS.
For legal-weight truck shipments, the use of casks carrying smaller payloads than those used in the
analysis (assuming the shipment of the same spent nuclear fuel) would lead to larger impacts for incident-
free transportation and traffic fatalities and about the same level of radiological accident risk. The
relationship is approximately linear; if the payloads of truck shipping casks in the mostly legal-weight
truck scenario were less by one-half, the incident-free impacts would increase by approximately a factor
of 2. Conversely, because the amount of radioactive material in a cask would be less (assuming shipment
of the same spent nuclear fuel), the radiological consequences of maximum reasonably foreseeable
accident scenarios would be less with the use of smaller casks. If smaller casks were used to
accommodate shipments of spent nuclear fuel with shorter cooling time and higher bumup, the
radiological consequences of maximum reasonably foreseeable accident scenarios would be about the
same.
J-22
Transportation
J.1.2.2 Transportation Routes
At this time, about 10 years before shipments could begin, DOE has not determined the specific routes it
would use to ship spent nuclear fuel and high-level radioactive waste to the proposed repository.
Nonetheless, this analysis used current regulations governing highway shipments and historic rail industry
practices to select existing highway and rail routes to estimate potential environmental impacts of national
transportation. Routing for shipments of spent nuclear fuel and high-level radioactive waste to the
proposed repository would comply with applicable regulations of the Department of Transportation and
the Nuclear Regulatory Commission in effect at the time the shipments occurred, as stated in the proposed
DOE revised policy and procedures for implementing Section 180(c) of the Nuclear Waste Policy Act
(DOE 1998b, all).
Approximately 4 years before shipments to the proposed repository began, the Office of Civilian
Radioactive Waste Management plans to identify the preliminary routes that DOE anticipates using in
state and tribal jurisdictions so it can notify governors and tribal leaders of their eligibility for assistance
under the provisions of Section 180(c) of the Nuclear Waste Policy Act. DOE has published a revised
proposed policy statement that sets forth its revised plan for implementing a program of technical and
financial assistance to states and Native American tribes for training public safety officials of appropriate
units of local government and tribes through whose jurisdictions the Department plans to transport spent
nuclear fuel or high-level radioactive waste (63 FR 83, January 2, 1998).
The analysis of impacts of the Proposed Action and Modules 1 and 2 used characteristics of routes that
shipments of spent nuclear fuel and high-level radioactive waste could travel from the originating sites
listed in Tables J-5 through J-8. Existing routes that could be used were identified for the mostly legal-
weight truck and mostly rail transportation scenarios and included the 10 rail and heavy-haul truck
implementing alternatives evaluated in the EIS for transportation in Nevada. The route characteristics
used were the transportation mode (highway, railroad, or navigable waterway) and, for each of the modes,
the total distance between an originating site and the repository. In addition, the analysis estimated the
fraction of travel that would occur in rural, suburban, and urban areas for each route. The fraction of
travel in each population zone was determined using 1990 census data (see Section J.1.1.2 and J. 1.1.3) to
identify population-zone impacts for route segments. The highway routes were selected for the analysis
using the HIGHWAY computer program and routing requirements of the Department of Transportation
for shipments of Highway Route-Controlled Quantities of Radioactive Materials (49 CFR 397.101).
Shipments of spent nuclear fuel and high-level radioactive waste would contain Highway Route-
Controlled Quantities of Radioactive Materials.
J.1 .2.2.1 Routes Used in the Analysis
Routes used in the analysis of transportation impacts of the Proposed Action and Inventory Modules 1
and 2 are highways and rail lines that DOE anticipates it could use for legal-weight truck or rail shipments
from each origin to Nevada. For rail shipments that would originate at sites not served by railroads,
routes used for analysis include highway routes for heavy-haul trucks or barge routes from the sites to
railheads. Figures J-5 and J-6 show the Interstate System highways and mainline railroads, respectively,
and their relationship to the commercial and DOE sites and Yucca Mountain. Tables J-1 1 and J- 12 list
the lengths of trips and the distances of the highway and rail routes, respectively, in rural, suburban, and
urban population zones. Sites that would be capable of loading rail casks, but that do not have direct rail
access, are listed in Table J-12. The analysis used four ending rail nodes in Nevada (Beowawe, Caliente,
Jean, and Apex) to select rail routes from the 77 sites. These rail nodes would be starting points for the
rail and heavy-haul truck implementing alternatives analyzed for transportation in Nevada.
J-23
Transportation
J-24
Transportation
J-25
Transportation
Table J-11. Highway distances for legal-weight truck shipments from commercial and DOE sites to
Yucca Mountain, mostly legal-weight truck transportation (kilometers)"'^ (page 1 of 2).
Origin
State
Total"
Rural
Suburban
Urban
Browns Ferry
AL
3,442
3,022
374
45
Joseph M. Farley
AL
4,229
3,647
520
62
Arkansas Nuclear One
AR
2,810
2,588
192
30
Palo Verde
AZ
1,007
886
100
21
Diablo Canyon
CA
1,016
828
119
68
Humboldt Bay
CA
1,749
1,465
192
92
Rancho Seco
CA
1,228
1,028
124
76
San Onofre
CA
694
517
89
88
Haddam Neck
CT
4,519
3,708
736
75
Millstone
CT
4,527
3,673
746
109
Crystal River
FL
4,319
3,606
653
59
St. Lucie
FL
4,588
3,793
729
64
Turkey Point
FL
4,842
3,888
821
132
Edwin I. Hatch
GA
3,986
3,373
553
58
Vogtle
GA
3,938
3,301
573
63
Duane Arnold
lA
2,773
2,544
189
40
Braidwood
IL
3,063
2,796
231
36
Byron
IL
3,032
2,773
223
36
Clinton
IL
3,104
2,814
252
38
Dresden/Morris
IL
3,059
2,798
225
36
La Salle
IL
3,017
2,766
215
36
Quad Cities
IL
2,877
2,631
211
36
Zion
IL
3,167
2,834
284
50
Wolf Creek
KS
2,374
2,226
131
16
River Bend
LA
3,446
2,941
420
85
Waterford
LA
3,531
3,003
444
84
Pilgrim
MA
4,722
3,697
930
94
Yankee-Rowe
MA
4,616
3,692
831
92
Calvert Cliffs
MD
4,278
3,511
684
82
Maine Yankee
ME
4,894
3,733
1,052
108
Big Rock Point
MI
3,866
3,266
547
52
D. C. Cook
MI
3,196
2,827
319
51
Fermi
MI
3,524
3,014
449
61
Palisades
MI
3,244
2,855
338
51
Monticello
MN
3,003
2,702
261
41
Prairie Island
MN
2,993
2,720
233
41
Callaway
MO
2,633
2,399
206
27
Grand Gulf
MS
3,354
2,989
311
54
Brunswick
NC
4,418
3,672
680
66
Shearon Harris
NC
4,187
3,493
630
63
McGuire
NC
3,991
3,415
516
58
Cooper Station
NE
2,523
2,328
160
36
Fort Calhoun
NE
2,348
2,165
148
35
Seabrook
NH
4,725
3,676
942
107
Oyster Creek
NJ
4,424
3,530
825
69
Salem/Hope Creek
NJ
4,350
3,531
739
79
Ginna
NY
4,089
3,357
642
91
Indian Point
NY
4,382
3,695
620
67
James FitzPatrick/Nine
NY
4,234
3,461
688
85
Mile Point
J-26
Transportation
Table J-11. Highway distances for legal-weight truck shipments from commercial and DOE sites to
Yucca Mountain,
mostly legal-weight truck
transportation (kilometers)^'
"(page 2 of 2).
Origin
State
Tota^
Rural
Suburban
Urban
Davis-Besse
OH
3,520
3,106
358
56
Perry
OH
3,693
3,157
464
73
Trojan
OR
2,137
1,865
237
36
Beaver Valley
PA
3,779
3,215
500
64
Limerick
PA
4,287
3,484
741
62
Peach Bottom
PA
4,205
3,479
662
64
Susquehanna
PA
4,126
3,539
528
59
Three Mile Island
PA
4,147
3,443
643
60
Catawba
SC
3,994
3,364
575
54
Oconee
SC
3,853
3,264
532
55
H. B. Robinson
SC
4,112
3,417
628
65
Summer
SC
3,996
3,383
557
55
Sequoyah
TN
3,500
3,039
414
45
Watts Bar
TN
3,578
3,138
394
45
Comanche Peak
TX
2,794
2,547
213
34
South Texas
TX
3,011
2,652
295
64
North Anna
VA
4,081
3,503
515
63
Surry
VA
4,255
3,577
610
67
Vermont Yankee
VT
4,616
3,675
847
94
WPPSS" 2
WA
1,880
1,669
178
32
Kewaunee
WI
3,347
2,979
314
55
La Crosse
WI
3,014
2,773
198
43
Point Beach
WI
3,341
2,972
314
55
Ft. St. Vrain'
CO
1,415
1,311
93
10
ineel'
ID
1,201
1,044
130
27
West Valley^
NY
3,959
3,322
562
75
Savannah River'
SC
3,961
3,321
574
64
Hanford«
WA
1,881
1,671
178
32
a. To convert kilometers to miles, multiply by 0.62137.
b. Distances determined for purposes of analysis using HIGHWAY computer program.
c. Totals might differ firom sums due to method of calculation and rounding.
d. IX)E spent nuclear fuel site.
e. DOE spent nuclear fuel and high-level waste site.
f. DOE high-level waste site.
g. WPPSS = Washington Public Power Supply System.
STATE-DESIGNATED PREFERRED ROUTES
Department of Transportation regulations specify that states and tribes can designate preferred
routes that are alternatives, or in addition to, Interstate System highways including bypasses or
beltways for the transportation of Highway Route-Controlled Quantities of Radioactive Materials.
Highway Route-Controlled Quantities of Radioactive Materials include spent nuclear fuel and high-
level radioactive waste in quantities that would be shipped on a truck or railcar to the repository. If a
state or tribe designated such a route, shipments of spent nuclear fuel and high-level radioactive
waste would use the preferred route if (1) it was an alternative preferred route, (2) it would result in
reduced time in transit, or (3) it would replace pickup or delivery routes. Ten states — Alabama,
Arkansas, California, Colorado, Iowa, Kentucky, Nebraska, New Mexico, Tennessee, and Virginia —
have designated alternative or additional preferred routes (Rodgers 1998, all). Although Nevada has
designated a State routing agency to the Department of Transportation (Nevada Revised Statutes,
Chapter 408.141), the State has not designated alternative preferred routes for Highway Route-
Controlled Quantities of Radioactive Materials.
J-27
Transportation
Table J-12. Rail transportation distances from commercial and DOE sites to Nevada ending rail nodes"
(kilometers)''''' (page 1 of 5)
Site
State
Destination
Total"
Rural Suburban Urban
Commercial sites with direct rail access
Joseph M. Farley
Arkansas Nuclear One
Palo Verde
Rancho Seco
San Onofre
Millstone
Edwin I. Hatch
Vogtle
Duane Arnold
Braidwood
Byron
Clinton
Dresden/Morris
La Salle
Quad Cities
AL
AR
AZ
CA
CA
CT
GA
GA
lA
IL
IL
IL
IL
IL
IL
Apex
4,495
3,872
562
60
Caliente
4,322
3,698
562
60
Beowawe
4,177
3,593
535
48
Jean
4,577
3,937
574
65
Apex
3,170
2,960
181
29
Caliente
2,996
2,786
181
29
Beowawe
2,852
2,681
154
17
Jean
3,251
3,024
193
34
Apex
976
864
89
23
Caliente
1,149
1,038
89
23
Beowawe
1,908
1,524
274
109
Jean
894
800
77
18
Apex
985
781
151
53
Caliente
1,159
955
151
53
Beowawe
706
589
83
32
Jean
904
717 .
139
48
Apex
576
409
105
63
Caliente
750
582
105
63
Beowawe
1,576
1,167
286
121
Jean
495
344
93
58
Apex
4,728
3,526
994
208
Caliente
4,555
3,353
994
208
Beowawe
4,411
3,247
966
197
Jean
4,810
3,591
1,005
213
Apex
4,403
3,830
514
58
Caliente
4,229
3,656
514
58
Beowawe
4,085
3,551
486
47
Jean
4,484
3,894
525
64
Apex
4,459
3,877
523
58
Caliente
4,286
3,703
523
58
Beowawe
4,141
3,598
495
47
Jean
4,541
3,942
534
64
Apex
2,745
2,547
167
31
Caliente
2,572
2,374
167
31
Beowawe
2,428
2,268
140
20
Jean
2,827
2,612
178
36
Apex
3,166
2,798
284
85
Caliente
2,993
2,624
285
85
Beowawe
2,849
2,518
257
73
Jean
3,248
2,862
296
90
Apex
2,979
2,740
205
35
Caliente
2,806
2,566
205
35
Beowawe
2,662
2,461
177
24
Jean
3,061
2,805
216
41
Apex
3,172
2,891
228
53
Caliente
2,998
2,718
228
53
Beowawe
2,854
2,612
201
42
Jean
3,253
2,956
239
58
Apex
3,087
2,786
255
46
Caliente
2,914
2,613
255
46
Beowawe
2,769
2,507
227
35
Jean
3,169
2,851
266
51
Apex
3,060
2,831
196
33
Caliente
2,887
2,657
196
33
Beowawe
2,953
2,691
225
37
Jean
3,403
3,201
181
20
Apex
3,003
2,759
210
33
Caliente
2,829
2,586
210
33
Beowawe
2,895
2,619
238
38
Jean
3,345
3,130
195
21
J-28
Transportation
Table J-12. Rail transportation distances from commercial and DOE sites to Nevada ending rail nodes'
(kilometers)'''^ (page 2 of 5).
Site
State Destination Total Rural Suburban Urban
Commercial sites with direct rail access (continued}
Zion
Wolf Creek
River Bend
Waterford
Yankee-Rowe
Maine Yankee
Big Rock Point
D. C. Cook
Fermi
Prairie Island
Brunswick
Shearon Harris
McGuire
Seabrook
PitzPatrick/Nine Mile Point
IL
KS
LA
LA
MA
ME
MI
MI
MI
MN
NC
NC
NC
NH
NY
Apex
3,119
2,765
279
75
Caliente
2,946
2,591
279
75
Beowawe
2,801
2,486
252
64
Jean
3,201
2,829
291
81
Apex
2,685
2,528
131
27
Caliente
2,512
2,354
131
27
Beowawe
2,368
2,249
103
16
Jean
2,767
2,593
142
32
Apex
3,509
3,114
322
73
Caliente
3,380
2,944
377
59
Beowawe
3,445
2,975
406
65
Jean
3,428
3,049
311
68
Apex
3,551
3,173
304
74
Caliente
3,423
3,003
359
61
Beowawe
3,487
3,033
388
66
Jean
3,470
3,108
293
69
Apex
4,471
3,466
823
183
Caliente
4,298
3,292
823
183
Beowawe
4,153
3,187
7%
171
Jean
4,553
3,530
835
188
Apex
4,908
3,629
1,075
204
Caliente
4,734
3,455
1,075
204
Beowawe
4,590
3,350
1,048
193
Jean
4,989
3,693
1,087
209
Apex
3,835
3,299
431
105
Caliente
3,662
3,126
431
105
Beowawe
3,517
3,020
404
93
Jean
3,917
3,364
443
110
Apex
3,209
2,799
324
86
Caliente
3,035
2,625
324
86
Beowawe
2,891
2,520
297
75
Jean
3,290
2,863
336
91
Apex
3,649
3,046
469
135
Caliente
3,476
2,872
469
135
Beowawe
3,332
2,767
442
123
Jean
3,731
3,110
481
140
Apex
2,980
2,715
238
28
Caliente
2,807
2,541
238
28
Beowawe
2,663
2,436
210
16
Jean
3,062
2,780
249
33
Apex
4,768
3,972
724
71
Caliente
4,594
3,799
724
71
Beowawe
4,450
3,693
697
59
Jean
4,849
4,037
736
76
Apex
4,669
3,910
689
69
Caliente
4,495
3,737
689
69
Beowawe
4,351
3.631
662
58
Jean
4,751
3,975
701
75
Apex
4,539
3,779
683
77
Caliente
4,366
3,605
683
77
Beowawe
4,221
3,500
656
65
Jean
4,621
3,844
694
82
Apex
4,755
3,567
987
201
Caliente
4,582
3,393
987
201
Beowawe
4,437
3,288
960
190
Jean
4,837
3,632
999
206
Apex
4,213
3,2%
728
188
Caliente
4,039
3,123
728
188
Beowawe
3,895
3,017
701
177
Jean
4,294
3,361
740
193
J-29
Transportation
Table J-12. Rail transportation distances from commercial and DOE sites to Nevada ending rail nodes"
(kilometers)'''^ (page 3 of 5).
Site
State
Destination
Total"
Rural
Suburban Urban
Commercial sites with direct rail access (continued)
Davis Besse OH
Perry
Trojan
Beaver Valley
Limerick
Susquehanna
Three Mile Island
Catawba
H. B. Robinson
Summer
Sequoyah
Watts Bar
Comanche Peak
South Texas
North Anna
OH
OR
PA
PA
PA
PA
SC
SC
SC
TN
TN
TX
TX
VA
Apex
3,590
3,133
342
114
Caliente
3,416
2,960
342
114
Beowawe
3,272
2,854
315
103
Jean
3,671
3,198
354
120
Apex
3,692
3,131
416
145
Caliente
3,519
2,958
416
145
Beowawe
3,374
2,852
389
133
Jean
3,774
3,196
428
150
Apex
2,202
1,897
244
61
Caliente
2,031
1,871
136
23
Beowawe
1,539
1,445
85
9
Jean
2,121
1,833
233
56
Apex
3,819
3,212
499
108
Caliente
3,645
3,039
499
108
Beowawe
3,501
2,933
472
96
Jean
3,901
3,277
510
113
Apex
4,389
3,349
843
197
Caliente
4,216
3,175
843
197
Beowawe
4,072
3,070
816
186
Jean
4,471
3,414
855
203
Apex
4,406
3,412
819
175
Caliente
4,232
3,238
819
175
Beowawe
4,088
3,133
791
164
Jean
4,487
3,477
830
180
Apex
4,283
3,330
767
186
Caliente
4,110
3,157
767
186
Beowawe
3,966
3,051
739
175
Jean
4,365
3,395
778
191
Apex
4,537
3,756
702
77
Caliente
4,363
3,583
702
77
Beowawe
4,219
3,477
675
66
Jean
4,618
3,821
714
82
Apex
4,513
3,745
688
78
Caliente
4,339
3,572
688
78
Beowawe
4,195
3,466
661
67
Jean
4,594
3,810
700
83
Apex
4,472
3,782
621
68
Caliente
4,299
3,609
621
68
Beowawe
4,154
3,503
594
57
Jean
4,554
3,847
633
74
Apex
3,890
3,480
361
48
Caliente
3,716
3,307
361
48
Beowawe
3,572
3,201
333
37
Jean
3,971
3,545
372
53
Apex
3,887
3,544
286
57
Caliente
3,714
3,370
286
57
Beowawe
3,569
3,265
259
46
Jean
3,969
3,608
298
62
Apex
2,890
2,639
213
38
Caliente
2,716
2,465
213
38
Beowawe
2,791
2,512
236
43
Jean
2,445
2,338
101
5
Apex
3,055
2,800
206
49
Caliente
3,228
2,973
206
49
Beowawe
3,320
2,948
330
43
Jean
2,973
2,735
194
44
Apex
4,521
3,669
686
165
Caliente
4,347
3,496
686
165
Beowawe
4,203
3,390
659
153
Jean
4,602
3,734
698
170
J-30
Transportation
Table J-12. Rail transportation distances from commercial and DOE sites to Nevada ending rail nodes"
(kilometers)''''' (page 4 of 5).
Site
Stale Destination Total Rural Suburban Urban
Commercial sites with direct rail access (continued)
Vermont Yankee
WPPSS' 2
Commercial sites with indirect rail access
Browns Ferry
HH - 55.4 kilometers
Diablo Canyon
HH- 43.5 kilometers
St. Lucie
HH - 23.3 kilometers
Turkey Point
HH- 17.4 kilometers
Calvert Cliffs
HH- 41.9 kilometers
Palisades
HH- 41.9 kilometers
Callaway
HH- 18.5 kilometers
Grand Gulf
HH - 47.8 kilometers
Cooper Station
HH - 53.8 kilometers
Fort Calhoun
HH - 6.0 kilometers
Salem/Hope Creek
HH- 51.0 kilometers
Oyster Creek
HH - 28.5 kilometers
VT
WA
AL
CA
FL
FL
MD
NO
MO
MS
NE
NE
NJ
NJ
Apex
4,551
3,519
846
186
Caliente
4,378
3,345
846
186
Beowawe
4,233
3,240
818
175
Jean
4,633
3,584
857
192
Apex
1,946
1,807
116
22
Caliente
1,772
1,634
116
22
Beowawe
1,565
1,490
66
9
Jean
2,027
1,872
128
28
Apex
3,741
3,332
357
52
Caliente
3,567
3,158
357
52
Beowawe
3,423
3,053
329
41
Jean
3,822
3,397
368
57
Apex
893
609
174
110
Caliente
1,067
783
174
110
Beowawe
1,157
872
203
82
Jean
812
544
162
105
Apex
4,938
4,073
780
85
Caliente
4,765
3,899
780
85
Beowawe
4,621
3,794
753
73
Jean
4,863
4,006
732
125
Apex
5,285
4,305
841
138
Caliente
5,111
4,132
841
138
Beowawe
4,967
4,026
814
126
Jean
5,366
4,370
853
143
Apex
4,543
3,448
881
213
Caliente
4,369
3,275
881
213
Beowawe
4,225
3,169
854
201
Jean
4,625
3,513
893
218
Apex
3,257
2,816
353
88
Caliente
3,083
2,642
353
88
Beowawe
2,939
2,537
326
77
Jean
3,339
2,881
365
93
Apex
2,807
2,636
140
32
Caliente
2,634
2,462
140
32
Beowawe
2,490
2,357
113
20
Jean
2,889
2,701
151
37
Apex
3,686
3,355
291
39
Caliente
3,512
3,181
291
39
Beowawe
3,368
3,076
264
28
Jean
3,767
3,419
303
44
Apex
2,429
2,252
141
36
Caliente
2,256
2,078
141
36
Beowawe
2,111
1,973
114
25
Jean
2,511
2,317
153
42
Apex
2,313
2,189
102
21
Caliente
2,139
2,015
102
21
Beowawe
1,995
1,910
75
10
Jean
2,394
2,254
114
27
Apex
4,551
3,375
946
229
Caliente
4,378
3,202
946
229
Beowawe
4,234
3,097
919
218
Jean
4,633
3,440
958
235
Apex
4,568
3,395
952
221
Caliente
4,395
3,222
952
221
Beowawe
4,251
3,116
925
209
Jean
4,650
3,460
964
226
J-31
Transportation
Table J-12. Rail transportation distances from commercial and DOE sites to Nevada ending rail nodes"
(kilometers) '^ (page 5 of 5).
Site
State
Destination
Total"
Rural
Suburban Urban
Commercial sites with indirect rail access (continued)
Peach Bottom PA
HH - 58.9 kilometers
Oconee
HH - 17.5 kilometers
Surry
HH - 75.2 kilometers
Kewaunee
HH - 9.7 kilometers
Point Beach
HH - 36.4 kilometers
SC
VA
WI
WI
DOE spent nuclear fuel and high-level waste (direct rail access)
Ft. St. Vrain* 00
INEEL"
West Valley'
Savannah River Site
Hanford Site""
ID
NY
SC
WA
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
Apex
Caliente
Beowawe
Jean
4,304
4,131
3,986
4,386
4,257
4,084
3,940
4,339
4,505
4,332
4,188
4,587
3,444
3,270
3,126
3,526
3.397
3,224
3,080
3,479
1,561
1,387
1,298
1,643
1,059
88S
741
1,140
3,972
3,798
3,654
4,053
4,374
4,201
4,057
4,456
1,933
1,760
1,553
2,015
3,335
3,161
3,056
3,400
3,662
3,488
3,383
3,726
3,927
3,753
3,648
3,992
2,954
2,780
2,675
3,019
2,938
2,765
2,659
3,003
1,453
1,280
1,266
1,518
978
804
699
1,042
3,169
2,995
2,890
3,234
3,690
3,517
3,411
3,755
1,795
1,622
1,477
1,860
778
778
751
790
534
534
507
545
512
512
484
523
395
395
368
406
370
370
343
381
93
93
29
105
66
66
39
78
638
638
611
650
609
609
581
620
116
116
66
128
190
190
179
196
61
61
50
66
66
66
55
72
95
95
84
100
89
89
78
94
14
14
3
20
15
15
4
21
165
165
153
170
75
75
64
80
22
22
9
28
a. The ending rail nodes (INTERLINE computer program designations) are Apex- 14763; Caliente- 14770; Beowawe- 14791; and Jean- 16328.
b. To convert kilometers to miles, multiply by 0.62137.
c. This analysis used the INTERLINE computer program to estimate distances.
d. Totals might differ from sums due to method of calculation and rounding.
e. NP = nuclear plant.
f. DOE spent nuclear fuel.
g. DOE spent nuclear fuel and high-level radioactive waste,
h. DOE high-level radioactive waste.
i. WPPSS = Washington Public Power Supply System.
Selection of Highway Routes. The analysis of national transportation impacts used route
characteristics of existing highways, such as distances, population densities, and state-level accident
statistics. The analysis of highway shipments of spent nuclear fuel and high-level radioactive waste used
the HIGHWAY computer model (Johnson et al. 1993a, all) to determine highway routes using regulations
of the Department of Transportation (49 CFR 397.101) that specify how routes are selected. The
selection of "preferred routes" is required for shipment of these materials. DOE has determined that the
HIGHWAY program is appropriate for calculating highway routes and related information (Maheras and
J-32
Transportation
Pippen 1995, pages 2 to 5). HIGHWAY is a routing tool that DOE has used in previous EISs [for
example, the programmatic EIS on spent nuclear fuel (DOE 1995, page 1-6) and the Waste Isolation Pilot
Plant Supplement II EIS (DOE 1997a, pages 5 to 13)] to determine highway routes for impact analysis.
Because the regulations require that the preferred routes result in reduced time in transit, changing
conditions, weather, and other factors could result in the use of more than one route at different times for
shipments between the same origin and destination. However, for this analysis the program selected only
one route for travel from each site to the Yucca Mountain site.
Although shipments could use more than one preferred route in national highway transportation to
comply with Department of Transportation regulations (49 CFR 397.101), under current Department of
Transportation regulations all preferred routes would ultimately enter Nevada on Interstate 15 and travel
to the repository on U.S. Highway 95. States can designate alternative or additional preferred routes for
highway shipments (49 CFR 397.103). At this time the State of Nevada has not identified any alternative
or additional preferred routes that DOE could use for shipments to the repository.
Selection of Rail Routes. Rail transportation routing of spent nuclear fuel and high-level radioactive
waste shipments is not regulated by the Department of Transportation. As a consequence, the routing
rules used by the INTERLINE computer program (Johnson et al. 1993b, all) assumed that railroads would
select routes using historic practices. DOE has determined that the INTERLINE program is appropriate
for calculating routes and related information for use in transportation analyses (Maheras and Pipp)en
1995, pages 2 to 5). Because the routing of rail shipments would be subject to future, possibly different
practices of the involved railroads, DOE could use other rail routes.
For the 19 commercial sites that have the capability to handle and load rail casks but do not have direct
rail service, DOE used the HIGHWAY computer program to identify routes for heavy-haul transportation
to nearby railheads. For such routes, routing agencies in affected states would need to approve the
transport and routing of overweight and overdimensional shipments.
J.1 .2.2.2 Routes for Shipping Rail Casl<s from Sites Not Served by a Railroad
In addition to routes for legal-weight trucks and rail shipments, 19 commercial sites that are not served by
a railroad, but that have the capability to load rail casks, could ship spent nuclear fuel to nearby railheads
using heavy-haul trucks (see Table J-12). Fourteen of these sites are on navigable waterways; some of
these could ship by barge to railheads. Distances to the nearest railheads for barge shipments were
estimated for each of the 14 reactor sites. These distances are listed in Table J-13.
J.1 .2.2.3 Sensitivity of Analysis Results to Routing Assumptions
Routing for shipments of spent nuclear fuel and high-level radioactive waste to the proposed repository
would comply with regulations of the Department of Transportation and the Nuclear Regulatory
Commission in effect at the time shipments would occur. Unless the State of Nevada designates
alternative or additional preferred routes, to comply with Department of Transportation regulations all
preferred routes would ultimately enter Nevada on Interstate 15 and travel to the repository on U.S.
Highway 95. States can designate alternative or additional preferred routes for highway shipments. At
this time the State of Nevada has not identified any alternative or additional preferred routes DOE could
use for shipments to the repository. Section J.3.1.3 examines the sensitivity of transportation impacts
both nationally and regionally (within Nevada) to changes in routing assumption within Nevada.
t
J-33
Transportation
Table J-13. Barge transportation distances from sites to intermodal rail nodes (kilometers).
a,b
Site State Total'' Rural Suburban Urban
Browns Ferry
Diablo Canyon
St. Lucie
Turkey Point
Calvert Cliffs
Palisades
Grand Gulf
Cooper
Salem/Hope Creek
Oyster Creek
Surry
Kewaunee
Point Beach
AL
57
52
5
0
CA
143
143
0
0
FL
140
50
52
39
FL
54
53
0
1
MD
99
98
2
0
MI
256
256
0
0
MS
51
51
0
0
ME
117
100
16
1
NJ
30
30
0
0
NJ
130
77
36
17
VA
71
60
8
3
WI
293
285
2
7
WI
301
293
2
7
a. To convert kilometers to miles, multiply by 0.62137.
b. Distances estimated with INTERLINE (Johnson et al. 1993b, all).
c. Intermodal rail nodes selected for purpose of analysis. Source: TRW (1999a, Section 4).
d. Totals might differ from sums due to methods of calculation and rounding.
J.1.3 ANALYSIS OF IMPACTS FROM INCIDENT-FREE TRANSPORTATION
DOE analyzed the impacts of incident-free transportation for shipments of commercial and DOE spent
nuclear fuel and DOE high-level radioactive waste that would be shipped under the Proposed Action and
Inventory Modules 1 and 2 from 77 sites to the repository. The analysis estimated impacts to the public
and workers and included impacts of loading shipping casks at commercial and DOE sites and other
preparations for shipment as well as intermodal transfers of casks from heavy-haul trucks or barges to rail
cars.
J.1.3.1 Methods and Approach for Analysis of Impacts for Loading Operations
The analysis used methods and assessments developed for spent nuclear fuel loading operations at
commercial sites to estimate radiological impacts to involved workers at commercial and DOE sites.
Previously developed conceptual radiation shield designs for shipping casks (Schneider et al. 1987,
Sections 4 and 5), rail and truck shipping cask dimensions, and estimated radiation dose rates at locations
where workers would load and prepare casks (Smith, Daling, and Faletti 1992, page 4.2) for shipment
were the analysis bases for loading operations. In addition, tasks and time-motion evaluations from these
studies were used to describe spent nuclear fuel handling and loading. These earlier evaluations were
based on normal, incident-free operations that would be conducted according to Nuclear Regulatory
Commission regulations that establish radiation protection criteria for workers.
The analysis assumed that noninvolved workers would not have tasks that would result in radiation
exposure. In a similar manner, the analysis projected that the dose to the public from loading operations
would be extremely small, resulting in no or small impacts. A separate evaluation of the potential
radiation dose to members of the public from loading operations at commercial nuclear reactor facilities
showed that the dose would be very low, less than 0.(X)1 person-rem per metric ton uranium of spent
nuclear fuel loaded (DOE 1986, page 2.42, Figure 2.9). Public doses from activities at commercial and
DOE sites generally come from exposure to airborne emissions and, in some cases, waterbome effluents
containing low levels of radionuclides. However, direct radiation at publicly accessible locations near
these sites typically is not measurable and contributes negligibly to public dose and radiological impacts.
Though DOE expects no releases from loading operations, this analysis estimated that the dose to the
public would be 0.001 person-rem per metric ton uranium, and metric ton equivalents, for DOE spent
nuclear fuel and high-level radioactive waste. Noninvolved workers could also be exposed to low levels
J-34
Transportation
of radioactive materials and radioactivity from loadout operations. However, because these workers
would not work in radiation areas they would receive a very small fraction of the dose received by
involved workers. DOE anticipates that noninvolved workers would receive individual doses similar to
those received by members of the public. Because the population of noninvolved workers would be small
compared to the population of the general public near the 77 sites, the dose to these workers would be a
small fraction of the public dose.
The analysis used several basic assumptions to evaluate impacts from loading operations at DOE sites:
• Operations to load spent nuclear fuel and high-level radioactive waste at DOE facilities would be
similar to loading operations at commercial facilities.
• Commercial spent nuclear fuel would be in storage pools or in dry storage at the reactors and DOE
spent nuclear fuel would be in dry storage, ready to be loaded directly in Nuclear Regulatory
Commission-certified shipping casks and then on transportation vehicles. In addition, DOE high-
level radioactive waste could be loaded directly in casks. All preparatory activities, including
packaging, repackaging, and validating the acceptability of spent nuclear fuel for acceptance at the
repository would be complete prior to loading operations.
• Commercial spent nuclear fuel to be placed in the shipping casks would be uncanistered or canistered
fuel assemblies, with at least one assembly in a canister. DOE spent nuclear fuel and high-level
radioactive waste would be in disposable canisters. Typically, uncanistered assemblies would be
loaded into shipping casks under water in storage pools (wet storage). Canistered spent nuclear fuel
could be loaded in casks directly from dry storage facilities or storage pools.
In addition, because handling and loading operations for DOE spent nuclear fiiel and high-level
radioactive waste and commercial spent nuclear fuel would be similar, the analysis assumed that impacts
to workers during the loading of commercial spent nuclear fuel could represent those for the DOE
materials, even though the radionuclide inventory of commercial fuel and the resultant external dose rate
would be higher than those of the DOE materials. This conservative assumption of selecting impacts
from commercial handling and loading operations overestimated the impacts of DOE loading operations,
but it enabled the use of detailed real information developed for commercial loading operations to assess
impacts for DOE operations. Equivalent information was not available for operations at DOE facilities.
To gauge the conservatism of the assumption DOE compared the radioactivity of contents of shipments of
commercial and DOE spent nuclear fuel and high-level radioactive waste. Table J-14 compares typical
inventories of important contributors to the assessment of worker and public health impacts. These are
cesium-137 and actinide isotopes (including plutonium) for rail shipments of commercial spent nuclear
fuel, DOE spent nuclear fuel, and DOE high-level radioactive waste. Although other factors are also
important (for example, material form and composition), these indicators provide an index of the relative
hazard potential of the materials. Appendix A contains additional information on the radionuclide
inventory and characteristics of spent nuclear fiiel and high-level radioactive waste.
J.1 .3.1 .1 Radiological Impacts of Loading Operations at Commercial Sites
In 1987, DOE published a study of the estimated radiation doses to the public and workers resulting from
the transport of spent nuclear fuel from commercial nuclear power reactors to a hypothetical deep
geologic repository (Schneider et al. 1987, all). This study was based on a single set of spent nuclear fuel
characteristics and a single split [30 percent/70 percent by weight; 900 metric tons uranium/2, 1(X) metric
tons uranium per year] between truck and rail conveyances. DOE published its findings on additional
radiological impacts on monitored retrievable storage workers in an addendum to the 1987 report (Smith,
Daling, and Faletti 1992, all). The technical approaches and impacts summarized in these DOE reports
J-35
Transportation
Table J-14. Typical cesium- 137, actinide isotope, and total radioactive material content (curies) in a rail
shipping cask."
Actinides
Material
Cesium- 137
(excluding uranium)*"
Total
Commercial spent nuclear fuel
810,000
650,000
2,000,000
High-level radioactive waste
120,000
40,000"=
280,000
DOE spent nuclear fuel (except naval
260,000
160,000
620,000
spent nuclear fuel)
Naval spent nuclear fuel
550,000
30,000
1,200,000
a. Source: Appendix A. Source estimated based on 36 typical pressurized-water reactor fuel assemblies for commercial spent
nuclear fuel; one dual-purf)ose shipping canister for naval spent fuel; five canisters of DOE spent nuclear fuel; and five
canisters of high-level radioactive waste.
b. Uranium would not be an important contributor to health and safety risk.
c. Includes plutonium can-in-canister with high-level radioactive waste.
were used to project involved worker impacts that would result from commercial at-reactor spent nuclear
fuel loading operations. DOE did not provide a separate analysis of noninvolved worker impacts in these
reports. For the analysis in this EIS, DOE assumed that noninvolved workers would not receive radiation
exposures from loading operations. This assumption is appropriate because noninvolved workers would
be personnel with managerial or administrative support functions directly related to the loading tasks but
at locations, typically in offices, away from areas where loading activities took place.
In the DOE study, worker impacts from loading operations were estimated for a light-water reactor with
pool storage of spent nuclear fuel. The radiological characteristics of the spent nuclear fuel in the analysis
was 10-year-old, pressurized-water reactor fuel with an exposure history (bumup) of 35,000 megawatt-
days per metric ton. In addition, the reference pressurized-water reactor and boiling-water reactor fuel
assemblies were assumed to contain 0.46 and 0.19 MTU, respectively, prior to reactor irradiation. These
parameters for spent nuclear fuel are similar to those presented in Appendix A of this EIS. The use of the
parameters for spent nuclear fuel presented in Appendix A would be likely to lead to similar results.
In the 1987 study, radiation shielding analyses were done to provide information on (1) the conceptual
configuration of postulated reference rail and truck transportation casks, and (2) the direct radiation levels
at accessible locations near loaded transportation casks. The study also presented the results of a detailed
time-motion analysis of work tasks that used a loading concept of operations. This task analysis was
coupled with cask and at-reactor direct radiation exposure rates to estimate radiation doses to involved
workers (that is, those who would participate directly in the handling and loading of the transportation
casks and conveyances). Impacts to members of the public from loading operations had been shown to be
small [fraction of a person-millirem population dose; (Schneider et al. 1987, page 2.9)] and were
eliminated from further analysis in the 1987 report. The at-reactor-loading concept of operations included
the following activities:
1 . Receiving the empty transportation cask at the site fence
2. Preparing and moving the cask into the facility loading area
3. Removing the cask from the site prime mover trailer
4. Preparing the cask for loading and placing it in the water-filled loading pit
5. Transferring spent nuclear fuel from its pool storage location to the cask
6. Removing the cask from the pool and preparing it for shipment
J-36
Transportation
Rail"
Truck'
Total
2,100
900
3,000
6.5/6.70
0.92y0.93
NA'
320
970
1,290
2.3/2.5
1.3/1.4
NA
0.06/0.077
0.29/0.31
NA
7. Placing the cask on the site prime mover trailer
8. Moving the loaded cask to the site fence where the trailer is connected to the transportation carrier's
prime mover for offsite shipment
The results for loading operations are listed in Table J-15.
Table J-15. Principal logistics bases and results for the reference at-reactor loading operations."
Conveyance
Parameter
Annual loading rate (MTU/year)''
Transportation cask capacity, PWR - BWR (MTU/cask)
Annual shipment rate (shipments/year)
Average loading duration,' PWR - BWR (days)
Involved worker specific CD,^ PWR - BWR (person-rem/MTU)
a. Source: Schneider et al. (1987, pages 2.5 and 2.7).
b. 14 pressurized-waste reactor and boiling-water reactor spent nuclear fuel assemblies per rail transportation cask.
c. 2 pressurized-waste reactor and boiling-water reactor spent nuclear fuel assemblies per tmck transportation cask.
d. MTU = metric tons of uranium.
e. NA = not applicable.
f. Based on single shift operations; carrier dropy-off and pick-up delays were not included.
g. Collective dose expressed as the sum of the doses accumulated by all loading (involved) workers, regardless of the total
number of workers assigned to loading tasks.
The loading activities that the study determined would produce the highest collective unit impacts are
listed in Table J- 16. As listed in this table, the involved worker collective radiation doses would be
dominated by tasks in which the workers would be near the transportation cask when it contained spent
nuclear fuel, particularly when they were working around the cask lid area. These activities would deliver
at least 40 percent of the total collective worker doses. Worker impacts from the next largest dose-
producing tasks (working to secure the transportation cask on the trailer) would account for 12 to 19
percent of the total impact. The impacts are based on using crews of 13 workers [the number of workers
assumed in the Schneider et al. (1987, Section 2) study] dedicated solely to performing cask-handling
work. The involved worker collective dose was calculated using the following formula:
Collective dose (person-rem) = AxBxCxDxE
where: A = number of pressurized-water or boiling-water reactor spent nuclear fuel shipments being
analyzed under each transportation scenario (from Tables J-5 and J-6)
B = number of transportation casks included in a shipment (set at 1 for both transportation
scenarios)
C = number of pressurized-water or boiling-water reactor spent nuclear fuel assemblies in a
transportation cask (from Table J-3)
D = amount of uranium in the spent nuclear fuel assembly prior to reactor irradiation,
expressed as metric tons uranium per assembly (from Table J-15)
E = involved worker-specific collective dose in person-rem/metric ton uranium for each fuel
type (from Table J-15)
J-37
Transportation
Table J-16. At-reactor reference loading operations — collective impacts to involved workers.^
Rail Truck
CD/MTU"
Percent of
CD/MTU
Percent of
Task description
(PWR - BWR)'^
total impact
(PWR - BWR)
total impact
Install cask lids; flush cask interior;
0.025/0.024
40/31
0.126/0.126
43/40
drain, dry and seal cask
Install cask binders, impact limiters.
0.010/0.009
15/12
0.056/0.055
19/18
personnel barriers
Load SNF into cask
0.011/0.027
17/35
0.011/0.027
4/9
On-vehicle cask radiological
0.003/0.003
5/4
0.018/0.018
6/6
decontaminadon and survey
Final inspection and radiation surveys
0.002/0.002
4/3
0.016/0.015
5/5
All other (19) activities
0.011/0.012
19/16
0.066/0.073
23/23
Task totals
0.062/0.077
100/100
0.29/0.31
100/100
a. Source: Schneider et al. (1987, page 2.9).
b. CD/MTU = Collective dose (f)erson-rem effective dose equivalent) per metric ton uranium. The at-reactor loading
c. crew size is 13 involved workers.
d. PWR = pressurized-water reactor; BWR = boiling-water reactor.
Because worker doses are linked directly to the number of loading operations performed, the highest
average individual doses under each transportation scenario would occur at the reactor sites having the
most number of shipments. Accordingly, the average individual dose impacts were calculated for the
limiting site using the equation:
Average individual dose (rem per involved worker) = (AxBxCxDxE) + F
where: A = largest value for the number of shipments from a site under each transportation scenario
(from Tables J-5 and J-6)
B = number of transportation casks included in a shipment (set at 1 for both transportation
options)
C = number of spent nuclear fuel assemblies in a transportation cask (from Table J-3)
D = amount of uranium in the spent nuclear fuel assembly prior to reactor irradiation in metric
tons uranium per assembly (from Table J- 15)
E = involved worker-specific collective dose in person-rem per metric ton uranium for each
fuel type (from Table J- 15)
F = involved worker crew size (set at 13 persons for both transportation options; from
Table J-16)
J.I .3.1 .2 Radiological Impacts of DOE Spent Nuclear Fuel and High-Level Radioactive
Waste Loading Operations
The methodology used to estimate impacts to workers during loading operations for commercial spent
nuclear fuel was also used to estimate impacts of loading operations for DOE spent nuclear fuel and high-
level radioactive waste. The exposure factor for loading boiling-water reactor spent nuclear fuel in truck
casks at commercial facilities (person-rem per MTU) was used (see Table J-16). The exposure factor for
truck shipments of boiling-water reactor spent nuclear fuel was based on a cask capacity of five
J-38
Transportation
boiling-water reactor spent nuclear fuel assemblies (about 0.9 MTHM). The analysis used this factor
because it would result in the largest estimates for dose per operation.
J.I. 3.2 Methods and Approach for Analysis of Impacts from Incident-Free Transportation
The potential exists for human health impacts to workers and members of the public from incident-free
transportation of spent nuclear fuel and high level radioactive waste. Incident-free transportation means
normal accident-free shipment operations during which traffic accidents and accidents in which
radioactive materials could be released do not occur; these are addressed separately in Section J. 1.4.
Incident-free impacts could occur from exposure to (1) external radiation in the vicinity of the
transportation casks, or (2) transportation vehicle emissions, both during normal transportation.
J.1 .3.2.1 Incident-Free Radiation Dose to Populations
The analysis used the RADTRAN4 computer program (Neuhauser and Kanipe 1992, all) to evaluate
incident-free impacts for populations. The RADTRAN4 input parameters used to estimate incident-free
impacts are listed in Table J-17. Through extensive review (Maheras and Pippen 1995, Section 3 and 4),
DOE has determined that this program provides valid estimates of population doses for use in the
evaluation of risks of transporting radioactive materials, including spent nuclear fuel and high-level
radioactive waste. DOE has used the RADTRAN4 code to analyze transportation impacts for other
environmental impact statements (for example, DOE 1995, Appendix E; DOE 1997b, Appendixes F and
G). The program used population densities from 1990 census data to calculate the collective dose to
populations that live along transportation routes [within 800 meters (0.5 mile) of either side of the route].
Table J- 18 lists the estimated number of people who live within 800 meters of national routes.
The analysis used five kinds of information to estimate collective doses to populations:
• External radiation dose rate around shipping casks
• Number of people who would live within 8(X) meters (0.5 mile) along the routes of travel
• Distances individuals would live from the routes
• Amount of time each individual would be exposed as a shipment passed by
• Number of shipments that would be transported over each route
The first four were developed using the data listed in Table J-19. The fifth kind of information (the
number of shipments that would use a transportation route) was developed with the use of the CALVIN
computer program discussed in Section J. 1.1.1, the DOE Throughput Study (TRW 1997, Section 6.1.1),
data on DOE spent nuclear fuel and high-level radioactive waste inventories in Appendix A, and data
from DOE sites (Jensen 1998, all). The analysis used CALVIN to estimate the number of shipments from
each commercial site. The Throughput Study provided the estimated number of shipments of high-level
radioactive waste from the four DOE sites. Information provided by the DOE National Spent Nuclear
Fuel Program (Jensen 1998, all) and in Appendix A was used to estimate shipments of DOE spent nuclear
fuel.
The analysis used a value of 10 millirem per hour at a distance of 2 meters (6.6 feet) from the side of a
transport vehicle for the external dose rate around shipping casks. This value is the maximum allowed by
regulations of the Department of Transportation for shipments of radioactive materials [49 CFR
173.441(b)]. Dose rates at distances greater than 2 meters from the side of a vehicle would be less. The
dose rate at 30 meters (1(X) feet) from the vehicle would be less than 0.2 millirem per hour; at a distance
of 800 meters (2,625 feet) the dose rate would be less than 0.(XX)2 millirem per hour.
J-39
Transportation
Table J-17. Input parameters and parameter values used for
trancnnrtntir\n nn^lvcic
the incident-free national truck and rail
transportation analysis.
Parameter
Legal-weight truck
transportation
Rail
transportation
Legal-weight truck
and rail
Package type
Type B shipping cask
Package dimension
4.77 meters" long
Dose rate
10 millirem per hour,
2 meters from side of
vehicle
Number of crewmen
2
5
Distance from source to crew
Speed
Rural
3 meters
88 km*" per hour
152 meters
64 km per hour
Suburban
40 km per hour
Urban
24 km per hour
Stop time per km
0.011 hours per km
0.033 hours per km"^
Number of people exposed while stopped
50
Based on suburban
population density
Number of people per vehicle sharing
route
Population densities (persons per kn^f
Rural
Suburban
Urban
One-way traffic count (vehicles per hour)
Rural
Suburban
Urban
2
470
780
2,800
3
1
5
5
(e)
(e)
(e)
a. To convert meters to feet, multiply by 3.2808.
b. To convert kilometers (km) to miles, multiply by 0.62137.
c. Assumes general freight rather than dedicated service.
d. To convert square kilometers to square miles, multiply by 0.3861.
e. Population densities along transpwrtation routes were estimated using the HIGHWAY and INTERLINE computer programs.
These programs used 1990 Census data.
Table J-18. Population within 800 meters (0.5 mile) of routes
for incident-free transportation using 1990 census data.
Transportation scenario 1990 Census data
Mostly legal-weight truck
Mostly rail
7,200,000
11,100,000
a. Source: TRW (1999a, pages 18 and 19).
The second kind of information used in the analysis was the number of people who potentially would be
close enough to shipments to be exposed to radiation from the casks. The analysis determined the
estimated offlink number of people [those within the 1.6-kilometer (1-mile) region of influence] by
multiplying the population densities (persons per square kilometer) in population zones through which a
route would pass by the 1.6-kilometer width of the region of influence and by the length of the route
through the population zones. Onlink populations (those sharing the route and people at stops along the
route) were estimated using assumptions from other EISs that have evaluated transportation impacts
(DOE 1995, Appendix I; DOE 1996a, Appendix E; DOE 1997b, Appendixes F and G). The travel
distance in each population zone was determined for legal-weight truck shipments by using the
HIGHWAY computer program (Johnson et al. 1993a, all) and for rail shipments by using the
J-40
Transportation
Table J-19. Information used for analysis of incident-free transportation impacts.
Travel speed
Population within (kilometers per hour)
800 meters' Legal- weight Heavy-haul Dose rate 2 meters'" from
Population zones (per kilometer of route) truck truck Rail vehicle (millirem per hour)
Urban
(c)
24
24
24"
10
Suburban
(c)
40"
40
40
10
Rural
(c)
88
40
64
10
a. 800 meters = about 2,600 feet.
b. 2 meters = about 6.6 feet.
c. Estimates of population within 800 meters of a route are based on analysis of census block data using HIGHWAY (Johnson
et al. 1993a, all) and INTERLINE (Johnson et al. 1993b, all) computer programs. The analysis used actual populations
along routes based on the 1990 Census.
d. Analysis of impacts for shipments of naval spent nuclear fuel used 40 kilometers (25 miles) per hour for heavy-haul truck
speed and 24 kilometers (15 miles) f)er hour for train speed in urban, suburban, and rural zones.
INTERLINE program (Johnson et al. 1993b, all). These programs used 1990 census block group data to
identify where highways and railroads enter and exit each type of population zone, which the analysis
used to determine the total lengths of the highways and railroads in each population zone.
The third kind of information — the distances individuals live from the route used in the analysis — is the
estimated the number of people who live within 8(X3 meters (about 2,600 feet) of the route. The analysis
assumed that population density is uniform in population zones.
The determination of the fourth kind of information used in the analysis — the time that people could be
exposed as shipments passed — was based on the assumed travel speed of shipments in each population
zone along the route. For example, travel at 24 kilometers (15 miles) an hour in urban areas would lead to
a longer exposure time than travel at 88 kilometers (55 miles) an hour in rural areas. Persons in vehicles
traveling along a route with a shipment of spent nuclear fuel or high-level radioactive waste or persons
who lived near railyards where shipments would be switched between trains could be exposed for longer
periods.
With the five kinds of information, the analysis used RADTRAN4 to calculate exposures for the
following groups:
• Public along the route (Off link Exposure): Collective doses for persons living or working within
0.8 kilometer (0.5 mile) on each side of the transportation route.
•
•
Public sharing the route (Onlink Exposure): Collective doses for persons in vehicles sharing the
transportation route; this includes persons traveling in the same or opposite direction and those in
vehicles passing the shipment.
Public during stops (Stops): Collective doses for people who could be exposed while a shipment
was stopped en route. For truck transportation, these would include stops for refueling, food, and
rest. For rail transportation, stops would occur in railyards along the route to switch railcars from
inbound trains to outbound trains traveling toward the Yucca Mountain site, and to change train crews
and equipment (locomotives).
Worker exposure (Occupational Exposure): Collective doses for truck and rail transportation
crew members.
J-41
Transportation
• Security escort exposure (Occupational Exposure): Collective doses for security escorts. In
calculating doses to workers the analysis conservatively assumed that the maximum number of
escorts required by regulations (10 CFR 73.37) would be present for urban, suburban, and rural
population zones.
The sum of the doses for the first three categories is the total nonoccupational (public) dose.
Unit dose factors were used to calculate collective dose. These factors, which are listed in Table J-20,
represent the dose that would be received by a population of 1 person per square kilometer for one
shipment of radioactive material moving a distance of 1 kilometer (0.62 mile) in the indicated population
density zone. The unit dose factors for incident-free transportation reflect the assumption that the dose
rate external to shipments of spent nuclear fuel and high-level radioactive waste would be the maximum
value allowed by Department of Transportation regulations — 10 millirem per hour at 2 meters (6 feet)
from the side of the transport vehicle (49 CFR 173.441). The incident-free dose from transporting a
single shipment was determined by multiplying the appropriate unit dose factors by corresponding
distances in each of the population zones the shipment route passes through and the population density of
the zone. The collective dose from all shipments from a site were determined by multiplying the dose
from a single shipment by the number of shipments that would be required to transport the site's spent
nuclear fuel or high-level radioactive waste to the repository. Collective dose was converted to the
estimated number of latent cancer fatalities using conversion factors recommended by the International
Commission on Radiological Protection (ICRP 1991, page 22). These values are 0.0004 for radiation
workers and 0.0005 for the general population.
Table J-20. Unit dose factors for incident-free national truck and
rail transportation of spent nuclear fuel and high-level radioactive
waste.
Unit dose factors
Exposure
group
(person-rem per kilometer)^
Mode
Rural
Suburban
Urban
Truck
Involved worker
4.56x10"'
1x10""
1.67x10"*
Public
Offlink"
3.2x10"*
3.52x10"*
4.33x10"*
Onlink'
7.81x10-*
2.25x10"'
2.32x10"*
Stops
1.87x10"*
1.87x10-*
1.87x10"*
Rail
Involved worker^
1.22x10"'
1.22x10"'
1.22x10"'
Public
Offlink
4.38x10"*
7.02x10"*
1.17x10"'
Onlink
1.03x10"'
1.32x10"*
3.65x10*
Stops'
7.42x10"*
7.42x10"*
7.42x10*
a. The methodology, equations, and data used to develop the unit dose factors
are discussed in Madsen et al. (1986, all) and Neuhauser and Kanipw (1992,
page 4-15). Cashwell et al. (1986, page 44) contains a detailed explanation of
the use of unit factors.
b. Offlink general population included persons within 800 meters (2,625 feet) of
the road or railway.
c. Onlink general population included persons sharing the road or railway.
d. The nonlinear component of incident-free rail dose for crew workers because
of railcar inspections and classifications is 0.014 person-rem per shipment.
Ostmeyer (1986, all) contains a detailed explanation of the rail exposure
model.
e. The nonlinear component of incident-free rail dose for the general population
because of railcar inspections and classifications is 0.0014 person-rem per
shipment. Ostmeyer (1986, all) contains a detailed explanation of the rail
exposure model.
J-42
Transportation
J.I .3.2.2 Methods Used To Evaluate Incident-Free Impacts to Maximally Exposed
Individuals.
To estimate impacts to maximally exposed individuals, the same kinds of information as those used for
population doses (except for population size) was needed. The analysis of doses to maximally exposed
individuals used projected exposure times, the distance a hypothetical individual would be from a
shipment, the number of times an exposure event could occur, and the assumed external radiation dose
rate 2 meters (6.6 feet) from a shipment (10 millirem per hour). These analyses used the RISKIND
computer program (Yuan et al. 1995, all). DOE has used RISKEW for analyses of transportation impacts
in other environmental impact statements (DOE 1995, Appendix J; DOE 1996a, Appendix E; DOE
1997b, Appendix E). RISKIND provides appropriate results for analyses of incident-free transportation
and transportation accidents involving radioactive materials (Maheras and Pippen 1995, Sections 5.2 and
6.2; Biweretal. 1997, all).
The maximally exposed individual is a hypothetical person who would receive the highest dose. Because
different maximally exposed individuals can be postulated for different exposure scenarios, the analysis
evaluated the following exposure scenarios.
•
•
Crew Members. In general, truck crew members, including security escorts and rail security
escorts, would receive the highest doses during incident-free transportation (see discussion in
J. 1.3.2.2.1 below). The analysis assumed that the crews would be limited to a total job-related
exposure of 2 rem per year (DOE 1994, Article 21 1).
Inspectors (Truck and Rail). Inspectors would be Federal or state vehicle inspectors. On the basis
of information provided by the Commercial Vehicle Safety Alliance (Battelle 1998, all; CVSA 1999,
all), the analysis assumed an average exposure distance of 1 meter (3 feet) and an exposure duration
of 1 hour (see discussion in J. 1.3.2.2).
Railyard Crew Member. For a railyard crew member working in a rail classification yard
assembling trains, the analysis assumed an average exposure distance of 10 meters (33 feet) and an
exposure duration of 2 hours (DOE 1997b, page E-50).
Resident. The analysis assumed this maximally exposed individual is a resident who lives 30 meters
(100 feet) from a point where shipments would pass. The resident would be exposed to all shipments
along a particular route (DOE 1995, page 1-52).
Individual Stuck in Traffic (Truck or Rail). The analysis assumed that a member of the public
could be 1.2 meter (4 feet) from the transport vehicle carrying a shipping cask for 1 hour. Because
these circumstances would be random and unlikely to occur more than once for the same individual,
the analysis assumed the individual to be exposed only once.
Resident near a Rail Stop. The analysis assumed a resident who lives within 200 meters (660 feet)
of a switchyard and an exposure time of 20 hours for each occurrence. The analysis of exposure for
this maximally exposed individual assumes that the same resident would be exposed to all rail
shipments to the repository (DOE 1995, page 1-52).
Person at a Truck Service Station. The analysis assumed that a member of the public (a service
station attendant) would be exposed to shipments for 1 hour for each occurrence at a distance of
20 meters (70 feet). The analysis also assumed this individual would work at a location where all
truck shipments would stop.
J-43
Transportation
As discussed above for exposed populations, the analysis converted radiation doses to estimates of
radiological impacts using dose-to-risk conversion factors of the International Commission on
Radiological Protection.
J.1 .3.2.2.1 Incident-Free Radiation Doses to Inspectors. DOE estimated radiation doses to the
state insp)ectors who would inspect shipments of spent nuclear fuel and high-level radioactive waste
originating in, passing through, or entering a state. For legal-weight truck and railcar shipments, the
analysis assumed that:
• Each inspection would involve one individual working for 1 hour at a distance of 2 meters (6.6 feet)
from a shipping cask.
• The radiation field surrounding the cask would be the maximum permitted by regulations of the
Department of Transportation (49 CFR 173.441).
• There would be no shielding between an inspector and a cask.
For rail shipments, the analysis assumed that:
• There would be a minimum of two inspections per trip — one at origin and one at destination — with
additional inspections in route occurring about once every 500 kilometers (300 miles) of railcar
travel.
• Rail crews would conduct the remaining along-the-route inspections.
For legal-weight truck shipments, the analysis assumed that:
• On average, state officials would conduct two inspections during each trip - one at the origin and one
at the destination.
• The inspectors would use the Enhanced North American Uniform Inspection Procedures and Out-of-
Service Criteria for Commercial Highway Vehicles Transporting Transuranics, Spent Nuclear Fuel,
and High-Level Radioactive Waste (CVS A 1999, all).
• The shipments would receive a Commercial Vehicle Safety Alliance inspection sticker on passing
inspection and before departing from the 77sites.
• Display of such a sticker would provide sufficient evidence to state authorities along a route that a
shipment complied with Department of Transportation regulations (unless there was contradictory
evidence), and there would be no need for additional inspections.
The analysis determined doses to state inspectors in two ways. For rail shipments, inspector doses were
based on the equations and assumptions used in the RADTRAN4 computer program. The program uses
an empirically derived equation that is based on observations of rail classification yard operations, as
follows:
Dose = Ko X dose rate x casks per shipment x number of shipments x 0. 16 x 0.(X)1
where:
dose = rem of exposure to an inspector
J-44
Transportation
Ko = a shape factor for the cask assumed for purposes of analysis (meters);
6 meters for rail cask that would ship spent nuclear fuel
dose rate = the dose rate in millirem per hour 1 meter from the surface of the
cask; set to 14 millirem per hour for the analysis
casks per shipment = the average number of casks (one cask per railcar) in a train; set to 1
for the analysis
number of shipments = number of shipments inspected (set to 1 for the analysis)
0. 16 = exposure factor that translates the product of cask dose rate and shape
factor into inspector dose (meters per hour)
0.001 = conversion factor to convert millirem per hour to rem per hour.
The equation shows that the calculated value for whole-body dose to an individual inspector for one
inspection would be 13.4 millirem. An inspector in Nevada who inspected all rail shipments under the
mostly rail scenario would receive a whole body dose of 470 x 13.4 = 6.3 rem in a year. If the same
inspector inspected all shipments over the 24 years of the Proposed Action, he or she would be exposed to
150 rem. Using the dose to risk conversion factors published by the Intemational Commission on
Radiation Protection, this exposure would increase the likelihood of the inspector incurring a fatal cancer.
This would add 6 percent to the likelihood for fatal cancers from all other causes, increasing the
likelihood from approximately 23 percent (ACS 1998, page 10) to 29 percent.
For shipments by legal-weight truck, the analysis used the RISKIND computer program to estimate doses
to inspectors (Yuan et al. 1995, all). The data used by the code to calculate dose includes the estimated
value for dose rate at 1 meter (3.3 feet) from a cask surface, the length and diameter of the cask, the
distance between the location of the individual and the cask surface, and the estimated time of exposure.
For this calculation, the analysis assumed that an inspector following Commercial Vehicle Safety
Alliance procedures (CVSA 1999, all) would work for 1 hour at an average distance of 2 meters (6.6 feet)
from the cask. The analysis assumed that a typical legal-weight truck cask would be about 1 meter in
diameter and about 5 meters (16 feet) long and that the dose rate 1 meter from the cask surface would be
14 millirem per hour. A dose rate of 14 millirem per hour 1 meter from the surface of a truck cask is
approximately equivalent to the maximum dose rate allowed by Department of Transportation regulations
for exclusive-use shipments of radioactive materials (49 CFR 173.441).
Using this data, the RISKIND computer code calculated an expected dose of 18 millirem for an individual
inspector. Under the mostly legal-weight truck scenario in which approximately 2,100 legal -weight truck
shipments would arrive in Nevada annually, a Nevada inspector working 1,8(X) hours per year could
inspect as many as 470 shipments in a year. This inspector would receive a whole-body dose of 8.5 rem.
If this same inspector inspected all shipments over the 24 years of the Proposed Action, he or she would
be exposed to 204 rem. Using the dose to risk conversion factors published by the Intemational
Commission on Radiation FVotection, this exposure would increase the likelihood of this individual
contracting a fatal cancer. This would add about 8 percent to the likelihood for fatal cancers from all
other causes, increasing the likelihood from approximately 22 percent (ACS 1998, page 10) to 32 percent.
Under the mostly legal-weight truck scenario, the annual committed dose to inspectors in a state that
inspected all incoming legal-weight truck shipments containing spent nuclear fuel or high-level
radioactive waste would be about 38 person-rem. Over 24 years, the population dose for these inspectors
would be about 910 person-rem. This would result in about 0.34 latent cancer fatality (this is equivalent
J-45
Transportation
to a 36-percent likelihood that there would be 1 additional latent cancer fatality among the exposed
group).
DOE implements radiation protection programs at its facilities where there is the potential for worker
exposure to cumulative doses from ionizing radiation. The Department anticipates that the potential for
individual whole-body doses such as those reported above would lead an involved state to implement
such a radiation protection program. If similar to those for DOE facilities, the administrative control limit
on individual dose would not exceed 2 rem per year (DOE 1994, Article 21 1) and the expected maximum
exposure for inspectors would be less than 500 millirem per year.
J.I .3.2.2.2 Incident-Free Radiation Doses to Escorts. Transporting spent nuclear fuel to the
Yucca Mountain site would require the use of physical security and other escorts for the shipments.
Regulations (10 CFR 73.37) require escorts for highway and rail shipments. These regulations require
two escorts (individuals) for truck shipments traveling in highly populated (urban) areas. One of the
escorts must be in a vehicle that is separate from the shipment vehicle. For rail shipments in urban areas,
at least two escorts must maintain visual surveillance of a shipment from a railcar that accompanies a cask
car.
In areas that are not highly populated (suburban and rural), one escort must accompany truck shipments.
The escort can ride in the cab of the shipment vehicle. At least one escort is required for rail shipments in
suburban and rural areas. However, for rail shipments, the escort must occupy a railcar that is separate
from the cask car and must maintain visual surveillance of the shipment at all times.
For legal-weight truck shipments, the analysis assumed that a second driver, who would be a member of
the vehicle crew, would serve as an escort in all areas. The analysis assigned a second escort for travel in
urban areas and assumed that this escort would occupy a vehicle that followed or led the transport vehicle
by at least 60 meters (about 200 feet). The analysis assumed that the dose rate at a location 2 meters
(6.6 feet) behind the vehicle would be 10 millirem per hour, which is the limit allowed by Department of
Transportation regulations (49 CFR 173.441). Using this information, the analysis used the RISKIND
computer program to calculate a value of approximately 0. 1 1 millirem per hour for the dose rate 60
meters behind the transport vehicle; this is the estimated value for the dose rate in a following escort
vehicle. The value for the dose rate in an escort vehicle that preceded a shipment would be lower.
Because the dose rate in the occupied crew area of the transport vehicle would be less than 2 millirem per
hour, the dose rate 2 meters in front of the vehicle would be much less than 10 millirem per hour, the
value assumed for a location 2 meters behind the vehicle. The value of 2 millirem per hour in normally
occupied areas of transport vehicles is the maximum allowed by Department of Transportation
regulations (49 CFR 173.441).
To calculate the dose to escorts, the analysis assumed that escorts in separate vehicles would be required
in urban areas as shipments traveled to the Yucca Mountain site. The calculations used the RISKIND
computer program (Yuan et al. 1995, all); the distance of travel in urban areas provided by the
HIGHWAY and INTERLINE computer codes; and the estimated speed of travel in urban areas based on
data in Table J-19 to estimate the total dose to escorts. For example, truck shipments could be escorted
through an average of five urban areas on average for 30 minutes in each. Using these assumptions and
the estimated dose rate in an escort vehicle, the estimated dose for escorts in separate vehicles is 0.28
millirem per shipment (0.28 millirem = 5 areas per shipment x 0.5 hour per area x 0.1 1 millirem per
hour). For the 24 years of the Proposed Action, the total dose to escorts in separate vehicles would,
therefore, be about 14 rem (0.28 millirem per shipment x 50,000 shipments). This dose would lead to
0.02 latent cancer fatality in the population of escorts who would be affected.
J-46
Transportation
For rail shipments, the analysis assumed that escorts would be 30 meters (98 feet) away from the end of
the shippmg cask on the nearest railcar. This separation distance is the sum of the:
• Length of a buffer car [about 1 5 meters (49 feet)] between a cask car and an escort car required by
Department of Transportation regulations (49 CFR 174.89),
• Normal separation between cars [a total of about 2 meters (6.6 feet) for two separations],
• Distance from the end of a cask to the end of its rail car [about 5 meters (16 feet)], and
• Assumed average distance from the escort car's near-end to its occupants [5 to 10 meters (16 to
32 feet)].
This analysis assumed that the dose rate at 2 meters (6.6 feet) from the end of the cask car would be 10
millirem per hour, the maximum allowed by Department of Transportation regulations (49 CFR 173.441).
The analysis used these assumptions and the RISKfND computer program to estimate 0.46 millirem per
hour as the dose rate in the occupied areas of the escort railcar. For example, an individual escort who
occupied the escort car continuously for a 5-day cross-country trip would receive a maximum dose of
about 55 millirem. Escorting 26 shipments in a year, this individual would receive a maximum dose of
1.4 rem. Over the 24 years of the Proposed Action, if the same individual escorted 26 shipments every
year, he or she would receive a dose of about 34 rem. Using the dose-to-risk conversion factors
recommended by the International Commission on Radiation Protection (ICRP 1991, page 22), this dose
would increase the potential for the individual to contract a fatal cancer from about 22 percent (ACS
1998, page 10) to 24 percent.
J. 1.3.2.3 Vehicle Emission Impacts
Human health impacts from exposures to vehicle exhaust depend principally on the distance traveled in
an urban population zone and on the impact factors for particulates and sulftir dioxide from truck
(including escort vehicles) or rail emissions, fugitive dust generation, and tire abrasion (DOE 1995,
page 1-52).
The analysis estimated incident-free impacts from nonradiological causes using unit risk factors that
account for both fatalities associated with the emissions of pollution in urban, suburban, and rural areas
by transportation vehicles, including escort vehicles. Because the impacts would occur equally for trucks
transporting loaded or unloaded shipping casks, the analysis used round-trip distances. Escort vehicle
impacts were included only for loaded shipment miles.
The analysis used impact factors for effects on urban areas of 0.00000016 fatality per urban mile traveled
(0.0000001 fatality per kilometer) by trucks and 0.00000021 fatality per urban mile traveled (0.00000013
fatality per kilometer) by trains (Rao, Wilmot, and Luna 1982, all). The region of influence used in the
analysis for exposure to vehicle emissions was a band between 30 and 805 meters (98 and 2,640 feet)
wide on both sides of the transportation route.
In addition to unit risk factors used to estimate impacts from vehicle emissions in urban areas, an
additional factor was used to estimate health effects from vehicle exhaust emissions in rural areas. Based
on data in a study by the Environmental Protection Agency that addressed latent cancer consequences of
vehicle exhausts, a factor of 0.000000000072 fatality per kilometer traveled was calculated for use in
rural and suburban population zones (DOE 1995, page 1-52).
Although the analysis estimated human health and safety impacts of transporting spent nuclear fuel and
high-level radioactive waste, exhaust and other pollutants emitted by transport vehicles into the air would
J-47
Transportation
not measurably affect national air quality. National transportation of spent nuclear fuel and high-level
radioactive waste, which would use existing highways and railroads would average 14.2 million truck
kilometers per year for the mostly truck case and 3.5 million railcar kilometers per year from the mostly
rail case. The national yearly average for total highway and railroad traffic is 186 billion truck kilometers
and 49 billion railcar kilometers (BTS 1999, Table 3-22). Spent nuclear fuel and high-level radioactive
waste transportation would represent a very small fraction of the total national highway and railroad
traffic (0.008 percent of truck kilometers and 0.007 percent of rail car kilometers). In addition, the
contributions to vehicle emissions in the Las Vegas air basin, where all truck shipments (an average of
five per day) would travel under the mostly legal-weight truck scenario, would be small in comparison to
those from other vehicle traffic in the area. The annual average daily traffic on 1-15 0.3 kilometer (0.2
mile) north of the Sahara Avenue interchange is almost 200,000 vehicles (NDOT 1997, page 7), about 20
percent of which are trucks (Cerocke 1998, all). For these reasons, national transportation of spent
nuclear fuel and high-level radioactive waste by truck and rail would not constitute a meaningful source
of air pollution along the nation's highways and railroads.
J.1 .3.2.4 Sensitivity of Dose Rate to Characteristics of Spent Nuclear Fuel
For this analysis, DOE assumed that the dose rate external to all shipments of spent nuclear fuel and high-
level radioactive waste would be the maximum value allowed by regulations (49 CFR 173.441).
However, the dose rate for actual shipments would not be the maximum value of 10 millirem per hour at
2 meters (6.6 feet) from the sides of vehicles. Administrative margins of safety that are established to
compensate for limits of accuracy in instruments and methods used to measure dose rates at the time
shipments are made would result in lower dose rates. In addition, the characteristics of spent nuclear fuel
and high-level radioactive waste that would be loaded into casks would always be within the limit values
allowed by the cask's design and its Nuclear Regulatory Commission certificate of compliance.
For example, DOE used data provided in the GA-4 Legal- Weight Truck Cask Design Report (General
Atomics 1993, pages 5.5-18 and 5.5-19) to estimate dose rates 2 meters (6.6 feet) from transport vehicles
for various characteristics of spent nuclear fuel payloads. Figure J-7 shows ranges of bumup and cooling
times for spent nuclear fuel payloads for the GA-4 cask. The figure indicates the characteristics of a
typical pressurized-water reactor spent nuclear fuel assembly (see Appendix A). Based on the design data
for the GA-4 cask, a shipment of typical pressurized-water reactor spent nuclear fuel would result in a
dose rate of about 6 millirem per hour at 2 meters from the side of the transport vehicle, or about 60
percent of the limit established by Department of Transportation regulations (49 CFR 173.441).
Therefore, DOE estimates that, on average, dose rates at locations 2 meters (6.6 feet) from the sides of
transport vehicles would be about 50 to 70 percent of the regulatory limits. As a result, DOE expects
radiological risks to workers and the public from incident-free transportation to be no more than 50 to 70
percent of the values presented in this EIS.
J.1 .4 METHODS AND APPROACH TO ANALYSIS OF ACCIDENT SCENARIOS
J. 1.4.1 Accidents in Loading Operations
J.1 .4.1 .1 Radiological Impacts of Loading Accidents
The analysis used information in existing reports to consider the potential for radiological impacts from
accidents during spent nuclear fuel loading operations at the commercial and DOE sites. These included
a report that evaluated health and safety impacts of multipurpose canister systems (TRW 1 994, all) and
two safety analysis reports for onsite dry storage of commercial spent nuclear fuel at independent spent
fuel storage installations (PGE 1996, all; CP&L 1989, all). The latter reports address the handling and
loading of spent nuclear fuel assemblies in large casks similar to large transportation casks. In addition,
J-48
Transportation
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from side of transport vefiicle
(mllllrem per fiour)
A Typical pressurized-water
reactor spent nuclear fuel
a. GIgawatt days per metric ton of heavy metal.
Figure J-7. Comparison of GA-4 cask dose rate and spent nuclear fiiel bumup and cooling time.
\)m.
J-49
Transportation
DOE environmental impact statements on the management of spent nuclear fuel and high-level
radioactive waste (DOE 1995, all; DOE 1997b, all) provided information on radiological impacts from
loading accidents.
TRW (1994, Sections 3.2 and 4.2) discusses potential accident scenario impacts of four cask management
systems at electric utility and other spent nuclear fuel storage sites. This report concentrated on
unplanned contact (bumping) during lift-handling of casks, canisters, or fuel assemblies. The two safety
analysis reports for independent spent fuel storage installations for commercial spent nuclear fuel (PGE
1996, all; CP&L 1989, all) evaluated a comprehensive spectrum of accident-initiating events. These
events included fires, chemical explosions, seismic events, nuclear criticality, tornado strikes and tornado-
generated missile impacts, lightning strikes, volcanism, canister and basket drop, loaded shipping cask
drop, and interference (bumping, binding) between the transfer cask and storage module. The DOE
environmental impact statements for the interim management of spent nuclear fuel and high-level
radioactive waste (DOE 1995, Appendix E; DOE 1997b, Appendixes F and G) included radiological
impacts from potential accident scenarios associated with preparing, storing, and shipping these materials.
These EISs do not discuss quantitative radiological impacts for accident scenarios associated with
material loading, but do contain estimates of radiological impacts from accident scenarios for the spent
nuclear fuel and high-level radioactive waste management activities considered. As discussed for routine
loading operations, this analysis converted radiation doses to estimates of radiological impacts using
dose-to-risk conversion factors of the International Commission on Radiological Protection.
J.1 .4.1 .2 Industrial Safety Impacts of Loading Operations at Commercial Facilities
The principal industrial safety impact parameters of importance to commercial industry and the Federal
Government are (1) total recordable (injury and illness) cases, (2) lost workday cases associated with
workplace injuries and illnesses, and (3) workplace fatalities. The frequency of these impacts under the
Proposed Action and the inventory modules (Modules 1 and 2) was projected using the involved worker
level of effort, expressed as the number of full-time equivalent worker multiples, that would be needed to
conduct shipment tasks. The workplace loss incidence rate for each impact parameter [as shown in the
DOE Computerized Accident/Incident Reporting and Recordkeeping System (CAIRS) data base (DOE
1999, all)] was used as a multiplier to convert the level of effort to expected industrial safety losses.
DOE did not explicitly analyze impacts to noninvolved workers in its earlier reports (Schneider et al. 1987,
all; Smith, Daling, and Faletti 1992, all). However, for purposes of analysis in this EIS, DOE estimated
that impacts to noninvolved workers would be 25 percent of the impacts to the involved workforce. This
assumption is based on (1) the DOE estimate that about one of five workers assigned to a specific task
would perform administrative or managerial duties, and (2) the fact that noninvolved worker loss incidence
rates are generally less than those for involved workers (see Appendix F, Table F-2).
The estimated involved worker full-time equivalent multiples for each shipment scenario were estimated
using the following formula:
Involved worker full-time equivalent multiples = (AxBxCxD)^E
where: A = number of shipments (from Tables J-5 and J-6)
B = average loading duration for each shipment by fuel type and conveyance mode (workdays;
from Table J-15)
C = workday conversion factor = 8 hours per workday
J-50
Transportation
D = involved worker crew size (13 workers; from Table J-16)
E = full-time equivalent conversion factor = 2,000 worker hours per full-time equivalent
The representative CAIRS data base loss incidence rate for each total recordable case, lost workday case,
and fatality trauma category (for example, the number of total recordable cases per full-time equivalent)
was then multiplied by the involved worker full-time equivalent multiples to project the associated
incidence. The involved worker total recordable case incidence rate used was that reported in the DOE
CAIRS data base (DOE 1999, all) for the 1992 to 1997 period of record because neither the Nuclear
Regulatory Commission nor the Bureau of Labor Statistics maintains data on commercial power reactor
industrial safety losses. The total recordable case incidence rate, 410 cases in a workforce of
15,000 workers (0.03 total recordable case per full-time equivalent), is the averaged loss experience at the
three principal DOE sites: the Savannah River Site, Hanford Site, and Idaho National Environmental and
Engineering Laboratory. The DOE sites were chosen because the operations and hazards would be
representative of those encountered at commercial power reactor sites. Because lost workday cases are
linked to the total recordable case experience (that is, each lost workday case would have to be included
in the total recordable case category), the same DOE CAIRS data base period of record and facilities were
used in the selection of the involved worker lost workday case incidence rate [200 lost workday cases in a
workforce of 15,000 workers (0.013 lost workday case per full-time equivalent)].
The TRW (1994, all) study concluded that radiological impacts from handling incidents would be small.
The total person-rem exposure for accidents in handling the four cask systems considered in the study
would vary from 0. 1 rem to 0.04 rem. This exposure would be the total for all persons who would be
exposed, onsite workers as well as the public. The highest estimated exposure (0.1 person-rem) would
result in 0.(XXX)5 latent cancer fatality in the exposed population.
The involved worker fatality incidence rate used was that also reported in the DOE CAIRS data base, but
for the 1996 to 1997 (through the third quarter) period of record. The average DOE and contractor
fatality rates used (2.9 fatalities among 100,0(X) workers) represent losses among workers operating
equipment and handling waste materials at the principal DOE sites. This fatality incidence rate represents
government and contractor experience in the DOE complex and operations that are governed by safety
and administrative controls that would be similar to those used at commercial power reactor sites.
For comparison, the noninvolved worker total recordable case, lost workday case, and fatality incidence
rates using the same data base sources are 0.033, 0.016, and 0.000029, respectively. However, because
the CAIRS data base did not include fatality rates for noninvolved workers, the involved worker rate was
used.
J.I .4.1 .3 Industrial Safety Impacts of DOE Loading Operations
The technical approach and loss multipliers discussed in Section J. 1.4. 1.2 for commercial power reactor
sites analysis were used for the analysis of spent nuclear fuel and high-level radioactive waste loading
impacts at DOE sites. Because no information existed on the high-level radioactive waste loading
duration for the truck and rail transportation modes, DOE assumed that the number of full-time equivalent
involved workers for the two transportation modes would be the same as that for the DOE sites shipping
spent nuclear fuel. For those sites, the average number of full-time equivalent workers would be about
0.07 and 0.12 per shipment for the truck and rail transportation modes, respectively.
J-51
Transportation
J.1.4.2 Transportation Accident Scenarios
J.1 .4.2.1 Radiological Impacts of Transportation Accidents
A potential consequence and risk of transportation would be accidents that released and dispersed
radioactive material from safe containment in transportation packages. Such releases and dispersals, if
they occurred, would lead to impacts to human health and the environment. The following sections
describe the methods for analyzing the risks and consequences of accidents that could occur in the course
of transporting spent nuclear fuel and high-level radioactive waste to a nuclear waste repository at the
Yucca Mountain site. They discuss the bases for, and methods for, determining rates at which accidents
are assumed to occur, the severity of these accidents, and the amounts of materials that could be released.
Accident rates, severities, and the corresponding quantities of radioactive materials that could be released
are essential data used in the analyses. Appendix A presents the quantities of radioactive materials in a
typical pressurized-water reactor spent nuclear fuel assembly used in the analysis of accident
consequences and risks. Legal-weight truck casks would contain as many as four pressurized-water
reactor spent nuclear fuel assemblies, and rail casks would contain as many as 36 (see Table J-3).
In addition to accident rates and severities, an important variable in assessing impacts from transportation
accident scenarios is the type of material that would be shipped. Accordingly, this appendix presents
information used in the analyses of impacts of accidents that could occur in the course of transporting
conunercial pressurized- and boiling-water reactor fuels, DOE spent nuclear fuels, and DOE high-level
radioactive waste.
POTENTIAL EFFECTS OF HUMAN ERROR ON ACCIDENT IMPACTS
The accident scenarios described in this chapter would be mostly a direct consequence of error on
the part of transport vehicle operators, operators of other vehicles, or persons who maintain vehicles
and rights-of-way. The number and severity of the accidents would be minimized through the use of
trained and qualified personnel.
Others have argued that other kinds of human error could also contribute to accident consequences:
(1) undetected error in the design and certification of transportation packaging (cask) used to ship
radioactive material, (2) hidden or undetected defects in the manufacture of these packages, and (3)
error in preparing the packages for shipment. DOE has concluded that regulations and regulatory
practices of the Nuclear Regulatory Commission and the Department of Transportation address the
design, manufacture, and use of transportation packaging and are effective in preventing these kinds
of human error by requiring:
• Independent Nuclear Regulatory Commission review of designs to ensure compliance with
requirements (10 CFR Part 71)
• Nuclear Regulatory Commission-approved and audited quality assurance programs for design,
manufacturing, and use of transportation packages
In addition, Federal provisions (10 CFR Part 21) provide additional assurance of timely and effective
actions to identify and initiate corrective actions for undetected design or manufacturing defects.
Furthermore, conservatism in the approach to safety incorporated in the regulatory requirements and
practices provides confidence that design or manufacturing defects that might remain undetected or
operational deficiencies would not lead to a meaningful reduction in the performance of a package
under normal or accident conditions of transportation.
J-52
Transportation
For exposures to ionizing radiation following accidents, risks were analyzed in terms of dose and latent
cancer fatalities to the public and workers. The analyses of risk also addressed the potential for fatalities
that would be the direct result of mechanical forces and other nonradiological effects that occur in
everyday vehicle and industrial accidents.
The transportation of spent nuclear fuel and high-level radioactive waste from the 77 sites to the Yucca
Mountain site would be conducted in a manner that complied fully with regulations of the U.S.
Department of Transportation and Nuclear Regulatory Commission. These regulations specify
requirements that promote safety and security in transportation. The requirements apply to carrier
operations; in-transit security; vehicles; shipment preparations; documentation; emergency response;
quality assurance; and the design, certification, manufacture, inspection, use, and maintenance of
packages (casks) that would contain the spent nuclear fuel and high-level radioactive waste.
Because of the high level of performance required by regulations for transportation casks (49 CFR
Part 173 and 10 CFR Part 71), the Nuclear Regulatory Commission estimates that in 99.4 percent of rail
and truck accidents no cask contents would be released (Fischer et al. 1987, page 9-10). The 0.6 percent
of accidents that could cause a release of radioactive materials from casks can be described by a spectrum
of accident severity. As the severity of an accident increases, the fraction of radioactive material contents
that would be released from transportation casks also increases. However, as the severity of an accident
increases it is less likely to occur, hi its Modal Study (Fischer et al. 1987, all), the Nuclear Regulatory
Commission developed an accident analysis methodology that uses this concept of a spectrum of severe
accidents to calculate the probabilities and consequences of unlikely accidents that could occur in
transporting highly radioactive materials.
Although the Nuclear Regulatory Commission approach, which was used in this EIS, provides a method
for determining the frequency with which severe accidents can be expected to occur, their severity, and
their consequences, a method does not exist for predicting where along routes accidents would occur.
Therefore, for the analyses of impacts presented here the method used in the RADTRAN4 computer code
(Neuhauser and Kanipe 1992, all) is used. This method assumes that accidents could occur at any
location along routes, with their frequency of occurrence being determined by the accident rate
characteristic of the states through which the route passes and the number of shipments that travel the
route.
The transportation accident scenario analysis evaluated radiological impacts to populations and to
hypothetical maximally exposed individuals and estimated fatalities that could occur from traffic
accidents. It included both rail and legal-weight truck transportation. The analysis used the RADTRAN4
(Neuhauser and Kanipe 1992, all) and RISKIND (Yuan et al. 1995, all) computer programs to determine
accident consequences and risks. DOE has used both codes in recent DOE environmental impact
statements (DOE 1995, Appendix J; DOE 1996a, Appendix E; DOE 1997b, Appendixes F and G) that
address impacts of transporting radioactive materials. The analyses used seven kinds of information to
determine the consequences and risks of accidents for populations:
• Routes from the 77 sites to the repository and their lengths in each state and population zone
• The number of shipments that would be transported over each route
• State-specific accident rates
• The kind and amount of radioactive material that would be transported in shipments
• Probabilities of release and fractions of cask contents that could be released in accidents
J-53
Transportation
ESTIMATING ACCIDENT RISK
Assessing the radiological impact of accidents involves estimating the probability that an accident
might occur and estimating the accident consequences. The probability, or chance, that an accident
will occur is multiplied by the consequences of the accident to determine accident risk.
One method for estimating accident probabilities uses historic information on the rate at which
accidents of a similar type or severity occur (accidents per vehicle-mile traveled). Information of this
type is maintained as transportation accident data by the Department of Transportation and by
transportation safety organizations in state governments. Accident rates are multiplied by the total
number of miles that vehicles would travel to estimate the number of accidents.
Determining radiological accident consequences requires estimating the quantity of radionuclides
likely to be released and the environmental transport mechanisms that would bring the radionuclides
into contact with people and then calculating the resultant radiation dose. Because of the large
amounts of data these calculations require, conservative or bounding assumptions are commonly
used to simplify the calculation task. As a result, calculated risks tend to be overestimates.
• The number of people who could be exposed to accidents and how far they lived from the routes
• Exposure scenarios that include multiple exposure pathways, state-specific agricultural factors, and
atmospheric dispersion factors for neutral and stable conditions applicable to the entire country for
calculating radiological impacts
The analysis used the same routes and lengths of travel as the analysis of incident-free transportation
impacts discussed above.
DOE used the CALVIN computer code discussed earlier, the DOE Throughput Study (TRW 1997, all),
and information provided by the DOE National Spent Nuclear Fuel Program (Jensen 1998, all) to
calculate the number of shipments from each site and, thus, the number of shipments that would use a
particular route.
The state-specific accident rates (accidents and fatalities per kilometer of vehicle travel) used in the
analysis included accident statistics for commercial motor carrier operations for the Interstate Highway
System, other U.S. highways, and state highways for each of the 48 contiguous states (Saricks and
Tompkins 1999, all). The analysis also used average accident and fatality rates for railroads in each state.
The data specifically reflect accident and fatality rates that apply to commercial motor carriers and
railroads.
Appendix A contains information on the radioactive material contents of shipments. Appendix A,
Section A.2.1.5 describes the characteristics of the spent nuclear fuel and high-level radioactive waste that
would be shipped. The analysis assumed that the average inventory of radioactive materials in shipments
would be typical pressurized-water reactor spent nuclear fuel that had been removed from reactors for
25.8 years. Appendix A describes this inventory. The estimated impacts would be less if the analysis
used the characteristics of a typical boiling-water reactor spent nuclear fuel, DOE spent nuclear fuel
(including naval spent nuclear fuel, which the analysis assumed would be removed from reactors 5 years
before its shipment to the repository), or high-level radioactive waste.
J-54
Transportation
The analysis also used the number of people who potentially would be close enough to transportation
routes at the time of an accident to be exposed to radiation or radioactive material released from casks,
and the distances these people would be from the accidents. It used the HIGHWAY and INTERLINE
computer programs to determine this estimated number of people and their distances from accidents.
HIGHWAY and INTERLINE used 1990 Census data for this analysis. The analysis assumed that the
region of influence extended 80 kilometers (50 miles) from an accident.
Accident Severity Categories and Conditional Probabilities
The classification scheme used in the Modal Study for both truck and rail transportation accidents is
shown in Figure J-8. As shown, accident severity is a function of two variables. The first variable is the
mechanical force that occurs in impacts. In the figure, mechanical force is represented by the deformation
(strain) in a cask's containment (inner shell) that the force would cause. The second variable is thermal
energy, or the heat input to a cask engulfed by fire. In the figure, thermal energy is represented by the
midpoint temperature of a cask's lead shield wall following heating, as in a fire.
Because all accident scenarios that would involve casks can be described in these terms, the severity of
accidents can be analyzed independently of specific accident sequences. In other words, any sequence of
events that results in an accident in which a cask is subjected to mechanical forces, within a certain range
of values, and possibly fire is assigned to the accident severity category associated with the applicable
ranges for the two parameters. This accident severity scheme enables analysis of a manageable number
of accident situations while accounting for all reasonably foreseeable transportation accidents, including
accidents with low probabilities but high consequences and those with high probabilities but low
consequences.
For the analysis of impacts, a conditional probability was assigned to each accident severity category.
Figure J-8 also shows the conditional probabilities developed in the Modal Study for the accident severity
matrix. These conditional probabilities are used in the analysis of impacts presented in this chapter. The
conditional probabilities are the chances that accidents will involve the mechanical forces and the heat
energy in the ranges that apply to the categories. For example, accidents that would fall into the category
labeled R(l,l), which represents the least severe accident in the matrix, would be likely to make up 99.4
percent of all accidents that would involve truck and railcar shipments of casks carrying spent nuclear fuel
or high-level radioactive waste. The mechanical forces and heat in accidents in this category would not
exceed the regulatory design standards for casks. Using the information in the figure, an accident in this
category could cause a maximum of 0.2 percent strain (deformation) in a cask's containment and could
heat the lead shielding to 260°C (500°F) degrees. These damage conditions are within the range of
damage that would occur to casks subjected to the hypothetical accident conditions tests that Nuclear
Regulatory Commission regulations require a cask to survive (10 CFR Part 71). Category R(4,5)-
accidents, which would cause extensive damage to a cask, are very severe but very infrequent. The
Category R(4,5) accidents would occur an estimated 3.4 times in each 100 trillion rail accidents and less
than one time in each 10 quadrillion truck accidents.
The analysis of accident risks presented in this appendix used the frequency that would be likely for
accidents in each of the severity categories. This frequency was determined by multiplying the category's
conditional probability by the accident rates for each state's urban, suburban, and rural population zones
and by the shipment distances in each of these zones, and then adding the results. The accident rates in
the population density zones in each state are distinct and correspond to traffic conditions, including
average vehicle speed, traffic density, and other factors, including rural, suburban, or urban location.
In terms of potential to release radioactivity to the environment, the most severe of reasonably foreseeable
accidents are those that would fall into one of the eight categories of very severe accidents. For these
eight categories, the fractions and characteristics of radioactive materials that would be released in an
J-55
Transportation
Ee-
ls
CO O)
|1
I"
II
Jo"
Legend
R(x,y) =
P.=
P,=
S3
P,
Pr
R(4,1)
1.532x10-'
1.786x10-'
R(4.2)
3.926x10-'"
3.290 x10'3
R(4,3)
1.495x10-'"
2.137 x10-'3
R(4,4)
7.681 X 10"
1.644 x10'3
R(4,5)
<1 xlO"
3.459x10'"
(30)
s?
P,
P,
R(3,1)
1.7984x10-'
5.545x10-"
R(3,2)
1.574x10-'
1.021 xlO-'
R(3,3)
2.034x10-'
6.634x10'
R(3,4)
1.076x10'
5.162x10-'
R(3,5)
4.873x10-'
5.296x10'
(2)
s,
P,
P,
R(2,1)
3.8192x10-3
2.7204x10-3
R(2,2)
2.330x10-'
5.011 xlO-'
R(2,3)
3.008x10'
3.255x10'
R(2,4)
1.592x10'
2.531 x 10'
R(2,5)
7.201 xlO'
1.075x10'
(0.2)
P,
P,
R(1,1)
0.994316
0.993962
R(1,2)
1.687x10-5
1.2275x10-3
R(1,3)
2.362x10-=
7.9511x10-"
R(1,4)
1.525x10-=
6.140x10-"
R(1,5)
9.570x10'
1.249x10"
T,
(500)
T2
(600)
T3
(650)
T4
(1 ,050)
Thermal response (lead mid-thickness temperature, °F)
The label used to identify the cell in the accident response matrix located at the
X row frotn the bottotn of the matrix and y column from the left of the matrix. Thus,
(R1 ,1) is the identifier for the cell in the lower left corner of the matrix.
Probability of occurrence assuming a truck accident occurs.
Probability of occurrence assuming a rail accident occurs.
Note:
- Maximum strain between 0 and 0.2 percent (Si) for the inner shell of a cask would be
within the design conditions for a Nuclear Regulatory Commission-certified shipping
cask. There would be permanent deformation after the load is removed. Si strains
could occur in impacts against medium hardness structures (for example, bridge
abutments) at speeds up to 100 kilometers (60 miles per hour).
- Strains between 0.2 and 2 percent (S2) would result in small permanent deformations.
52 strains could occur in impacts against medium hardness structures at speeds up to
130 kilometers (80 miles per hour).
- Strains between 2 and 30 percent (S3) would result in large permanent deformations.
53 strains could occur in impacts against medium hardness structures at speeds
greater than 1 30 kilometers (80 miles per hour).
Source: Ftscher et al. (1987, pages 4-8, 7-25, and 7-26),
Figure J-8. Probability matrix for mechanical forces and heat in transportation accidents.
J-56
Transportation
accident were estimated to be the same. That is, for a shipment of spent nuclear fuel that is involved in an
accident classified as Category R(4,l), the amount and characteristics of radioactive material assumed to
be released would be the same as those for an accident that would fall into Category R(4,2), R(4,3),
R(4,4), R(4,5), R(l,5), R(2,5), or R(3,5). Because the releases of radioactive materials that could occur
are assumed to be the same for each of these eight categories, the probabilities of occurrence can be
summed. This sum is used to calculate a collective probability for the most severe of the accidents
addressed in this analysis. Thus, the conditional probability of a truck accident of the greatest severity
that is analyzed would be 0.0000098 per accident event (about 1 chance in 100,000 per accident).
By combining categories for which the releases of radioactive materials are assumed to be equivalent, the
20 accident categories in Figure J-8 are reduced to six collective categories. The first is the same as
severity category R(1,I); the second collects severity categories R(l,2) and R(l,3); the third R(2,l),
R(2,2) and R(2,3); the fourth R(3,l), R(3,2) and R(3,3); the fifth, R(l,4), R(2,4), and R(3,4); and, as
discussed above, the sixth collects R(4,l) through R(4,5) and R(l,5) through R(3,5).
Accident Releases
Radiological consequences were calculated by assigning cask release fractions to each accident severity
category for each chemically and physically distinct radioisotope. The release fraction is defined as the
fraction of the radioactivity in the cask that could be released from the cask in a given severity of
accident. Release fractions vary according to spent nuclear fuel type and the physical/chemical properties
of the radioisotopes. Most radionuclides in spent nuclear fuel are in chemically and physically stable,
solid, nondispersible forms. Gaseous radionuclides, such as krypton-85, would be released if both the
fuel cladding and cask containment boundary were compromised.
The Modal Study developed release fractions for commercial spent nuclear fuel from pressurized-water
reactors. These release fractions, listed in Table J-21, are based on best engineering judgment and are
believed to be conservative. The analysis estimated the amount of radioactive material released from a
cask in an accident by multiplying the approximate release fraction by the number of fuel assemblies in a
cask (see Table J-3) and the radionuclide activity of a spent nuclear fuel assembly (see Appendix A). To
provide perspective, the release fraction for a category 6 accident involving a large rail cask results in an
estimated release of about 1,600 curies of cesium isotopes. For this analysis, the release fractions
developed by the Modal Study were used only for commercial pressurized-water reactor fuel and spent
nuclear fuel from training, research and isotope reactors built by General Atomics (commonly called
TRIGA spent nuclear fuel), both of which are rod-type fuels. The availability of fuel-specific data for
other types of spent nuclear fuel that would be shipped to the repository allowed the use of release
fractions that more closely approximate expected release characteristics.
Table J-21. Fractions of selected radionuclides in commercial spent nuclear fuel projected to be released
from casks in transportation accidents for cask response regions.
Severity
Release fraction*
lodine-
Cesium-134, -
Ruthenium
ft Cask response region
category
Inert gas
129
135,-137
-106
Particulates
1 R(l,l)
1
0.0
0.0
0.0
0.0
0.0
P R(1,2),R(1,3)
2
9.9x10-^
7.5x10"'
6.0x10"'
8.1x10"'
6.0x10"*
R(2,1),R(2,2),R(2,3)
3
3.3x10"^
2.5x10"*
2.0x10"'
2.7x10"'
2.0x10"'
R(3,1),R(3,2),R(3,3)
4
3.3x10"'
2.5x10"^
2.0x10""
2.7x10"'
2.0x10"'
R(1,4),R(2,4),R(3,4)
5
3.9x10"'
4.3x10^
2.0x10""
4.8x10"'
2.0x10"'
R(1,5),R(2,5),R(3,5),R(4,5),
6
6.3x10'
4.3x10"^
2.0x10"'
4.8x10""
2.0x10"'
R(4,l),R(4,2),R(4,3)Jl(4,4)
Source: (DOE 1995, page 1-86).
J-57
Transportation
Release fractions for aluminum fuels (aluminum alloy fuel, aluminum cladding) were based on laboratory
measurements and the U.S. Nuclear Regulatory Commission Modal Study (Fischer et al. 1987, all).
Because of the lower melting point of aluminum compared to metals used in other metallic fuels, the
aluminum fuel release fractions are considered bounding for metallic fuels (that is. Savannah River
Production Reactor, Hanford N-Reactor, and Experimental Breeder Reactor-II Mark V spent nuclear
fuel). Release fractions for the aluminum and other metallic fuel types are listed in Table J-22. The
estimates of fractions for cask contents released in severe accidents were assumed to be independent of
the type of cask.
Table J-22. Fractions of selected radionuclides in aluminum and metallic spent nuclear fuel projected to
be released from casks in transportation accidents for cask response regions.^
Severity
Release fraction''
lodine-
Cesium- 134,
Ruthenium-
Cask response region
category
Inert gas
129
-135.-137
106
Particulates
R(l,l)
1
0.0
0.0
0.0
0.0
0.0
R(1,2),R(1,3)
2
9.9 X 10"^
1.1 X 10'^
3.0 X 10"*
4.1 X 10"'
3.0 X 10'"
R(2,1),R(2,2),R(2,3)
3
3.3 X 10-^
3.5 X 10-^
1.0x10"^
1.4x10"*
1.0x10"'
R(3,l)Jl(3,2),R(3,3)
4
3.3 X 10'
3.5 X 10'
1.0x10"*
1.4x10"^
1.0x10"*
R(1,4),R(2,4),R(3,4)
5
3.9 X 10'
6.0 X 10"*
1.0x10"*
2.4 x 10"^
1.0x10"*
R(1,5),R(2,5),R(3,5),R(4,5),
6
6.3 X 10"'
6.0 X 10"^
1.0x10"'
2.4 x 10"*
1.0x10"'
R(4,1),R(4,2), R(4,3),R(4,4)
a. Source: DOE (1995, page 1-87).
b. These release fractions are applicable to N-Reactor, Savannah River Site production reactor, and DOE research/test reactor
spent nuclear fuel tyjjes.
Atmospheric Conditions
For the analyses of accident risk and consequences, releases of radioactive materials from casks during
and following severe accidents were assumed to be into the atmosphere where these materials would be
carried by wind. Because it is not possible to predict specific locations where transportation accidents
would occur, atmospheric conditions that generally apply throughout the continental United States were
used.
Table J-23 lists the frequency at which atmospheric stability and wind speed conditions occur in the
contiguous United States. The data, which are averages for 177 meteorological data collection locations,
were used in conjunction with the RISKE^ID computer program (Yuan et al. 1995, all) to develop
estimates of the consequences of maximum reasonably foreseeable accidents and acts of sabotage.
In calculating estimated values for consequences, RISKIND used the atmospheric stability and wind
speed data to analyze the dispersion of radioactive materials in the atmosphere that could follow releases
in severe accidents. The dispersions were modeled as plumes of gases and particles. Using the results of
the dispersion analysis, RISKIND calculated values for radiological consequences (population dose and
dose to a maximally exposed individual). These results were placed in order from lowest to highest.
Following this order, the probabilities of the atmospheric conditions associated with each set of
consequences were accumulated. As the accumulated probability increased and the likelihood of an
exceedance of a set of atmospheric conditions decreased, estimated consequences increased. This
procedure was followed to identify the level of severe accident and sabotage consequences that would not
be exceeded 50 percent and 95 percent of the time. For atmospheric conditions that are called neutral, or
average, the consequences would not be exceeded 50 percent of the time. Thus, neutral atmospheric
conditions would be the conditions likely to prevail during a severe accident or act of sabotage. Under
stable, or quiescent, conditions the consequences would not be exceeded 95 percent of the time. The
J-58
Transportation
Table J-23. Frequency
of atmospheric and wind speed conditions - U.S. averages.
a
Atmospheric
Wind speed
condition
stability class
WS(1)
WS(2)
WS(3)
WS(4)
WS(5)
WS(6)
Total
A
0.00667
0.00444
0.00000
0.00000
0.00000
0.00000
0.01111
B
0.02655
0.02550
0.01559
0.00000
0.00000
0.00000
0.06764
C
0.01400
0.02931
0.05724
0.01146
0.00122
0.00028
0.11351
D
0.03329
0.07231
0.15108
0.16790
0.03686
0.01086
0.47230
E
0.00040
0.04989
0.06899
0.00146
0.00016
0.00003
0.12093
F
0.10771
0.08710
0.00110
0.00000
0.00000
0.00000
0.19591
G
0.01713
0.00146
0.00000
0.00000
0.00000
0.00000
0.01859
F+G
0.12485
0.08856
0.00110
0.00000
0.00000
0.00000
0.21451
Totals
0.20576
0.27000
0.29401
0.18082
0.03825
0.01117
1.00000
Wind speed (meters
per
0.89
2.46
4.47
6.93
9.61
12.52
second)''
a. Source: TRW (1999a, page 40).
b. To convert meters per second to miles per hour, multiply by 2.237.
analysis assumed that these conditions, which would be unlikely, would occur only for maximum
reasonably foreseeable accidents that had an annual probability greater than 2 chances in 1 million in a
year.
Exposure Pathways
Radiation doses were calculated for an individual who is postulated to be near the scene of an accident
and for populations within 80 kilometers (50 miles) of an accident location. Doses were determined for
rural, suburban, and urban population groups. Dose calculations considered a variety of exposure
pathways, including inhalation and direct exposure (cloudshine and immersion in a plume of radioactive
material) from a passing cloud of contaminants; ingestion from contaminated crops; direct exposure from
radioactivity deposited on the ground (groundshine); and inhalation of radioactive particles resuspended
by wind from the ground.
Emergency Response, Interdiction, Dose !\/litigation, and Evacuation
The RADTRAN4 computer program that DOE used to estimate radiological risks includes assumptions
about the postaccident remediation of radioactive material contamination of land where people live. The
program assumed that, after an accident, contaminants would continue to contribute to population dose
through three pathways — groundshine, inhalation of resuspended particulates, and, for accidents in rural
areas, ingestion of foods produced on the contaminated lands. It also assumed that medical and other
interdiction would not occur to reduce concentrations of radionuclides absorbed or deposited in human
tissues as a result of accidents.
Similarly, the RISKIND (Yuan et al. 1995, all) computer program includes assumptions about response,
interdiction, dose mitigation, and evacuation for calculating radiological consequences (dose to
populations and maximally exposed individuals). In estimating consequences of maximum reasonably
foreseeable accidents during the transportation of spent nuclear fuel and high-level radioactive waste to
the repository, the analysis assumed the following:
• Populations would continue to live on contaminated land for 1 year.
• There would be no radiological dose to populations from ingestion of contaminated food. Food
produced on land contaminated by a maximum reasonably foreseeable accident would be embargoed
from consun^tion.
J-59
Transportation
• Medical and other interdiction would not occur to reduce concentrations of radionuclides absorbed or
deposited in human tissues as a result of an accident.
The analysis of radiological risks to populations and estimates of consequences of maximum reasonably
foreseeable accidents did not explicitly address local, difficult-to-evacuate populations such as those in
prisons, hospitals, nursing homes, or schools. However, the analysis addressed the potential for accidents
to occur in urban areas with high population densities and used the assumptions regarding interdiction,
evacuation, and other intervention actions discussed above. These assumptions encompass the
consequences and risks that could arise from slowness in preventing the consequences of an accident for
some population groups.
Health Risk Conversion Factors
The health risk conversion factors used to estimate expected latent cancer fatalities from radiological
exposures are presented in International Commission on Radiological Protection Publication 60 (ICRP
1991, page 22). These factors are 0.0005 latent cancer fatality per person-rem for members of the public
and 0.0004 latent cancer fatality per person-rem for workers. For accidents in which individuals would
receive doses greater than 20 rem over a short period (high dose/high dose rate), the factors would be
0.0010 latent cancer fatality per rem for a member of the public and 0.0008 latent cancer fatality per rem
for workers.
Assessment of Accident Risk
The RADTRAN4 computer code (Neuhauser and Kanipe 1992, all) was used in calculating risks from
transportation of spent nuclear fuel and high-level radioactive waste. The code determined unit-risk
factors (person-rem per curie) for the radionuclides of concern in the inventory being shipped. The unit-
risk factors from RADTRAN4 were combined with conditional accident probabilities, state-specific
accident rates, release fractions for each of the six accident severity collective categories, and state-
specific food transfer factors to obtain risk per shipment for routes. The accident risks were estimated in
terms of collective radiation dose to the population within 80 kilometers (50 miles).
The analysis first calculated unit risk factors for a shipment for each state through which shipments would
pass. This was done for the three types of population zones in each state (using population density data
from the 1990 census) and for each accident severity category. The unit risk factors used actual
population densities within 8(X) meters (0.5 mile) of routes based on 1990 census data to estimate
populations within 80 kilometers (50 miles). This yielded values for each transportation mode, for each
type of impact, and for each state through which a shipment would pass. The unit risk factors for all the
applicable accident severity categories were summed for each population zone for each state. Also, for
the three types of population zone in a state, the lengths through areas of each type were summed for the
route used in the analysis. This yielded route lengths for each population zone in each state. The sum of
the route lengths and the sum of the unit risk factors for each population zone were multiplied together.
This was repeated for each population zone in each state through which a shipment would pass. The
results were summed to provide estimates of the accident risk for a shipment.
Estimating Consequences of IVIaximum Reasonably Foreseeable Accident Scenarios
In addition to analyzing the radiological and nonradiological risks that would result from the
transportation of spent nuclear fuel and high-level radioactive waste to the repository, DOE assessed the
consequences of maximum reasonably foreseeable accidents. This analysis provided information about
the magnitude of impacts that could result from the most severe accident that could reasonably be
expected to occur, although it could be highly unlikely. DOE concluded that, as a practical matter, events
with a probability less than 1 x 10"^ (1 chance in 10 million) per year rarely need to be examined (DOE
1993, page 28). This would be equivalent to about once in the course of 15 billion legal-weight truck
shipments. For perspective, an accident this severe in commercial truck transportation would occur about
J-60
Transportation
once in 50 years on U.S. highways. Thus, the analysis of maximum reasonably foreseeable accidents
postulated to occur during the transportation of spent nuclear fuel and high-level radioactive waste
evaluated only consequences for accidents with a probability greater than 1 x 10'' per year. The
consequences were determined for atmospheric conditions that could prevail during accidents and for
physical and biological pathways that would lead to exposure of members of the public and workers to
radioactive materials and ionizing radiation. The analysis used the RISKIND code (Yuan et al. 1995, all)
to estimate doses for individuals and populations.
The analysis assumed maximum reasonably foreseeable accident scenarios could occur anywhere, either
in rural or urbanized areas. The probability of such an accident would depend on the amount of exposure
to the transportation accident environment. In this case, exposure would be the product of the cumulative
shipment distance and the applicable accident rates. However, because of large differences in exposure,
principally because of the large differences in the distances traveled in the two types of population areas,
a severe accident scenario that might be reasonably foreseeable, in a rural area might not be reasonably
foreseeable in an urbanized area. Thus, a reasonably foreseeable accident postulated to occur in a rural
area (most travel would occur in rural areas) under meteorological conditions that would be exceeded
(resulting in greater consequences) only 5 percent of the time, might not be reasonably foreseeable in an
urbanized area where shipments would travel relatively few kilometers. For the mostly legal-weight truck
and mostly rail scenarios. Table J-24 lists the probability of a severe accident during national
transportation. These probabilities are for accidents that would:
• Occur in urbanized and rural areas
• Occur under median (50-percent) meteorological conditions and 95-percent conditions (95-percent
conditions would be exceeded, in terms of dose consequences, only 5 percent of the time)
• Occur for accidents in collective severity categories 5 and 6 that are postulated to result in the largest
releases of radioactive materials from shipping casks
• Involve rail and legal-weight truck casks
Table J-24. Annual probability of severe accidents in urbanized and rural areas - category 5 and 6
accidents, national transportation.
Meteorologic
conditions
exceeded
Probability of exceeding
threshold for Category 5
Probability of exceeding
threshold for Category 6
Scenario
Annual
probability for
urbanized area
Annual
probability for
rural area
Annual
probability for
urbanized area
Annual
probability for
rural area
Mostly rail
Truck shipments
50%
4xl0-'<'>
2x10'
3x10"^
1x10"'
95%
IxlO*"'
1x10-^
1x10*
7x10-*
Rail shipments
50%
IxlO'
4x10-'
3x10"'
8x10"'
95%
7x10''
2x10"'
2x10"^
4x10"^
Mostly legal-weight truck
Truck shipments
50%
6x10-*
4x10-'
4x10"'
2x10"'
95%
3x10-'
2x10'
2x10"^
IxlO"*
Rail shipments
50%
4x10*
1x10"'
8x10"'
4x10"^
95%
2x10*
5x10^
4xl0"
2x10*
a. Probabilities not in bold are reasonably foreseeable.
b. Probabilities in bold would occur less than one time in 10 million and therefore are not reasonably foreseeable.
J-61
Transportation
For the mostly legal-weight truck scenario, in which only naval spent nuclear fuel would be shipped by
rail, the likelihood would be less than 1x10"^ per year for the most severe rail accident (severity
category 6) to occur in an urbanized area. Thus, the highest severity rail accidents would only be
reasonably foreseeable in rural areas under average (50-percent) meteorological conditions (probability
greater than 1 in 10 million per year).
Table J-24 also lists the probabilities of other severe accidents the analysis considered. Under the mostly
rail scenario, the most severe types of legal-weight truck accidents (collective category 6) in rural and
urbanized areas under meteorological conditions that would be exceeded only 5 percent of the time would
not be reasonably foreseeable.
In total, 9 sets of accident conditions defined by scenario, shipment mode, meteorology, accident severity
category, and location (identified in the table by shaded cells) would not be reasonably foreseeable.
Nonetheless, although the probabilities would be remote for some accidents, the RADTRAN4 analysis of
radiological dose-risks (discussed above) included risk contributions of all accidents, including ones in
categories 1 through 4, regardless of their probability of occurrence or consequences. Thus, the analysis
addressed the contributions to risk from the spectrum of accidents that would range from low-
consequence, high-probability events to high-consequence, low-probability events.
The analysis of maximum reasonably foreseeable accidents evaluated only accidents from the 23 listed in
Table J-24 that would be reasonably foreseeable and that could result in maximum consequences.
From this collection of 23 possible accidents, the analysis evaluated three sets of accident conditions that
were determined as those with the greatest consequences — one for the mostly rail scenario and two for the
mostly legal-weight truck scenario — to identify the maximum reasonably foreseeable accident that would
have the greatest consequences. The results for these cases are listed in Table J-25. Based on these
results, the analysis identified one maximum reasonably foreseeable accident each for the mostly rail and
mostly legal-weight truck national transportation analysis scenarios. For the mostly legal-weight truck
scenario, the maximum reasonably foreseeable accident would be a severity category 6 accident involving
a legal-weight truck cask in an urbanized area under stable weather (meteorological conditions that would
be exceeded only about 5 percent of the time) conditions. For the mostly rail scenario, the accident would
also be a category 6 accident involving a rail cask in an urbanized area under stable weather conditions.
The analysis of consequences of maximum reasonably foreseeable accidents used data from the 1990
census to estimate the size of populations in urbanized areas that could receive exposures to radioactive
materials. The analysis used estimated populations in successive 8-kilometer (5-mile)-wide annular rings
around the centers of the 21 large urbanized areas (cities and metropolitan areas) in the continental United
States (TRW 1999a, page 22). The average population for each ring was used to form a population
distribution for use in the analysis. To be conservative in estimating consequences, the analysis assumed
that accidents in urbanized areas would occur at the center of the population zone, where the population
density would be greatest. This assumption resulted in conservative estimates of collective dose to
exposed populations.
J.1 .4.2.2 Methods and Approach for Analysis of Nonradiological Impacts of
Transportation Accidents
Nonradiological accident risks are risks of traffic fatalities. Traffic fatality rates are reported by state and
Federal transportation departments as fatalities per highway vehicle- or train-kilometer traveled. The
fatalities are caused by physical trauma in accidents. For nonradiological accident risks estimated in this
EIS for legal-weight truck transportation, accident fatality risks were based on state-level fatality rates for
Interstate Highways (Saricks and Tompkins 1999, all). Accident fatality risks for rail transportation were
J-62
Transportation
Table J- 25. Consequences of maximum reasonably foreseeable accidents in national
transportation.
Severity category 5 accidents
Severity category 6 accidents
Meteorologic
Consequences in
Consequences
Consequences in
Consequences
Scenario
conditions exceeded
urbanized area
in rural area
urbanized area
in rural area
Mostly rail
Truck accident
50%
+"
+
+
+
95%
_b
+
~
~
Rail accident
50% population dose
+
+
+
+
50% MEf dose
+
+
+
+
95% population dose
+
+
61,000(31)"
+
95% MEI dose
+
+
26 (0.013)'
+
Mostly legal-
weight truck
Truck accident
50% population dose
++'
++
++
++
50% MEI dose
++
++
++
++
95% population dose
++
++
9,400(5)
430 (0.2)
95% MEI dose
++
++
4 (0.002)
3.9 (0.002)
Rail accident
50%
++
_
++
95%
~
-
~
-
a. + = Consequences of these accidents are bounded by the rail accident in an urbanized area.
b. = probability less than 1 x 10-7 (not reasonably foreseeable).
c. MEI = maximally exposed individual.
d. Population consequence in person-rem (latent cancer fatality).
e. MEI consequences in rem (probability of increasing a latent cancer fatality).
f ++ = Consequences of these accidents are bounded by the truck accident in an urbanized area.
also calculated using state-specific rates (Saricks and Tompkins 1999, all). Section J.2.1 discusses
methods and data used to analyze accidents for barge transportation.
For truck transportation, the rates in Saricks and Tompkins (1999, Table 4) are specifically for heavy
combination trucks involved in interstate commerce. Heavy combination trucks are multiaxle tractor-
trailer trucks having a tractor and one to three freight trailers connected to each other. This kind of truck
with a single trailer would be used to ship spent nuclear fuel and high-level radioactive waste. Truck
accident rates were determined for each state based on statistics compiled by the Department of
Transportation Office of Motor Carriers for 1994 through 1996. The report presents accident
involvement and fatality counts, estimated kilometers of travel by state, and the corresponding average
accident involvement, fatality, and injury rates for the 3 years investigated. Fatalities include crew
members and all others attributed to accidents. Although escort vehicles would not be heavy combination
trucks, the fatality rate data used for truck shipments of loaded and empty spent fuel casks were also used
to estimate fatalities from accidents that would involve escort vehicles.
Rail accident rates were computed and presented similarly to truck accident rates, but a railcar is the unit
of haulage. The state-specific rail accident involvement and fatality rates are based on statistics compiled
by the Federal Railroad Administration for 1994 through 1996. Rail accident rates include both mainline
accidents and those occurring in railyards (Saricks and Tompkins 1999, page 9).
The accident rates used to estimate traffic fatalities were computed using data for all interstate shipments,
independent of the cargoes. Shippers and carriers of radioactive material generally have a higher-than-
average awareness of transport risk and prepare cargoes and drivers accordingly (Saricks and Kvitek
1994, all). These effects were not given credit in the assessment.
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Transportation
J.1 .4.2.3 Data Used To Estimate Incident Rates for Rail and Motor Carrier Accidents
In analyzing potential impacts of transporting spent nuclear fuel and high-level radioactive waste, DOE
considered both incident-free transportation and transportation accidents. Potential incident-free
transportation impacts would include those caused by exposing the public and workers to low levels of
radiation and other hazards associated with the normal movement of spent nuclear fuel and high-level
radioactive waste by truck, rail, or barge. Impacts from accidents would be those that could result from
exposing the public and workers to radiation, as well as vehicle-related fatalities.
In its analysis of impacts from transportation accidents, DOE relied on data collected by the U.S.
Department of Transportation and others (for example, the American Petroleum Institute) to develop
estimates of accident likelihood and their ranges of severity (see Fischer et al. 1987, pages 7-25 and 7-26).
Using these data, the analysis estimated that as many as 40 accidents could occur over 24 years in the
course of shipping spent nuclear fuel to the repository by legal-weight trucks; 1 or 2 rail accidents that
involved a railcar carrying a cask could occur if most shipments were by rail; and no accidents would be
likely for the limited use of barges.
Furthermore, in using data collected by the Department of Transportation, the analysis considered the
range of accidents, from slightly more than "fender benders" to high-speed crashes, that the DOE carrier
would have to report in accordance with the requirements of Department of Transportation regulations.
The accidents that could occur would be unlikely to be severe enough to affect the integrity of the
shipping casks.
The following paragraphs discuss reporting and definitions for transportation accidents and the
relationships of these to data used in analyzing transportation impacts in this EIS.
J.1 .4.2.3.1 Transportation Accident Reporting and Definitions. In the United States, the
reporting of transportation accidents and incidents involving trucks, railroads, and barges follows
requirements specified in various Federal and state regulations.
Motor Carrier Accident Reporting and Definitions
Regulations generally require the reporting of motor carrier accidents (regardless of the cargo being
carried) if there are injuries, fatalities, or property damage. These regulations have evolved through the
years, mostly in response to increasing values of transportation equipment and commodities. For
example, the Federal requirements in the following text box establish a functional threshold for damage to
vehicles rather than a value-of-damage threshold, which was used until the 1980s. Nonetheless, many
states continue to use value thresholds (for example, Ohio uses $500) for vehicle damage when
documenting reportable accidents.
Until March 4, 1993, Federal regulations (49 CFR Part 394) required motor carriers to submit accident
reports to the Federal Highway Administration Motor Carrier Management Information System using the
so-called "50-T" reporting format. The master file compiled from the data on these reports in the Federal
Highway Administration Office of Motor Carriers was the basis of accident, fatality, and injury rates
developed for the 1994 study of transportation accident rates (Saricks and Kvitek 1994, all).
The Final Rule of February 2, 1993 (58 FR 6726, February 2, 1993), modified the carrier reporting
requirement; rather than submitting reports, carriers now must maintain a register of accidents that meet
the definition of an accident for 1 year after such an accident occurs. Carriers must make the contents of
such a register available to Federal Highway Administration agents investigating specific accidents. They
must also give ". . .all reasonable assistance in the investigation of any accident including providing a full,
true, and correct answer to any question of inquiry" to determine if hazardous materials other than spilled
J-64
Transportation
COMMERCIAL MOTOR VEHICLE ACCIDENT
(49 CFR 390.5)
An occurrence involving a commercial motor vehicle operating on a public road in interstate or
intrastate commerce that results in:
• A fatality
• Bodily injury to a person who, as a result of the injury, immediately receives medical treatment
away from the scene of the accident
• One or more motor vehicles incurring disabling damage as a result of the accident, requiring the
motor vehicle to be transported away from the scene by a tow truck or other motor vehicle
The term accident does not include:
• An occurrence involving only boarding and alighting from a stationary motor vehicle
• An occurrence involving only the loading or unloading of cargo
• An occurrence in the course of the operation of a passenger car or a multipurpose passenger
vehicle by a motor carrier and is not transporting passengers for hire or hazardous materials of a
type and quantity that require the motor vehicle to be marked or placarded in accordance with 49
CFR Part 177, Subpart 823
fuel from the fuel tanks were released, and to furnish copies of all state-required accident reports [49 CFR
390. 15]. The reason for this rule change was the emergence of an automated State accident reporting
system compiled from law enforcement accident reports that, pursuant to provisions of the Intermodal
Surface Transportation Efficiency Act of 1991 [P.L. 102-240, 105 STAT. 1914], was established under
the Motor Carrier Safety Assistance Program.
Under Section 408 of Title FV of the Motor Carrier Act of 1991, a component of the Intermodal Surface
Transportation Efficiency Act, the Secretary of Transportation is authorized to make grants to states to
help them achieve uniform implementation of the police reporting system for truck and bus accidents
recommended by the National Governors Association. Under this system, called SAFETYNET, accident
data records generated by each state follow identical formatting and content instructions. They are
entered in a Federally maintained SAFETYNET data base on approximately a weekly basis. The
SAFETYNET data base, in turn, is compiled and managed as part of the Motor Carrier Management
Information System.
Accident data compiled from the Bureau of Motor Carrier Safety (now the Office of Motor Carriers in the
Federal Highway Administration), American Petroleum Institute, California Highway Patrol, and
California Department of Transportation provided the basis used by the Modal Study (Fischer et al. 1987,
page B-1) for estimating characteristics of accidents that might involve shipments of spent nuclear fuel
using "large trucks." Although reporting requirements have changed, these data were similar to data
being compiled by the SAFETYNET system for motor carrier accidents in 1999. Most important, the
definition of a motor carrier accident, the basis for reporting and data compilation, has remained basically
unchanged over the 40 years of data collection.
Because the Modal Study is the fundamental source for data that describes the severity of transportation
accidents used in this EIS, the relative constancy of the definition oi accident is important in establishing
confidence in estimated impact results. Thus, although the transportation environment has changed over
the 40 years of data collection, the constancy of the definition of accident tends to provide confidence that
the distribution of severity for reported accidents has remained relatively the same. That is, low-
consequence, fender-bender accidents are the most common, high-consequence, highly energetic
accidents are rare, and the proportions of these have remained roughly the same.
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Transportation
Changes in the transportation environment, such as changes in speed limits and safety technology, tend to
change the accident rate (accidents per vehicle-kilometer of travel). Overall, however, given that the
definition of accident does not change, such changes do not greatly affect the distribution of accident
severities. For example, recent increases in speed limits from 105 to 121 kilometers (65 to 75 miles) per
hour represent about a 25-percent increase in the maximum mechanical energy of vehicles. Other
information aside, this increase could lead to the conclusion that the resulting distribution of accidents
would show an increase for the most severe accidents in comparison to minor accidents. However, the
speed limit increases do not represent a corresponding increase in actual traffic speeds, and would be
unlikely to change the distribution of velocities and, thus, mechanical energies, of severe accidents from
those reported in the Modal Study. These velocities ranged to faster than 137 kilometers (85 miles) per
hour, even though at the time the National speed limit was 89 kilometers (55 miles) per hour.
Rail Carrier Accident Reporting and Definitions
As with regulations governing the reporting of motor carrier accidents. Federal Railroad Administration
regulations generally require the reporting of accidents if there are injuries, fatalities, or property damage.
These regulations have evolved through the years, mostly in response to increasing values of
transportation equipment and commodities. For example, the Federal requirements in the following text
box establish a value-based reporting threshold for damage to vehicles; the value has been indexed to
inflation since 1975.
RAILROAD ACCIDENT/INCIDENT
(49 CFR 225.11)
• An impact between railroad on-track equipment and an automobile, bus, truck, motorcycle,
bicycle, farm vehicle or pedestrian at a highway-rail grade crossing
• A collision, derailment, fire, explosion, act of God, or other event involving operation of railroad
on-track equipment (standing or moving) that results in reportable damages greater than the
current reporting threshold to railroad on-track equipment, signals, track, track structures, and
roadbed
• An event arising from the operation of a railroad which results in:
- Death to any person
- Injury to any person that requires medical treatment
- Injury to a railroad employee that results in:
• A day away from work
• Restricted work activity or job transfer
• Loss of consciousness
• Occupational illness
Rail carriers covered by these requirements must fulfill several bookkeeping tasks. The Federal Railroad
Administration requires the submittal of a monthly status report, even if there were no reportable events
during the period. This report must include accidents and incidents, and certain types of incidents require
immediate telephone notification. Logs of reportable injuries and on-track incidents must be maintained
by the railroads on which they occur, and a listing of such events must be posted and made available to
employees and to the Federal Railroad Administration, along with required records and reports, on
request. The data entries extracted from the reporting format are consolidated into an accident/incident
data base that separates reportable accidents from grade-crossing incidents. These are processed annually
into event, fatality, and injury count tables in the Federal Railroad Administration's Accident/Incident
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Transportation
Bulletin (Saricks and Tompkins 1999, all), which the Office of Safety publishes on the Internet
(http://safetydata.fra.dot.gov/officeofsafety/Prelim/1999/r01.htm).
In contrast to the regulations for motor carriers discussed above, the Federal Railroad Administration
regulations cited above call for the reporting of accidents and incidents. According to the Modal Study,
the Administration defines an accident as "any event involving on-track raikoad equipment that results in
damage to the railroad on-track equipment, signals, track, or track structure, and roadbed at or exceeding
the dollar damage threshold." Train incidents are defined as "events involving on-track railroad
equipment [and non-train incidents arising from the operation of a railroad] that result in the reportable
death and/or injury or illness of one or more persons, but do not result in damage at or beyond the damage
threshold." The Modal Study, because "damage to casks containing spent nuclear fuel will necessarily
involve severe accidents" (hence, substantial damage), used only "train accidents" to form the basis for
developing the conditional probabilities of accident severities.
As with motor carrier operations, the constancy of the definition of a train accident is important in
establishing confidence in the impact. For rail accidents the transportation environment has not changed
dramatically over the years of data collection, and the definition of accident has remained essentially
unchanged (with adjustments for inflation). The constancy of the definition provides confidence that the
distribution of severity for reported accidents has remained relatively the same — low-consequence,
limited-damage accidents are the most common and high-consequence, highly energetic accidents are
rare, and their proportions have remained about the same. Changes in the rail transportation environment,
as in safety and operations technology (for example, shelf-type couplers and tankcar head protection),
have resulted in lower accident rates (per railcar-kilometer of travel) and, in some cases, less severe
accidents. However, because the definition of accident has not changed appreciably, the changes that
have occurred are not the kind that would greatly affect the relative proportions of minor and severe
accidents.
Reporting and Definitions for Marine Casualties and Incidents
As with the regulations governing the reporting of motor carrier and rail accidents, U.S. law (46 USC
6101-6103) requires operators to report marine casualties and incidents if there are injuries, fatalities, or
property damage. In addition, the law requires the reporting of significant harm to the environment.
MARINE CASUALTY AND INCIDENT
(46 USC 6101-6103)
Criteria have been established for the required reporting (by vessel operators and owners) of marine
casualties and incidents involving all United States flag vessels occurring anywhere in the world and
any foreign flag vessel operating on waters subject to the jurisdiction of the United States. An
incident must be reported within five days if it results in:
The death of an individual
Serious injury to an individual
"Material" loss of property (threshold not specified; previously was $25,000)
Material damage affecting the seaworthiness or efficiency of the vessel
Significant harm to the environment
The states collect casualty data for incidents occurring in navigable waterways within their borders, and
there is a uniform state marine casualty reporting system for transmitting these reports to Federal
jurisdiction (the U. S. Coast Guard). Coast Guard Headquarters receives quarterly extracts of the Marine
J-67
Transportation
Safety Information System developed from these sources. This system is a network data base into which
Coast Guard investigators enter cases at each marine safety unit. The analysis uses a Relational Database
Management System. The Coast Guard Office of Investigations and Analysis compiles and processes the
casualty reports into the formats and partitioned data sets that comprise the Marine Safety Information
System data base, which includes maritime accidents, fatalities, injuries, and pollution spills dating to
1941 (however, the file is complete only from about 1991 to the present).
Hazardous Material Transportation Accident and Incident Reporting and Definitions
Radioactive material is a subset of the more general term hazardous material, which includes
commodities such as gasoline and chemical products. The U.S. Department of Transportation Office of
Hazardous Materials estimates that there are more than 800,000 hazardous materials shipments per day,
of which about 7,700 shipments contain radioactive materials.
Hazardous materials transportation regulations (49 CFR 171) contain no distinction between an accident
and an incident, and incident is the term used to describe situations that must be reported. Hazardous
materials regulations (49 CFR 171.15) require the reporting of incidents if:
A person is killed
A person receives injuries requiring hospitalization
The estimated property damage is greater than $50,000
An evacuation of the public occurs lasting one or more hours
One or more major transportation arteries are closed or shutdown for one or more hours
The operational flight pattern or routine of an aircraft is altered
Fire, breakage, spillage, or suspected radioactive contamination occurs involving shipment of
radioactive material
Fire, breakage, spillage, or suspected contamination occurs involving shipment of infectious agents
There has been a release of a marine pollutant in a quantity exceeding 450 liters (about 120 gallons)
for liquids or 400 kilograms (about 880 pounds) for solids
There is a situation that, in the judgement of the carrier, should be reported to the U.S. Department of
Transportation even though it does not meet the above criteria
These criteria apply to loading, unloading, and temporary storage, as well as to transportation. The
criteria involving infectious agents or aircraft are unlikely to be used for spent nuclear fuel or high-level
radioactive waste shipments. Based on these criteria, reportable motor vehicle and rail transportation
situations are far more exclusionary than hazardous material situations.
Carriers (not law enforcement officials) are required to report hazardous materials incidents to the U.S.
Department of Transportation. These reports are compiled in the Hazardous Materials Incident Report
data base. In addition, U.S. Nuclear Regulatory Commission regulations (20 CFR 20.2201, 20.2202,
20.2203) require the reporting of a loss of radioactive materials, exposure to radiation, or release of
radioactive materials.
J-68
Transportation
Sandia National Laboratories maintains the Radioactive Materials Incident Report (RMIR) data base,
which contains incident reports from the Hazardous Materials Incident Report data base that involve
radioactive material. In addition, RMIR contains data from the U.S. Nuclear Regulatory Commission,
state radiation control offices, the DOE Unusual Occurrence Report data base, and media coverage of
radioactive materials transportation incidents. DOE (1995, pages I-l 17) and McClure and Fagan (1998,
all) discuss historic incidents involving spent nuclear fuel that are reported in RMIR as well as incidents
that took place prior to the existence of this data base. RMIR characterizes incidents in three categories:
transportation accidents, handling accidents, and reported incidents. However, the definitions of these
categories are not consistent with the definitions used in other U.S. Department of Transportation data
bases. For example, from 1971 through 1998, RMIR lists one transportation accident involving a loaded
rail shipment of spent nuclear fuel. However, based on current Federal Railroad Administration reporting
requirements, this occurrence probably would be listed as a grade-crossing incident, not an accident. For
this reason and because of the small number of occurrences in the data base involving spent nuclear fuel,
the EIS analysis did not use RMIR to estimate transportation accident rates.
J.1. 4.2.3.2 Accident Rates for Transportation by Heavy-Combination Truck, Railcar, and
Barge in the United States. Saricks and Tompkins (1999, all) developed estimates of accident rates
for heavy-combination trucks, railcars, and barges based on data available for 1994 through 1996. The
estimates provide an update for accident rates published in 1994 (Saricks and Kvitek 1994, all) that
reflected rates from almost a decade earlier.
Rates for Accidents in Interstate Commerce for Heavy-Combination Trucks
Saricks and Tompkins (1999, all) developed basic descriptive statistics for state-specific rates of accidents
involving interstate-registered combination trucks for 1994, 1995, and 1996. The accident rate over all
road types for 1994 was 2.98 x 10'^ accident per truck-kilometer (Saricks and Tompkins, 1999, Table 3a);
for 1995 it was 2.97 x 10"'' accident per truck-kilometer (Saricks and Tompkins, 1999, Table 3b); and for
1996 it was 3.46 x 10"' accident per truck-kilometer (Saricks and Tompkins, 1999, Table 3c). The
composite mean from 1994 through 1996 was 3.21 x 10"' accident per truck-kilometer.
During the 24 years of the Proposed Action, the mostly legal-weight truck national transportation scenario
would involve as many as 50,000 truck shipments of spent nuclear fuel and high-level radioactive waste.
Based on the data in Saricks and Tompkins (1999, Table 4), the transportation analysis estimated that
those shipments could involve as many as 40 accidents. During the same period, the mostly rail scenario
would involve about 2,6(X) truck shipments, and the analysis estimated that as many as two accidents
could occur during these shipments. More than 99 percent of these accidents would not generate forces
capable of causing functional damage to the casks, and would have no radiological consequences. A
small fraction of the accidents could generate forces capable of damaging the cask.
Rates for Freight Railcar Accidents
Results for accident rates for freight railcar shipments from Saricks and Tompkins (1999, all), show that
domestic rail freight accidents, fatalities, and injuries on Class 1 and 2 raikoads have remained stable or
declined slightly since the late 1980s. Based on data from 1994 through 1996, these rates are 5.39 x 10"*,
8.64 X 10"*, and 1.05 x 10"* per railcar-kilometer, respectively (Saricks and Tompkins, 1999, Table 6).
This conclusion is based on applying denominators that do not include train and car kilometers for
intermodal shipments (containers and trailers-on-flatcar) not loaded by the carriers themselves. Thus, the
actual denominators are probably higher and the rates consequently lower, by about 20 percent.
During the 24 years of the Proposed Action, the mostly rail national transportation scenario would
involve as many as 1 1,000 rail shipments of spent nuclear fuel and high-level radioactive waste. Based
on the data in Saricks and Tompkins (1999, Table 6), the analysis estimated that these shipments could
involve one or two accidents. More than 99 percent of these accidents would not generate forces capable
J-69
Transportation
of causing functional damage to the cask; these accidents would have no radiological consequences. A
small fraction of the accidents could generate forces capable of damaging the cask. For the mostly legal-
weight truck scenario, rail accidents would be unlikely during the 300 railcar shipments of naval spent
nuclear fuel.
Rates for Barge Accidents
Waterway results show a general improvement over mid-1980s rates. The respective rates for
450-metric-ton (500-ton) shipments for waters internal to the coast (rivers, lakes, canals, etc.) for accident
and incident involvements and fatalities were 1.68 x 10"^ and 8.76 x lO' per shipment-kilometer,
respectively (Saricks and Tompkins 1999, Table 8b). Rates for lake shipping were lower — 2.58 x 10'^
and 0 per shipment-kilometer, for accidents and incidents and for fatalities, respectively. Coastal casualty
involvement rates have risen in comparison to the data recorded about 10 years ago, and are comparable
to rates for internal waters — 5.29 x 10"^ and 8.76 x 10"' per shipment-kilometer (Saricks and Tompkins
1999, Table 9b).
During the 24 years of the Proposed Action, the mostly rail national transportation scenario could involve
the use of barges to ship spent nuclear fuel from 14 commercial sites. Based on the data in Saricks and
Tompkins (1999, all), the analysis estimated that less than one accident could occur during such
shipments. A barge accident severe enough to cause measurable damage to a shipping cask would be
highly unlikely.
Rates for Safe Secure Trailer Accidents
DOE uses safe secure trailers to transport hazardous cargoes in the continental United States. The criteria
used for reporting accidents involving these trailers are damage in excess of $500, a fire, a fatality, or
damage sufficient for the trailer to be towed. From 1975 through 1998, 14 accidents involved safe secure
trailers over about 54 million kilometers (about 34 million miles) of travel, which yields a rate of
2.6 X 10'^ accident per kilometer (4.2 x 10"' per mile). This rate is comparable to the rate estimated by
Saricks and Tompkins (1999, Table 4) for heavy combination trucks, 3.2 x 10"' accident per kilometer
(5.1xlO"''peri™le)-
J.1. 4.2.3.3 Accident Data Provided by the States of Nevada, California, South Carolina,
Illinois, and Nebrasl(a. In May 1998, DOE requested the 48 contiguous states to provide truck and
rail transportation accident data for use in this EIS. Five states responded - Nevada, California, Illinois,
Nebraska, and South Carolina (Denison 1998, all; Caltrans 1997, all; Wort 1998, all; Kohles 1998, all;
SCDPS 1997, all). No states provided rail information.
•
Nevada. Nevada provided a highway accident rate of 1.1 x 10"* accident per kilometer (1.8 x 10"
per mile) for interstate carriers over all road types. This is higher than the accident rate estimated by
Saricks and Tompkins (1999, Table 4); 2.5 x 10"' accident per kilometer (3.9 x lO'per mile) for
heavy trucks over all road types in Nevada from 1994 to 1996.
The definition of accident used in Saricks and Tompkins (1999, page 4) is the Federal definition
(fatality, injury, or tow-away); in Nevada the accident criteria are fatality, injury, or $750 property
damage. Based on national data from the U.S. Department of Transportation Office of Motor Carrier
Information Analysis (FHWA 1997, page 2; FHWA 1998, pages 1 and 2), using the Federal
definition would reduce the accident rate from 1.1 x 10'* to about 4.1 x 10"' accident per kilometer
(1.8 X 10"* to 6.7 X 10"' per mile). The radiological accident risk in Nevada for the mostly legal-weight
truck scenario would increase over 24 years from 0.0(X)2 latent cancer fatality to about 0.(XX)5 latent
cancer fatality (a likelihood of 5 in 10,000 of one latent cancer fatality) if the accident rate reported by
Saricks and Tompkins for Nevada were replaced by the rate of 4.1 x lO' per kilometer. Thus, the
J-70
Transportation
impacts of the rate for accidents involving large trucks on Nevada highways reported by Nevada
(Denison 1998, all) would be comparable to the impacts derived using rate estimated by Saricks and
Tompkins.
California. California responded with highway accident rates that included all vehicles (cars, buses,
and trucks). The accident rate for hiterstate highways was 4.2 x 10"^ accident per kilometer
(6.8 X 10'^ per mile) for all vehicles in 1996. This rate is higher than the accident rate estimated by
Saricks and Tompkins (1999, Table 4), 1.6 x 10'^ accident per kilometer (2.6 x 10'^ per mile) for
heavy trucks on California interstate highways from 1994 to 1996.
The definition of accident in Saricks and Tompkins (1999, page 4) is the Federal definition (fatality,
injury, or tow-away); in California the accident criteria are fatality, injury, or $500 property damage.
Based on national data from FHWA (1997, page 2) and FHWA (1998, pages 1 and 2), using the
Federal definition would reduce the accident rate from 4.2 x 10"^ to about 1.6 x 10'^ accident per
kilometer (6.8 x 10'^ to 2.6 x 10'^ per mile), hi addition, the rate provided by California was for all
vehicles. Based on national data from the U.S. Department of Transportation Bureau of
Transportation Statistics, using the accident rate for large trucks would reduce the all-vehicle accident
rate from 1.6 x 10'^ to about 1.3 x 10'^ accident per kilometer (2.6 x 10'^ to 2.1 x lO'^per mile) for
large trucks. This rate is slightly less than the rate estimated by Saricks and Tompkins (1999, Table
4), 1.6 X 10"^ accident per kilometer.
Illinois. Dlinois provided highway data for semi -trucks from 1991 through 1995 over all road types.
Over this period, the accident rate was 1.8 x 10"* accident per kilometer (2.9 x 10"* per mile). From
1994 through 1996, Saricks and Tompkins (1999, all) estimated an accident rate of 3.0 x 10"^ accident
per kilometer (4.8 x 10"^ per mile) for heavy trucks over all road types in Illinois.
The definition of accident used in Saricks and Tompkins (1999, page 4) is the Federal definition
(fatality, injury, or tow-away); in Illinois the accident criteria are fatality, injury, or $500 property
damage. Based on national data from the U.S. Department of Transportation Office of Motor Carrier
Information Analysis (FHWA 1997, page 2; FHWA 1998, pages 1 and 2), using the Federal
definition would reduce the accident rate from 1.8 x 10"* to about 6.7 x 10"'' accident per kilometer
(2.9 X 10"* to 1.1 X 10"* per mile). This rate is comparable to the rate estimated by Saricks and
Tompkins (1999, all).
Nebraska. Nebraska provided a highway accident rate of 2.4 x 10'^ accident per kilometer
(3.8 x 10'^ per mile) for 1997. Nebraska did not specify if the rate was for interstate highways, but it
is for interstate truck carriers. This rate is slightly less than the accident rate estimated by Saricks and
Tompkins (1999, all) for Nebraska interstates, 3.2 x 10"' accident per kilometer (5.1 x 10"'' per mile)
for heavy trucks from 1994 through 1996.
South Carolina. South Carolina responded with highway accident rates that included all types of
tractor/trailers (for example, mobile homes, semi-trailers, utility trailers, farm trailers, trailers with
boats, camper trailers, towed motor homes, petroleum tankers, lowboy trailers, auto carrier trailers,
flatbed trailers, and twin trailers). The rate was 8.3 x 10"' accident per kilometer (1.3 x 10"* per mile),
for all road types. [This is higher than the accident rate estimated by Saricks and Tompkins (1999,
all), 4.7 X 10"' accident per kilometer (7.6 x 10'' per mile) for heavy trucks on all road types in South
Carolina from 1994 through 1996].
The definition of accident in Saricks and Tompkins (1999, page 4) is the Federal definition (fatality,
injury, or tow-away); in South Carolina the accident criteria are fatality, injury, or $1,000 property
J-71
Transportation
damage. Based on national data from the U.S. Department of Transportation Office of Motor Carrier
Information Analysis (FHWA 1997, page 2; FHWA 1998, pages 1 and 2), using the Federal
definition of an accident would reduce the accident rate from 8.3 x 10"^ to about 3.1 x 10"^ accident
per kilometer (1.3 x 10"^ to 5.0 x 10'^ per mile), which is slightly less than the rate estimated by
Saricks and Tompkins (1999, all), 4.7 x 10"^ accident per kilometer (7.6 x 10"^ per mile). In addition,
the accident rate estimated by Saricks and Tompkins (1999, all) was based on Motor Carrier
Management Information System vehicle configuration codes 4 through 8 (truck/trailer, bobtail,
tractor/semi-trailer, tractor/double, and tractor/triple), while the rate obtained fi-om South Carolina
included all truck/trailer combinations. Including all of the combinations tends to increase accident
rates; for example, light trucks have higher accident rates than heavy trucks (BTS 1999, Table 3-22).
DOE evaluated the effect of using the data provided by the five states on radiological accident risk for the
mostly legal-weight truck national transportation scenario. If the data used in the analysis for the five
states (Saricks and Tompkins 1999, Table 4) were replaced by the data provided by the states with the
adjustments discussed, the change in the resulting estimate of radiological accident risk would be small,
increasing from 0.067 to 0.071 latent cancer fatality. Using the unadjusted data provided by those states
would result in an increase in accident risk from 0.067 to 0.093 latent cancer fatality.
J.1 .4.2.4 Transportation Accidents Involving Nonradioactive Hazardous Materials
The analysis of impacts of transportation accidents involving the transport of nonradioactive hazardous
materials to and from Yucca Mountain used information presented in two U.S. Department of
Transportation reports (DOT 1998b, Table 1; BTS 1996, page 43) on the annual number of hazardous
materials shipments in the United States and the number of deaths caused by hazardous cargoes in 1995.
In total, there are about 300 million annual shipments of hazardous materials; only a small fraction
involve radioactive materials. In 1995, 6 fatalities occurred because of hazardous cargoes. These data
suggest a rate of 2 fatalities per 1(X) million shipments of hazardous materials. DOE anticipates about
40,000 shipments of nonradioactive hazardous materials (including diesel fuel and laboratory and
industrial chemicals) to and from the Yucca Mountain site during construction, operation and monitoring,
and closure of the repository. Assuming that the rate for fatalities applies to the transportation of
nonradioactive hazardous materials to and from Yucca Mountain, DOE does not expect fatalities from
40,000 shipments of these materials.
J.2 Evaluation of Rail and Intermodal Transportation Options
DOE could use several modes of transportation to ship spent nuclear fuel from the 77 sites. Legal-weight
trucks could be used to transport spent nuclear fuel and high-level radioactive waste contained in truck
casks that would weigh approximately 22,5(X) kilograms (25 tons) when loaded. For sites served by
railroads, rail casks placed on railcars could be used to ship directly to the Yucca Mountain site if a
branch rail line was constructed in Nevada or to ship to an intermodal transfer station in Nevada if heavy-
haul trucks were used.
For sites not served by a railroad that nonetheless have the capability to load rail casks, DOE could use
heavy-haul trucks or, for sites located on navigable waterways, barges to transport the casks between the
generating sites and nearby railheads.
For rail shipments, DOE could request the railroads provide dedicated trains to transport casks from sites
to a destination in Nevada or could deliver railcars with loaded casks to the railroads as general freight for
delivery in Nevada.
J-72
Transportation
J.2.1 IMPACTS OF THE SHIPMENT OF COMMERCIAL SPENT NUCLEAR FUEL BY
BARGE AND HEAVY-HAUL TRUCK FROM 19 SITES NOT SERVED BY A RAILROAD
An alternative to truck or rail transport of commercial spent nuclear fuel, barge transportation, was
evaluated. Nineteen commercial sites that have the capability to handle and load rail casks are not served
by a railroad. Accordingly, under the mostly rail transportation scenario the 19 sites were assumed to use
heavy-haul trucks to move the rail casks to nearby railheads. However, because 14 of the sites are on
navigable waterways (see Figure J-9), some could use barges to ship to nearby railheads. The following
sections present the analysis of impacts of using barges and compares these impacts from one of the
fourteen sites located on a navigable waterway (Turkey Point) to the impacts based on the use of heavy-
haul trucks and legal-weight truck. The analysis assumed that all five of the DOE sites would have
railroad service.
Unlike previous sections, where impacts were presented for all shipments by mode (mostly legal-weight
truck and mostly rail), impacts are reported on a per shipment basis and compared on that basis to
shipments via heavy-haul truck and legal-weight truck for the same reactor site.
J.2.1. 1 Routes for Barges and Heavy-Haul Trucks
The heavy-haul truck-to-railhead distances for the 19 sites range from about 6 to 75 kilometers (4 to
47 miles). Routing for heavy-haul trucks was estimated using the HIGHWAY computer code (Johnson
et al. 1993a, all). The INTERLINE computer code (Johnson et al. 1993b, all) was used to generate route-
specific distances that would be traveled by barges. The resulting estimates for route lengths for barges
and heavy-haul trucks are listed in Table J-26. Table J-27 lists the number of shipments from each site.
J.2.1 .2 Analysis of Incident-Free Impacts for Barge and Heavy-Haul Truck Transportation
J.2.1 .2.1 Radiological Impacts of Incident-Free Transportation
This section compares the radiological and nonradiological impacts to populations and maximally
exposed individuals of incident-free transportation of spent nuclear fuel from one commercial spent
nuclear fuel site (Turkey Point) for:
• Shipments using heavy-haul trucks to the nearest railhead and then to the Nevada Caliente node by
rail and finally to the Yucca Mountain site by rail using the Caliente-Chalk Mountain corridor.
• Shipments using barge to a nearby railhead (Port of Miami for the Turkey Point site) and then to the
Nevada Caliente node by rail and finally to the Yucca Mountain site by rail using the Caliente-Chalk
Mountain corridor.
• Shipments using legal-weight trucks to the Yucca Mountain site.
The radiological impacts of intermodal transfers at the interchange from heavy-haul trucks to railcars or
barges to railcars were included in the analysis. Workers would be exposed to radiation from casks
during transfer operations. However, because the transfers would occur in terminals and berths that are
remote from public access, public exposures would be small. Impacts of constructing intermodal transfer
facilities were not included because intermodal transfers were assumed to take place at existing facilities.
The analysis assumed that heavy-haul trucks, though they would be slower moving vehicles, would result
in the same types of impacts as, although somewhat higher than, an equal number of legal-weight truck
shipments over the same routes. Because travel distances to nearby railheads would be short, impacts of
J-73
Transportation
CO
O
00
a
u
J3
u
C
e
i
X)
3
O
S
J-74
Transportation
m
o
u
St
CO
u
5
X5
c
en
U
3
22
9
J-75
Transportation
Legend
ir Commercial
powerplant site
• Port
> Barge route
Figure J-9. Routes for barges from sites to nearby railheads (page 3 of 3).
J-76
Transportation
Table J-26. National transportation distances from commercial sites to Nevada ending rail nodes
(kilometers)'*' (page 1 of 2).
Site
State
Destination
Rail transportation
Barge transportation
(intermodal rail ncde)'
Total''
Rural
Suburban
Urban
Total"
Rural
Suburban
Urban
Browns Ferry NP°
AL
Apex
3,5%
3,269
281
46
57
52
5
0
Caliente
3,423
3,095
281
46
57
52
5
0
Beowawe
3,278
2,990
254
34
57
52
5
0
Jean
3,678
3,333
293
51
57
52
5
0
Diablo Canyon NP
CA
Apex
644
420
124
100
143
143
0
0
Caliente
817
594
124
100
143
143
0
0
Beowawe
1,439
1,005
291
141
143
143
0
0
Jean
562
355
112
94
143
143
0
0
St. Lucie NP
PL
Apex
5,203
4,293
812
97
140
50
52
39
Caliente
5,029
4,119
812
97
140
50
52
39
Beowawe
4,885
4,014
784
86
140
50
52
39
Jean
5,284
4,358
823
103
140
50
52
39
Turkey Point NP
FL
Apex
5,245
4,2%
820
127
54
53
0
1
Caliente
5,071
4,123
820
127
54
53
0
1
Beowawe
4,927
4,017
793
116
54
53
0
1
Jean
5,326
4,361
832
133
54
53
0
1
Calvert CUffs NP
MD
Apex
4,344
3,558
645
140
99
98
2
0
Caliente
4,170
3,385
645
140
99
98
2
0
Beowawe
4,026
3,279
618
129
99
98
2
0
Jean
4,425
3,623
657
145
99
98
2
0
PaUsades NP
MI
Apex
3,375
2,895
391
90
256
256
0
0
Caliente
3,202
2,722
391
90
256
256
0
0
Beowawe
3,058
2,616
363
78
256
256
0
0
Jean
3,457
2,960
402
95
256
256
0
0
Grand Gulf NP
MS
Apex
3,686
3,355
291
39
51
51
0
0
CaUente
3,512
3,181
291
39
51
51
0
0
Beowawe
3,368
3,076
264
28
51
51
0
0
Jean
3,767
3,419
303
44
51
51
0
0
Cooper NP
NE
Apex
2,345
2,193
119
33
117
100
16
1
Caliente
2,171
2,020
119
33
117
100
16
1
Beowawe
2,027
1,914
92
21
117
100
16
1
Jean
2,426
2,258
130
38
117
100
16
1
Salem/Hope Creek NP
NJ
Apex
4,423
3,410
818
194
30
30
0
0
Caliente
4,250
3,236
818
194
30
30
0
0
Beowawe
4,106
3,131
791
183
30
30
0
0
Jean
4,505
3,475
830
200
30
30
0
0
Oyster Creek NP
NJ
Apex
4,532
3,371
933
227
130
77
36
17
Caliente
4,358
3,198
933
227
130
77
36
17
Beowawe
4,214
3,092
906
216
130
77
36
17
Jean
4,613
3,436
944
232
130
77
36
17
Surry NP
VA
Apex
4,583
3,982
532
68
71
60
8
3
Caliente
4,409
3,809
532
68
71
60
8
3
Beowawe
4,265
3,703
505
57
71
60
8
3
Jean
4,664
4,047
544
73
71
60
8
3
Kewaunee NP
WI
Apex
3,180
2,789
312
79
293
285
2
7
Caliente
3,007
2,616
312
79
293
285
2
7
Beowawe
2,863
2,510
285
68
293
285
2
7
Jean
3,262
2,854
323
84
293
285
2
7
Point Beach NP
WI
Apex
3,180
2,789
312
79
301
293
2
7
CaUente
3,007
2,616
312
79
301
293
2
7
Beowawe
2,863
2,510
285
68
301
293
2
7
Jean
3,262
2,854
323
84
301
293
2
7
Callaway NP
MO
Apex
2,7%
2,625
140
31
__f
--
--
--
HH-18.S kilometers
Caliente
2,624
2,452
140
31
-
~
-
-
Beowawe
2,491
2,358
113
20
-
-
-
-
Jean
2,878
2,689
151
37
--
--
--
--
Fort Calhoun NP
NE
Apex
2,301
2,177
102
21
--
--
-
--
HH - 6.0 kilometers
Caliente
2,129
2,005
102
21
~
--
--
-
Beowawe
1,9%
1,911
75
10
--
-
--
-
Jean
2,383
2,242
114
27
-
-
-
-
J-77
Transportation
Table J-26. National transportation distances from commercial sites to Nevada ending rail nodes
(kilometers)"'' (page 2 of 2).
Site
State
E>estination
Rail transportation
Barge transportation
(intermodal rail node)'
Total"
Rural
Suburban
Urban
Total" Rural Suburban Urban
Peach Bottom NP°
PA
Apex
4,294
3,324
779
191
-'
HH - 58.9 kilometers
Caliente
4,121
3,151
779
191
-
Beowawe
3,988
3,057
752
179
..
Jean
4,375
3,388
790
196
-
Oconee NP
SC
Apex
4,247
3,651
534
61
—
HH- 17.5 kilometers
Caliente
4,074
3,479
534
61
„
Beowawe
3,941
3,385
507
50
..
Jean
4,328
3,716
546
66
-
a. To convert kilometers to miles, multiply by 0.62137.
b. Distances estimated using INTERLINE computer program.
c. Intermodal rail nodes selected for purpose of analysis. Source; TRW (1999a, all).
d. Totals might differ from sums of rural, suburban, and urban distances due to method of calculation and rounding.
e. NP = nuclear plant.
f. -- = the four sites that are not located on a navigable waterway.
Table J-27. Barge shipments and ports.
Number of shipments
Proposed
Modules 1
Barge ports assumed for barge-to-rail
Plant name
State
Action
and 2
intermodal transfer
Browns Ferry 1
AL
176
253
Wilson L/D
Browns Ferry 3
AL
67
114
Wilson L/D
Diablo Canyon 1
CA
64
129
Port Huememe
Diablo Canyon 2
CA
59
149
Port Huememe
St. Lucie 2
FL
56
103
Port Everglades
Turkey Point 3
FL
56
80
Port of Miami
Turkey Point 4
FL
57
89
Port of Miami
Calvert Cliffs 1
MD
144
204
Port of Baltimore
Palisades
MI
70
70
PortofMuskegan
Grand Gulf 1
MS
79
154
Port of Vicksburg
Cooper Station
NE
103
159
Port of Omaha
Hope Creek
NJ
59
146
Port of Wilmington
Oyster Creek 1
NJ
87
87
Port of Newark
Salem 1
NJ
63
104
Port of Wilmington
Salem 2
NJ
57
112
Port of Wilmington
Surry 1
VA
102
128
Port of Norfolk
Kewaunee
WI
57
70
Port of Milwaukee
Point Beach 1
WI
90
102
Port of Milwaukee
Totals
1,833
2,970
heavy-haul truck transportation would be much less than the impacts of national rail shipments. The
analysis of impacts for barge shipments assumed the transport would employ commercial vessels
operated by maritime carriers on navigable waterways and that these shipments would follow direct
routing from the sites to nearby railheads. For both modes, intermodal transfers would be necessary to
transfer rail casks to railcars.
Radiological impacts were estimated for workers and the general population. For heavy-haul truck
shipments, workers included vehicle drivers and escorts. For barge shipments, the work crew included
five members on board during travel and workers close to the shipping casks during inspections or
intermodal transfers. The general population for truck shipments included persons within 8(X) meters
(about 2,6(X) feet) of the road (offlink), persons sharing the road (onlink), and persons at stops. The
general population for barging included persons within a range of 200 to 1,000 meters (about 660 to
3,3(X) feet) of the route, and persons at stops. On-link exposures to members of the public during barging
J-78
Transportation
were assumed to be small. Incident-free unit risk factors were developed to calculate occupational and
general population collective doses. Table J-28 lists the unit risk factors for heavy-haul truck and barge
shipments. The unit risk factors for heavy-haul truck shipments reflect the effects of slower operating
speeds for those vehicles in comparison to those for legal-weight trucks.
Table J-28. Risk factors for incident-free heavy-haul truck and barge transportation
of spent nuclear fuel and high-level radioactive waste.
Incident free risk factors
Exposure group
(person
-rem per kilometer)"
Mode
Rural
Suburban
Urban
Heavy-haul truck
Occupational
General population
1.1x10-'
1.1x10"'
1.9x10"'
Offlink''
7.3x10*
7.7x10"*
8.3x10"*
Onlink'
1.1x10^
1.2x10"^
5.5x10"^
Stops
1.9x10"^
1.9x10"^
1.9x10"^
Storage"*
1.9x10'
1.9x10"'
1.9x10"'
Totals
2.2x10-'
2.3x10"'
2.7x10'
Barge
Occupational^
General population
9,4x10"'
1.9x10"'
4.8x10"*
Offlink"
8.6x10"*
1.7x10"'
4.3x10"'
Onlink'
0.0
0.0
0.0
Stops
5.4x10"'
5.4x10"'
5.4x10"'
Totals
5.4x10"'
5.4x10'
5.5x10"'
The methodology, equations, and data used to develop the unit dose factors are discussed in Madsen
et al. (1986, all) and Neuhauser and Kanipe (1992, all). Cashwell et al. (1986, all) contains a detailed
explanation of the use of unit factors.
Offlink general population included persons within 800 meters (about 2,600 feet) of the road or
railway.
Onlink general population included persons sharing the road or railway.
The storage unit risk factor is only applied for heavy-haul truck shipments requiring an ovemight stop.
Table J-29 lists the incident- free impacts on a per shipment basis from the Turkey Point nuclear power
plant using the three shipment scenarios listed above. This is presented to compare the impacts on a per
shipment basis using barge, heavy-haul truck or legal weight truck. Impacts of intermodal transfers are
included in the results. Occupational impacts would include the estimated radiological exposures of
security escorts.
Table J-29. Comparison of population doses and impacts from incident-
free national transportation for heavy-haul-to-rail, barge-to-rail, and legal-
weight truck options.^''
Category
Heavy-haul
to rail
Barge to rail
Legal-weight
truck
Involved worker
Collective dose (person-rem)
Estimated LCFs'
0.15
0.00006
0.13
0.00005
0.32
0.00013
Public
Collective dose (person-rem)
Estimated LCFs
0.12
0.00006
0.41
0.0002
1
0.0005
Maximally exposed individual
Impacts would be the same as those
Chapter 6, Tables 6-9 and 6-12
in
a. Rail impacts are presented for the Caliente-Chalk Mountain rail implementing alternative.
b. Impacts presented on a per shipment basis for the Turkey Point site.
c. LCF = latent cancer fatality.
J-79
Transportation
As indicated in Table J-29, differences in radiological impacts between the use of heavy-haul trucks and
barges would be small. The impacts to maximally exposed individuals would be the same because both
cases use the same assumptions for locations of such individuals in relation to shipments and times of
exposure.
J.2.1 .2.2 Nonradiological Impacts of Incident-Free Transportation (Vehicle Emissions)
Table J-30 compares the estimated number of fatalities from vehicle emissions from shipments, assuming
the use of heavy-haul trucks or barges to ship to nearby railheads.
Table J-30. Population health impacts from vehicle emissions during
incident-free national transportation for mostly legal-weight truck
scenario.^
Legal-weight
Category Heavy-haul to rail Barge to rail truck
Estimated fatalities 0.00004 0.00004 0.00003
a. Impacts are presented on a per shipment basis for the Turkey Point site.
J.2.1 .3 Analysis of Impacts of Accidents for Barge and l-ieavy-Haul Truck Transportation
J.2.1 .3.1 Radiological Impacts of Accidents
The analysis of risks from accidents during heavy-haul truck, rail, and legal-weight truck fransport of
spent fuel and high-level radioactive waste used the RADTRAN4 computer code (Neuhauser and Kanipe
1992, all) and the analysis approach discussed in Section J. 1.4.2. The analysis of risks due to barging
used the same methodology with the exception of conditional probabilities. For barge shipments, the
conditional accident probabilities (Table J-3 1) for each cask response category were based on a review of
other barge accident analyses.
Table J-31. Conditional probabilities for barge fransportation.
Severity category 12 3
4
5
6
Conditional probability 0.93794 0.005 0.000
0.057
0.000051
0.0000058
When radioactive material is shipped by barge, it is possible to have both water and land contamination.
The analysis assumed that airborne releases could occur in accidents involving barges. Any portion of a
release plume over water would result in water contamination. Thus, there are two mechanisms for
contaminating water and one, the airborne release, for contaminating land surfaces.
For accident scenarios that result in releases of radioactive material, part of the plume would be deposited
on water and part on land. For coastal and lake shipping, the analysis assumed that, 50 percent of the
time, the plume would be entirely deposited on water. For the other 50 percent, the analysis assumed that
the accident would occur about 200 meters (660 feet) from the shore and any material deposited in the
first 200 meters would be into water. The analysis used the methods used by the RISKIND computer
program (Yuan et al. 1995 all) to estimate plume depletion into water for D stability and a wind speed of
3 meters per second. For these conditions, about 20 percent of the plume would be depleted in the first
200 meters. Based on this information, the analysis assumed that for coastal and lake shipping, 60
percent of the plume would be deposited on water and for river transport only 20 percent of the release
would occur over water.
The analysis accommodated this split by allocating 60 percent of coastal and lake shipping to what was
called a "water" state and the remaining 40 percent to an adjoining state (Florida in the case of Turkey
J-80
Transportation
Point). For river transport, 20 percent of the mileage was allocated to the water state representing the
river and the remaining 80 percent of the mileage was allocated to the adjacent state (Mississippi in the
case of Browns Ferry).
The dose from plume release to water was limited to an ingestion dose. The transfer coefficients that
were used in the calculation are listed in Table J-32. The selection of isotopes and the transfer
coefficients was based on models used in the Foreign Spent Nuclear Fuel EIS (DOE 1996a, page E-126).
The same water uptake models were used. Both the freshwater and ocean models considered fish
consumption. The freshwater model included irrigation and domestic water consumption by both the
general population and livestock. The ocean model included uptake from eating shellfish.
Table J-32. Food transfer factors used in the barge
analysis.
Isotope
Ocean release Freshwater release
Hydrogen-3 (tritium)
0.000020
Niobium-95
0.080
Ruthenium- 106
0.00014
Cesium- 134
0.00037
0.000022
Cesium- 137
0.00037
0.000022
In addition, the analysis of barge accident risks used the following assumptions:
• Release fractions that determine the source term for dispersion to the waterway are the same as those
developed for airborne release scenarios
For freshwater river systems, the analysis assessed the following exposure pathways:
• Drinking water
• Ingestion of fish by humans
• Ingestion of irradiated foods
• Shoreline deposits
• External irradiation from immersion during swimming
For marine coastal systems, the following exposure pathways were assessed:
• Ingestion of fish and invertebrates by humans
• External irradiation from shoreline deposits
• External irradiation from immersion during swimming
Route-specific collective doses were calculated using population distributions along the routes developed
from 1990 Census data. As an example. Table J-33 presents the dose risk per shipment for the Turkey
Point nuclear power plant.
Table J-33. Accident risks for shipping spent nuclear fuel from Turkey Point.
Category
Heavy-haul to rail
Barge to rail Legal- weight truck
Dose risk (person-rem)
Dose risk (LCF)'
Traffic fatalities
0.0038
0.000002
0.00039
0.0019
0.0000009
0.00039
0.0023
0.000001
0.00011
a. LCF = latent cancer fatality.
J-81
Transportation
J.2.1 .3.2 Nonradiological Accident Risks
The fatalities per shipment for heavy-haul truck, barge, and legal-weight truck transport from Turkey
Point would be 3.9 x 10"*, 3.9 x IQ-* and 1.1 x 10"'* , respectively.
J.2.1 .3.3 Maximum Reasonably Foreseeable Accidents
With the relatively short barging distance relative to the rail distance traveled, the probability of a barge
accident is much lower than the 1 x 10"^-criteria used for accidents that are reasonably foreseeable.
J.2.2 EFFECTS OF USING DEDICATED TRAINS OR GENERAL FREIGHT SERVICE
The Association of American Railroads recommends that only special (dedicated) trains move spent
nuclear fuel and certain other forms of radioactive materials (DOT 1998b, page 2-6). In developing its
recommendation, the Association concluded that the use of special trains would provide operational (for
railroads and shippers) and safety advantages over shipments that used general freight service.
Notwithstanding this recommendation, the Department of Transportation study (DOT 1998b, all)
compared dedicated and regular freight service using factors that measure impacts to overall public
safety. The results of this study indicated that dedicated trains could provide advantages over regular
trains for incident-free transportation but could be less advantageous for accident risks. However,
available information does not indicate a clear advantage for the use of either dedicated trains or general
freight service. Thus, DOE has not determined the commercial arrangements it would request from
railroads for shipment of spent nuclear fuel and high-level radioactive waste. Table J-34 compares the
dedicated and general freight modes. These comparisons are based on the findings of the Department of
Transportation study and the Association of American Railroads.
J.3 Nevada Transportation
With the exceptions of the possible construction of a branch rail line or upgrade of highways for use by
heavy-haul trucks and the construction of an intermodal transfer station, the characteristics of the
transportation of spent nuclear fuel and high-level radioactive waste in Nevada would be similar to those
for transportation in other states across the nation. Unless the State of Nevada designated alternative or
additional preferred routes as prescribed under regulations of the Department of Transportation (49 CFR
397.103), Interstate System Highways (1-15) would be the preferred routes used by legal-weight trucks
carrying spent nuclear fuel and high-level radioactive waste. Unless alternative or non-Interstate System
routes have been designated by states. Interstate system Highways would also be the preferred routes used
by legal-weight trucks in other states during transit to Nevada.
In Nevada as in other states, rail shipments would, for the most part, be transported on mainline tracks of
major railroads. Operations over a branch rail line in Nevada would be similar to those on a mainline
railroad, except the frequency of train travel would be much lower. Shipments in Nevada that used
heavy-haul trucks would use Nevada highways in much the same way that other overdimensional,
overweight trucks use the highways along with other commercial vehicle traffic.
In some cases State-specific assumptions were used to analyze human health and safety impacts in
Nevada. A major difference would be that much of the travel in the State would be in rural areas where
population densities are much lower than those of many other states. Another difference would be for
travel in an urban area in the state. The most populous urban area in Nevada is the Las Vegas
metropolitan area, which is also a major resort area with a high percentage of nonresidents. The analysis
also addressed the channeling of shipments from the commercial and DOE sites into the transportation
arteries in the southern part of the State. Finally, the analysis addressed the commuter and commercial
J-82
Transportation
Table J-34. Comparison of general freight and dedicated train service.
Attribute
General freight
Dedicated train
Overall accident rate for
accidents that could damage
shipping casks
Grade crossing, trespasser,
worker fatalities
Security
Incident-free dose to public
Radiological risks from
accidents
Occupational dose
Utilization of resources
Same as mainline raitoad accident
rates
Same as mainline railroad rates for
fatalities
Security provided by escorts required
by NRC regulations
Low, but more stops in classification
yards than dedicated trains. However,
classification yards would tend to be
remote from populated areas.
Low, but greater than dedicated trains
Duration of travel influences dose to
escorts
Long cross-country transit times
could result in least efficient use of
expensive transportation cask
resources; best use of railroad
resources; least reliable delivery
scheduling; most difficult to
coordinate state notifications.
Expected to be lower than general
freight service because of operating
restrictions and use of the most up-to-
date railroad technology.
Uncertain. Greater number of trains
could result in more fatalities in grade
crossing accidents. Fewer stops in
classification yards could reduce work
related fatalities and trespasser fatalities.
Security provided by escorts required
by NRC regulations; fewer stops in
classification yards than general freight
service.
Lower than general freight service.
Dedicated trains could be direct routed
with fewer stops in classification yards
for crew and equipment changes.
Lower than general freight service
because operating restrictions and
equipment could contribute to lower
accident rates and reduced likelihood of
maximum severity accidents.
Shorter travel time would result in
lower occupational dose to escorts.
Direct through travel with on-time
deliveries would result in most efficient
use of cask resources; least efficient use
of railroad resources. Railroad resource
demands from other shippers could lead
to schedule and throughput conflicts.
Easiest to coordinate notification of
state officials.
a NRC = U.S. Nuclear Regulatory Commission.
travel that would occur on highways in the southern part of the State as a consequence of the construction,
operation and monitoring, and closure of the proposed repository.
This section presents information specific to Nevada that DOE used to estimate impacts for transportation
activities that would take place in the State. It includes results for cumulative impacts that would occur in
Nevada for transportation associated with Inventory Modules I and 2.
J.3.1 TRANSPORTATION MODES, ROUTES, AND NUMBER OF SHIPMENTS
J.3.1 .1 Routes in Nevada for Legal-Weight Trucks
The analysis of impacts that would occur in Nevada used the characteristics of (1) highways in Nevada
that would be used for shipments of spent nuclear fuel and high-level radioactive waste by legal-weight
trucks, (2) rail routes from the border to rail nodes where the implementing alternatives would connect,
and (3) rail corridors and highway routes analyzed for the rail and heavy-haul truck implementing
alternatives in the State.
J-83
Transportation
Figure J-10 shows the routes in Nevada that legal-weight trucks would use unless the State designated
alternative or additional preferred routes. The figure shows estimates for the number of legal-weight
truck shipments that would travel on each route segment for the mostly legal-weight truck and mostly rail
transportation scenarios. The inset on Figure J-10 shows the proposed Las Vegas Beltway and the routes
DOE anticipates legal-weight trucks traveling to the repository would use.
J.3.1.2 Routes in Nevada for Transporting Rail Casks
The rail and heavy-haul truck implementing alternatives for transportation in Nevada include five
possible rail corridors and five possible routes for heavy-haul trucks; the corridors and routes for these
implementing alternatives are shown in Figures J-1 1 and J-12. These figures also show the estimated
number of rail shipments that would enter the State on mainline railroads. These numbers indicate
shipments that would arrive from the direction of the bordering state for each of the implementing
alternatives for the mostly rail transportation scenario.
Table J-35 lists the total length and cumulative distance in rural, suburban, and urban population zones in
the State of Nevada used to analyze impacts of the implementing alternatives. Table J-36 lists the total
population that lives within 800 meters (0.5 mile) of rail lines in Nevada. The estimated population that
would live along each branch rail line was based on population densities along existing mainline railroads
in Nevada.
Nevada Heavy-Haul Truck Scenario
Tables J-37 through J-41 summarize the road upgrades for each of the five possible routes for heavy -haul
trucks that DOE estimates would be needed before routine use of a route to ship casks containing spent
nuclear fuel and high-level radioactive waste.
Nevada Rail Corridors
Under the mostly rail scenario, DOE could construct and operate a branch rail line in Nevada. Based on
the studies listed below, DOE has narrowed its consideration for a new branch rail line to five potential
rail corridors — the Carlin, Caliente, Caliente-Chalk Mountain, Jean, and Valley Modified routes. DOE
identified the five rail corridors through a process of screening potential rail alignments that it had studied
in past years. Several studies evaluated rail options.
•
•
The Feasibility Study for Transportation Facilities to Nevada Test Site study (Holmes «fe Narver
1962, all) determined the technical and economic feasibility of constructing and operating a railroad
from Las Vegas to Mercury.
The Preliminary Rail Access Study (Tappen and Andrews 1990, all) identified 13 and evaluated 10
rail corridor alignment options. This study recommended the Carlin, Caliente, and Jean corridors for
detailed evaluation.
The Nevada Railroad System: Physical, Operational, and Accident Characteristics (DOE 1991, all)
described the operational and physical characteristics of the current Nevada railroad system.
The High Speed Surface Transportation Between Las Vegas and the Nevada Test Site (NTS) report
(Raytheon 1994, all) explored the rationale for a potential high-speed rail corridor between Las Vegas
and the Nevada Test Site to accommodate personnel.
J-84
Transportation
lis
Approximately
43,950 shipments
over 24 years
under the mostly
legal-weight trudt I"
scenario
^>^oapa ^Q5/"*Mesquite
Reservation
Arizona
Las Vegas Metropolitan Area
Potential routes for legal-weight truck shipments in Nevada
comply with U.S. Department of Transportation regulations
(49 CFR 397.101) for selecting "preferred routes" and
"delivery routes" for motor carrier shipments of highway
route-controlled quantities of radioactive materials. The
State of Nevada could designate alternative routes as
specified in 49 CFR 397.103.
t
■"-■-•- Route for highway route-controlled
quantities of radioactive material
Highways
State line
County line
10
20 Miles
10 0 10 20 Kilometers
Soufce: Derived from DOE (1997c. eX),
andDOE(1996d.all).
Figure J-10. Potential Nevada routes for legal-weight truck shipments of spent nuclear fuel and high-level
radioactive waste to Yucca Mountain.
J-85
Transportation
Oregon Idaho
Approximate rail
shipments over 24 years
under ttie mostly rail
scenario
Caliente route
Carlin route
Caliente-Ctialk
Mountain route
Jean route
Valley Modified
route
0
12,227
Approximate rail
shipments over 24 years
under the mostly rail
scenario
12,701
0
Caliente route
Carlin route
Caliente-Chalk
Mountain route 12,701
Joan route 1 1 ,579
Valley Modified
12,571
Approximate rail
shipments over 24 years
under the mostly rail
scenario
Caliente route
Carlin route
Caliente-Chalk
Mountain route
Jean route
Valley Modified
route
0
625
Approximate rail
shipments over 24 years
under the mostly rail
scenario
Caliente route
Carlin route
Caliente-Chalk
Mountain route
Jean route
Valley Modified
route
Approximately
2,600 truck
shipments
Legend
I I I I I
Existing rail line
Highway
State line
County line
Potential rail corridor
Variation of potential
rail corridor
Approximate total truck
shipments = 2,600; approximate
total rail shipments = 13,416.
Caliente route
Carlin route
Caliente-Chalk
Mountain route
Jean route
Valley Modified
route
0 truck
shipments
40 Miles
I
50
50 Kilometers
Source: Modified Iroin DOE (19
Figure J-11. Potential Nevada rail routes to Yucca Mountain and approximate number of shipments for
each route.
J-86
Transportation
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J-87
Transportation
Table J-35. Route characteristics for rail and heavy-haul truck
implementing alternatives.
Alternative
Rail
node
Distance (kilometers)'
Rural Suburban Urban Total
Rail
Caliente
Caliente
513
0
Carlin
Beowawe
520
0
Caliente-Chalk Mountain
Caliente
345
0
Jean
Jean
181
0
Valley Modified
Apex
159
0
Heavy-haul'^
Caliente
Caliente
533
0
Caliente-Chalk Mountain
Caliente
282
0
Caliente-Las Vegas
Caliente
356
21
Apex/Dry Lake
Apex
162
21
Sloan/ Jean
Jean
145
43
0
513
0
520
0
345
0
181
0
159
0
533
0
282
0
377
0
183
0
188
a. To convert kilometers to miles, multiply by 0.62137.
b. Rounded to the nearest kilometer.
c. Heavy-haul distances are based on using the Northern,
in the Las Vegas area. These beltways are assumed to
Western, and Southern Beltways
have suburban population density.
Table J-36. Populations in Nevada within 800 meters (0.5 mile) of routes.
Population
Transportation scenario 1990 Census
Legal-weight truck routes"
Rail routes Nevada border to branch rail line
Caliente
Carlin
Caliente-Chalk Mountain
Jean
Valley Modified
Branch rail lines'^
Caliente
Carlin
Caliente-Chalk Mountain
Jean
Valley Modified
60,000
30,000
52,000
30,000
30,000
30,000
2,600
2,700
1,800
900
800
a. Source: TRW (1999a, Table 5-1).
b. Source: TRW (1999a, Table 5-2).
c. Estimated using 3.2 persons p)er square kilometer - the highest value for rural populations
along mainline railroads in Nevada (TRW 1999a, Table 5-2).
• The Nevada Potential Repository Preliminary Transportation Strategy, Study 1 (TRW 1995, all),
reevaluated 13 previously identified rail routes and evaluated a new route called the Valley Modified
route. This study recommended four rail routes for detailed evaluation — the Caliente, Carlin, Jean,
and Valley Modified routes.
• The Nevada Potential Repository Preliminary Transportation Strategy, Study 2 (TRW 1996, all),
further refined the analyses of potential rail corridor alignments presented in Study 1.
Public comments submitted to DOE during hearings on the scope of this environmental impact statement
resulted in addition of a fifth potential rail corridor — Caliente-Chalk Mountain.
J-88
Transportation
Table J-37. Potential road upgrades for Caliente route."
Route
Upgrades
Intermodal transfer station to U.S. 93
U.S. 93 to State Route 375
State Route 375 to U.S. 6
U.S. 6 to U.S. 95
U.S. 95 to Lathrop Wells Road
Lathrop Wells Road to Yucca Mountain
site
Pave existing gravel road.
Asphalt overlay on existing pavement, truck lanes where grade is
greater than 4 percent (minimum distance of 460 meters'" per lane),
turnout lanes every 32 kilometers'^ (distance of 305 meters per lane),
widen road.
Remove existing pavement, increase road base and overlay to
remove frost restrictions, truck lanes where grade is greater than 4
degrees (minimum distance of 460 meters per lane), turnout lanes
every 32 kilometers (distance of 305 meters per lane), widen road.
Same as State Route 375 to U.S. 6.
Remove existing pavement on frost restricted portion, increase base
and overlay to remove frost restrictions, turnout lanes every 8
kilometers (distance of 305 meters per lane), construct bypass
around intersection at Beatty, bridge upgrade near Beatty.
Asphalt overlay on existing roads.
a. Source: TRW ( 1 999b, Heavy-Haul Truck Files, Item 4).
b. To convert meters to feet, multiply by 3.2808.
c. To convert kilometers to miles, multiply by 0.62137.
Table J-38. Potential road upgrades for Caliente-Chalk Mountain route."
Route
Upgrades
Intermodal transfer station to U.S. 93
U.S. 93 to State Route 375
State Route 375 to Rachel
Rachel to Nellis Air Force Range
Nellis Airforce Range Roads
Nevada Test Site Roads
Pave existing gravel road.
Asphalt overlay on existing pavement, truck lanes where grade is
greater than 4 percent (minimum distance 460 meters'" per lane),
tiunout lanes every 32 kilometers'^ (distance of 305 meters per
lane), widen road.
Remove existing pavement, increase road base and overlay to
remove frost restrictions, turnout lanes every 32 kilometers
(distance of 305 meters per lane), widen road.
Pave existing gravel road.
Rebuild existing road.
Asphalt overlay on existing roads.
a. Source: TRW (1999b, Heavy-Haul Truck Files, Item 9).
b. To convert meters to feet, multiply by 3.2808.
c. To convert kilometers to miles, multiply by 0.62137.
DOE has identified 0.4-kilometer (0.25-niile)-wide corridors along each route within which it would need
to obtain a right-of-way to construct a rail line and an associated access road. A corridor defines the
boundaries of the route by identifying an established "zone" for the location of the raikoad. For this
analysis, DOE identified a single alignment for each of the corridors. These single alignments are
representative of the range of alignments that DOE has considered for the corridors from engineering
design and construction viewpoints. The following paragraphs describe the alignments that have been
identified for the corridors. Before siting a branch rail line, DOE would conduct engineering studies in
each corridor to determine a specific alignment for the roadbed, track, and right-of-way for a branch rail
line.
Carlin Rail Corridor Implementing Altemative. The Carlin corridor originates at the Union Pacific
main line railroad near Beowawe in north-central Nevada. The corridor is about 520 kilometers (331
J-89
Transportation
Table J-39. Potential road upgrades for Caliente-Las Vegas route.'
Route
Upgrades
Intermodal transfer station to U.S. 93
U.S. 93 to Interstate 15
Interstate 15 to U.S. 95
U.S. 95 to Mercury
Mercury Exit to Yucca Mountain site
Pave existing gravel road.
Asphalt overlay on existing pavement, truck lanes where grade is
greater than 4 percent (minimum distance 460 meters'" per lane),
turnout lanes every 32 kilometers'^ (distance of 305 meters per
lane), widen road, rebuild Interstate 15 interchange.
Increase existing two-lane Las Vegas Beltway to four lanes, asphalt
overlay on U.S. 95.
Asphalt overlay on U.S. 95.
Asphalt overlay on Jackass Flats Road, rebuild road when required.
a. Source: TRW (1999b, Heavy-Haul Truck Files, Item 4).
b. To convert meters to feet, multiply by 3.2808.
c. To convert kilometers to miles, multiply by 0.62137.
Table J-40. Potential road upgrades for Apex/Dry Lake route.^
Route
Upgrades
Intermodal transfer station to Interstate 15
Rebuild frontage road to U.S. 93. Rebuild U.S. 93/lnterstate 15
interchange.
Increase existing two-lane Las Vegas Beltway to four lanes.
Asphalt overlay on U.S. 95.
Asphalt overlay on Jackass Flats Road, rebuild road when required.
Source: TRW (1999b, Heavy-Haul Truck Files, Item 4).
Table J-41. Potential road upgrades for Sloan/Jean route.^
Interstate 15 to U.S. 95
U.S. 95 to Mercury Exit
Mercury Exit to Yucca Mountain site
Route
Upgrades
Intermodal transfer station to Interstate 15
Interstate 15 to U.S. 95
U.S. 95 to Mercury Exit
Mercury Exit to Yucca Mountain site
Overlay and widen existing road to Interstate 15 interchange, rebuild
Interstate 15 interchange.
Increase existing two-lane Las Vegas Beltway to four lanes.
Asphalt overlay on U.S. 95.
Asphalt overlay on Jackass Flats Road, rebuild road when required.
a. Source: TRW (1999b, Heavy-Haul Truck Files, Item 4).
miles) long from the tie-in point with the Union Pacific line to the Yucca Mountain site. Table J-42 lists
possible variations in the alignment of this corridor.
Caliente Rail Corridor Implementing Alternative. The Caliente corridor originates at an existing
siding to the Union Pacific mainline railroad near Caliente, Nevada. The Caliente and Carlin corridors
converge near the northwest boundary of the Nellis Air Force Range. Past this point, they are identical.
The Caliente corridor would be 513 kilometers (320 miles) long from the Union Pacific line connection to j
the Yucca Mountain site. Table J-43 lists possible alignment variations for this corridor.
Caliente-Chalk Mountain Rail Corridor Implementing Alternative. The Caliente-Chalk Mountain
corridor is identical to the Caliente corridor until it approaches the northern boundary of the Nellis Air
Force Range. At this point the Caliente-Chalk Mountain corridor turns south through the Nellis Air Force 1
Range and the Nevada Test Site to the Yucca Mountain site. The corridor would be 345 kilometers (214
miles) long from the tie-in point at the Union Pacific line to the Yucca Mountain Site. Table J-44 lists
possible alignment variations for this corridor.
J-90
Transportation
Table J-42. Possible alignment variations of the Carlin corridor/
Corridor
Description
Crescent Valley Would diverge from the analyzed alignment near Cortez Mining Operation; would travel
through nonagricultural lands adjacent to alkali flats but would affect larger area of private
land.
Wood Spring Would diverge from the analyzed alignment and use continuous 2-percent grade to descend
from Dry Canyon Summit in Toiyabe range; would be shorter than the analyzed alignment
but would have steeper grade.
Rye Patch Would travel through Rye Patch Canyon, which has springs, riparian areas, and game
habitats; would divert from the analyzed alignment, maintaining distance of 420 meters'"
from Rye Patch Spring and at least 360 meters from riparian areas throughout Rye Patch
Canyon, except at crossing of riparian area near south end of canyon; would avoid game
habitat (sage grouse strutting area).
Steiner Creek Would diverge from the analyzed alignment at north end of Rye Patch Canyon. Would
avoid crossing private lands, two known hawk-nesting areas, and important game habitat
(sage grouse strutting area) in the analyzed alignment.
Monitor Valley Would travel through less populated Monitor Valley (in comparison to Big Smokey
Valley).
Mud Lake' Would travel farther from west edge of Mud Lake, which has known important
archaeological sites.
Goldfield*^ Would avoid crossing Nellis Air Force Range boundary near Goldfield, avoiding potential
land-use conflicts with Air Force.
Bonnie Claire*^ Would avoid crossing Nellis Air Force Range boundary near Scotty's Junction, avoiding
potential land-use conflicts with Air Force.
Oasis Valley"^ Would enable flexibility in crossing environmentally sensitive Oasis Valley area. If DOE
selected route through this area, fiirther studies would ensure small environmental impacts.
Beatty Wash' Would provide a corridor through Beatty Wash that was longer, but required less severe
earthwork than the analyzed alignment.
a. Source: TRW (1999b, Rail Files, Item 6).
b. To convert meters to feet, multiply by 3.2808.
c. Common with Caliente corridor.
Table J-43. Possible alignment variations of the Caliente corridor.'
Corridor
Description
Caliente''
Crestline''
White River
Garden Valley
Mud Lake'
Goldfield'
Bonnie Claire*^
Oasis Valley'
Beatty Wash'
I Source: TRW
b. Common with
c. Common with
Would connect with Union Pacific line at existing siding in Town of Caliente.
Would connect with Union Pacific line near east end of existing siding at Crestline.
Would avoid potential conflict with Weepah Spring Wilderness Study Area.
Would put more distance between rail corridor and private lands in Garden Valley and
Coal Valley.
Would travel farther from west edge of Mud Lake, which has known important
archaeological sites.
Would avoid crossing Nellis Air Force Range boundary near Goldfield, avoiding potential
land-use conflicts with Air Force.
Would avoid crossing Nellis Air Force Range boundary near Scotty's Junction, avoiding
potential land-use conflicts with Air Force.
Would enable flexibility in crossing environmentally sensitive Oasis Valley area. If DOE
selected route through this area, further studies would ensure small environmental impacts.
Would provide corridor through Beatty Wash that was longer, but required less severe
earthwork than the analyzed alignment.
(1999b, Rail Files, Item 6).
Caliente-Chalk Mountain corridor.
Carhn corridor.
J-91
Transportation
Table J-44. Possible alignment variations of the Caliente-Chalk Mountain corridor/
Corridor Description
Mercury Highway
Tonopah
Mine Mountain
Area 4
To provide flexibility in choosing path, would travel north through center of Nevada
Test Site.
To provide flexibility in choosing path through Nevada Test Site; would travel north
along western boundary of Nevada Test Site.
Would provide flexibility in minimizing impacts to local archaeological sites.
Would provide flexibility in choosing path through Nevada Test Site.
a. Source: TRW (1999b, Rail Files, Item 8).
Jean Rail Corridor Implementing Alternative. The Jean corridor originates at the existing Union
Pacific mainline railroad near Jean, Nevada. The corridor would be 181 kilometers (112 miles) long from
the tie-in point at the Union Pacific line to the Yucca Mountain site. Table J-45 lists possible variations
for this corridor.
Table J-45. Possible alignment variations of the Jean corridor.'
Corridor
Description
North Pahrump Would minimize impacts to approximately 4 kilometers'" of private land on northeast
side of Pahrump.
Stateline Pass Would provide option to crossing Spring Mountains at Wilson Pass; would diverge
from analyzed alignment in Pahrump Valley; would parallel Nevada-California border,
traveling along southwestern edge of Spring Mountains and crossing border twice.
a. Source: TRW (1999b, Rail Files, Item 6).
b. 4 kilometers = 2.5 miles (approximate).
Valley Modified Rail Corridor Implementing Alternative. The Valley Modified corridor originates at
an existing rail siding off the Union Pacific mainline railroad northeast of Las Vegas. The corridor is
about 159 kilometers (98 miles) long from the tie-in point with the Union Pacific line to the Yucca
Mountain site. Table J-46 lists the possible variations in alignment for this corridor.
Table J-46. Possible alignment variations of the Valley Modified corridor.'
Corridor
Description
Indian Hills
Sheep Mountain
Valley Connection
Would avoid entrance to Nellis Air Force Range north of Town of Indian Springs by
traveling south of town.
Would increase distance from private land in Las Vegas and proposed 30-square-
kilometer'' Bureau of Land Management land exchange with city.
Would locate transfer operations at Union Pacific Valley Yard rather than Dike siding.
Overflights of Dike siding from Nellis Air Force Base could conflict with switching
operations.
a. Source: TRW (1999b, Rail Files, Item 6).
b. 30 square kilometers = 7,410 acres (approximate).
J.3.1.3 Sensitivity of Analysis Results to Routing Assumptions
hi addition to analyzing the impacts of using highway routes that would meet Department of
Transportation requirements for transporting spent nuclear fuel, DOE evaluated how the estimated
impacts would differ if legal-weight trucks used other routes in Nevada. Six other routes identified in a
1989 study by the Nevada Department of Transportation (Ardila-Coulson 1989, pages 36 and 45) were
J-92
Transportation
selected for this analysis. The Nevada Department of Transportation study described the routes as
follows:
Route A. Minimum distance and minimum accident rate.
South on U.S. 93A, south on U.S. 93, west on U.S. 6, south on Nevada 318, south on U.S. 93, south
on 1-15, west on Craig Road, north on U.S. 95
Route B. Minimum population density and minimum truck accident rate.
South on U.S. 93A, south on U.S. 93, west on U.S. 6, south on U.S. 95.
Both of these two routes use the U.S. 6 truck bypass in Ely.
Alternative route possibilities were identified between I- 1 5 at Baker, California and l-AO at Needles,
California to Mercury. These alternative routes depend upon the use of U.S. 95 in California, California
127 and the Nipton Road.
Route C. From Baker with California 127.
North on California 127, north on Nevada 373, south on U.S. 95
Route D. From Baker without California 127.
North on 1-15, west on Nevada 160, south on U.S. 95
Route E. From Needles with U.S. 95, California 127, and the Nipton Road.
North on U.S. 95, west on Nevada 164, west on 1-15, north on California 127, north on Nevada 373,
south on U.S. 95
Route F. From Needles without California 127 and the Nipton Road.
West on 1-40, east on 1-15, west on Nevada 160, south on U.S. 95
Table J-47 identifies the sensitivity cases evaluated based on the Nevada Department of Transportation
routes. Table J-48 lists the range of impacts in Nevada of using these different routes for the mostly
legal-weight truck analysis scenario. The tables compare the impacts estimated for the highways
identified in the Nevada study to those estimated for shipments that would follow routes allowed by
current Department of Transportation regulations for Highway Route-Controlled Quantities of
Radioactive Materials. Because the State of Nevada has not designated alternative or additional preferred
routes for use by these shipments, as permitted under Department of Transportation regulations (49 CFR
397.103), DOE has assumed that shipments of spent nuclear fuel and high-level radioactive waste would
Table J-47. Nevada routing sensitivity cases analyzed for a legal-weight truck.
Case Description
Case 1 To Yucca Mountain via Barstow, California, using 1-15 to Nevada 160 to Nevada 160 (Nevada D and
F)
Case 2 To Yucca Mountain via Barstow using 1-15 to California route 127 to Nevada 373 to US 95 (Nevada
C)
Case 3 To Yucca Mountain via Needles using U.S. 95 to Nevada 164 to 1-15 to California 127 to Nevada 373
and U.S. 95 (Nevada E)
Case 4 To Yucca Mountain via Needles using U.S. 95 to Nevada 164 to 1-15 to Nevada 160 (variation of
Nevada E)
Case 5 To Yucca Mountain via Wendover using U.S. 93 Alternate to U.S. 93 to US 6 to U.S. 95 (Nevada B)
Case 6 To Yucca Mountain via Wendover using U.S. 93 Alternate to U.S. 93 to Nevada 3 1 8 to U.S. 93 to
1-15 to the Las Vegas Beltway to U.S. 95 (Nevada A)
J-93
Transportation
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J-94
Transportation
enter Nevada on 1-15 from either the northeast or southwest. The analysis assumed that shipments
traveling on 1-15 from the northeast would use the northern Las Vegas Beltway to connect to U.S. 95 and
continue to the Nevada Test Site. Shipments from the southwest on 1-15 would use the southern and
western Las Vegas Beltway to connect to U.S. 95 and continue to the Nevada Test Site.
J.3.2 ANALYSIS OF INCIDENT-FREE TRANSPORTATION IN NEVADA
The analysis of incident-free impacts to populations in Nevada addressed transportation through urban,
suburban, and rural population zones. The population densities that were assumed for the analysis were
determined using the HIGHWAY and DVTERLINE computer programs. The population in the 800-meter
(0.5-mile) region of influence used to evaluate the impacts of incident-free transportation for both legal-
weight truck and rail shipments is listed in Table J-36.
Results for incident-free transportation of spent nuclear fuel and high-level radioactive waste for
Inventory Modules 1 and 2 are presented in Section J. 3.4.
J.3.3 ANALYSIS OF TRANSPORTATION ACCIDENT SCENARIOS IN NEVADA
Section J. 1 .4 discusses the methodology for estimating the risks of accidents that could occur during rail
and truck transportation of spent nuclear fuel and high-level radioactive waste. Section J.3.5 describes the
results of the accident risk analysis for Inventory Modules 1 and 2.
J.3.3.1 Intermodal Transfer Station Accident Methodology
Shipping casks would arrive at an intermodal transfer station in Nevada by rail, and a gantry crane would
transfer them from the railcars to heavy-haul trucks for transportation to the repository. The casks, which
would not be opened or altered in any way at the intermodal transfer station, would be certified by the
Nuclear Regulatory Commission and would be designed for accident conditions specified in 10 CFR
Part 71. Impact limiters, which would protect casks against collisions during transportation, would
remain in place during transfer operations at the intermodal transfer station.
DOE performed an accident screening process to identify credible accidents that could occur at an
intermodal transfer station with the potential for compromising the integrity of the casks and releasing
radioactive material. The external events listed in Table J-49 were considered, along with an evaluation
of their potential applicability.
As indicated from Table J-49, the only accident-initiating event identified from among the feasible
external events was the aircraft crash. Such events would be credible only for casks being handled or on
transport vehicles at an intermodal transfer station in the Las Vegas area (Apex/Dry Lake or Sloan/Jean).
For a station in the Las Vegas area, an aircraft crash would be from either commercial aircraft operations
at McCarran airport or military operations from Nellis Air Force Base.
Among the internal events, the only potential accident identified was a drop of the cask during transfer
operations. This accident would bound the other events considered, including drops from the railcar or
truck (less fall height would be involved than during the transfer operations). Collisions, derailments, and
other accidents involving the transport vehicles at the intermodal transfer would not damage the casks due
to the requirement that they be able to withstand high-speed impacts and the low velocities of the
transport vehicles at the intermodal transfer station.
Sabotage events were also considered as potential accident-initiating events at an intermodal transfer
station. Section J. 1.5 evaluates such events.
J-95
Transportation
Table J-49. Screening analysis of external events considered potential
accident initiators at intermodal transfer station.
Event
Applicability
Aircraft crash
Retained for further evaluation
Avalanche
(a)
Coastal erosion
(a)
Dam failure
See flooding
Debris avalanching
(a)
Dissolution
(b)
Epeirogenic displacement
(tilting of the earth's crust)
(c)
Erosion
(b)
Extreme wind
(c)
Extreme weather
(e)
Fire (range)
(b)
Hooding
(d)
Denudation
(b)
Fungus, bacteria, algae
(b)
Glacial erosion
(b)
High lake level
(b)
High tide
(a)
High river stage
See flooding
Hurricane
(a)
Inadvertent future intrusion
(b)
Industrial activity
Bounded by aircraft crash
Intentional future intrusion
(b)
Lightning
(c)
Loss of off/on site power
(c)
Low lake level
(b)
Meteorite impact
(e)
Military activity
Retained for further evaluation
Oogenic diastrophism
(e)
Pipeline accident
(b)
Rainstorm
See flooding
Sandstorm
(c)
Sedimentation
(b)
Seiche
(a)
Seismic activity, uphfting
(c)
Seismic activity, earthquake
(c)
Seismic activity, surface fault
(c)
Seismic activity, subsurface fault
(c)
Static fracturing
(b)
Stream erosion
(b)
Subsidence
(c)
Tornado
(c)
Tsunami
(a)
Undetected past intrusions
(b)
Undetected geologic features
(b)
Undetected geologic processes
(c)
Volcanic eruption
(e)
Volcanism, magmatic activity
(e)
Volcanism, ash flow
(c)
Volcanism, ash fall
(b)
Waves (aquatic)
(a)
a. Conditions at proposed sites do not allow event.
b. Not a potential accident initiator.
c. Bounded by cask drop accident considered in the internal events analysis.
d. Shipping cask designed for event.
e. Not credible, see evaluation for repository.
J-96
Transportation
Accident Analysis
1 . Cask Drop Accident. The only internal event retained after the screening process was a failure of
the gantry crane (due to mechanical failure or human error) during the transfer of a shipping cask
from a railcar to a heavy-haul truck. The maximum height between the shipping cask and the ground
during the transfer operation would be less than 6 meters (19 feet) (TRW 1999a, Heavy-Haul Files,
Item 11). The casks would be designed to withstand a 9-meter (30-foot) drop. Therefore, the cask
would be unlikely to fail during the event, especially because the impact energy from the 6-meter
drop would be only 65 percent of the minimum design requirement.
2. Aircraft Crash Accident. Two of the three intermodal transfer station locations are near airports that
handle large volumes of air traffic. The Apex/Dry Lake location is about 16 kilometers (10 miles)
northeast of the Nellis Air Force Base runways. Between 60,(KX) and 67,000 takeoffs and landings
occur at Nellis Air Force Base each year (Luedke 1997, all). The Sloan/Jean intermodal transfer area
begins about 16 kilometers southwest of McCarran International Airport in Las Vegas. In 1996,
McCarran had an average of 1,300 daily aircraft operations (Best 1998, all). Because of the large
number of aircraft operations at these airports, the probability of an aircraft crash on the proposed
intermodal transfer station could be within the credible range. To assess the consequences of an
aircraft crash, an analysis evaluated the ability of large aircraft projectiles [jet engines and jet engine
shafts (DOE 1996b, page 58)] to penetrate the shipping casks. The analysis used a recommended
formula (DOE 1996b, page 69) for predicting the penetration of steel targets, as follows:
T'^= 0.5 X M X V^h- 17,400 x KjX D' ^
where:
T = predicted thickness to just perforate a steel plate (inches)
M = projectile mass (weight/gravitational acceleration)
V = projectile impact velocity (feet per second)
Kj = constant depending on the grade of steel (usually about 1.0)
D = projectile diameter (inches)
The projectile characteristics listed in Table J-50 are from Davis, Strenge, and Mishima (1998, all). The
velocity used is about 130 meters (427 feet) per second, which is representative of aircraft velocities near
airports (maximum velocity during takeoff and landing operations). A higher velocity [about 180 meters
(590 feet) per second] was assumed for the projectile found to be limiting in terms of ability to penetrate
(commercial engine shaft) to provide perspective on the influence of velocity on the penetration
thickness. Table J-5 1 lists the results of the penetration calculation.
Table J-50. Projectile characteristics."
Engine weight Engine diameter
Aircraft (kilograms)'' (centimeters)*^
Small military 420 71
Commercial 3,900 270
a. Source: Davis, Strenge, and Mishima (1998, Table 1).
b. To convert kilograms to pounds, multiply by 2.2046.
c. To convert centimeters to inches, multiply by 0.3937.
The results indicate that none of the aircraft projectiles considered would penetrate the shipping casks,
which would have metal shield walls about 18 centimeters (7 inches) thick (JAI 1996, all).
This evaluation found no credible accidents with the potential for radioactive release at an intermodal
transfer station.
J-97
Transportation
Velocity
Penetration thickness
(meters per second)''
(centimeters)''''
130
2.5
130
2.5
130
3.0
130
3.7
180
5.9
Table J-51. Results of aircraft projectile penetration analysis.
Projectile
Small military engine
Small military shaft
Commercial engine
Commercial shaft
Commercial shaft
a. Source: Davis, Strenge, and Mishima (1998, Table 2).
b. To convert meters to feet, multiply by 3.2808.
c. To convert centimeters to inches, multiply by 0.3937.
d. Penetration through steel plate.
J.3.4 IMPACTS IN NEVADA FROM INCIDENT-FREE TRANSPORTATION FOR INVENTORY
MODULES 1 AND 2
This section presents the analysis of impacts to occupational and public health and safety in Nevada from
incident-free transportation of spent nuclear fuel and high-level radioactive waste in Inventory Modules 1
and 2. The analysis assumed that the routes, population densities, and shipment characteristics (for
example, radiation ft-om shipping casks) for shipments under the Proposed Action and hiventory Modules I
1 and 2 would be the same. The only difference was the projected number of shipments that would travelj
to the repository.
The following sections provide detailed information on the range of potential impacts to occupational andj
public safety and health from incident-free transportation of Modules 1 and 2 that result from legal-
weight trucks and the 10 alternative transportation routes considered in Nevada. National impacts of
incident-free transportation of Modules 1 and 2 incorporating Nevada impacts are discussed together with
other cumulative impacts in Chapter 8.
J.3.4.1 Mostly Legal-Weight Truck Scenario
Tables J-52 and J-53 list estimated incident-free impacts in Nevada for the mostly legal-weight truck
scenario for shipments of materials included in Inventory Modules 1 and 2.
J.3.4.2 Nevada Rail Implementing Alternatives
Table J-54 lists the range of estimated incident-free impacts in Nevada for the operation of a branch rail
line to ship the materials included in Inventory Modules 1 and 2. It lists impacts that would result from
operations for a branch line in each of the five possible rail corridors DOE is evaluating. These include
the impacts of about 2,600 legal-weight truck shipments from commercial sites that could not use rail
casks to ship spent nuclear fuel.
J.3.4.3 Nevada Heavy-Haul Truck Implementing Alternatives
Radiological Impacts
Intermodal Transfer Station Impacts. Involved worker exposures (the analysis assumed that the
noninvolved workers would receive no radiation exposure and thus required no further analysis) would
occur during both inbound (to the repository) and outbound (to the 77 sites) portions of the shipment
campaign. DOE used the same involved worker level of effort it used in the analysis of intermodal
transfer station worker industrial safety impacts to estimate collective involved worker radiological
impacts (that is, 16 full-time equivalents per year). The collective worker radiation doses were adapted
from a study (Smith, Daling and Faletti 1992, all) of a spent nuclear fuel transportation system, which
J-98
Transportation
Table J-52. Population doses and radiological impacts from incident-free Nevada transportation for
mostly legal-weight truck scenario - Modules 1 and 2/
Category
Legal-weight Rail shipments of naval
truck shipments spent nuclear fuel''
Total'
Module 1
Involved worker
Collective dose (person-rem)
Estimated latent cancer fatalities
Public
Collective dose (person-rem)
Estimated latent cancer fatalities
Module 2
Involved worker
Collective dose (person-rem)
Estimated latent cancer fatalities
Public
Collective dose (person-rem)
Estimated latent cancer fatalities
2,900
30
2,900
1.2
0.01
1.2
5,100
26
5,100
2.5
0.01
2.5
3,000
40
3,000
1.2
0.02
1.2
5,300
30
5,300
2.6
0.02
2.6
a. Impacts are totals for shipments over 38 years.
b. Includes impacts at intermodal transfer stations.
c. Totals might differ from sums due to rounding.
Table J-53. Population health impacts from vehicle emissions during incident-free Nevada transportation
for the mostly legal-weight truck scenario - Modules 1 and 2.'
Vehicle emission-related fatalities
Legal-weight
truck shipments
Rail shipments of naval
spent nuclear fiiel''
Total'
Module 1
Module 2
0.01
0.01
0.0004
0.0005
0.01
0.01
a. Impacts are totals for shipments over 38 years.
b. Includes heavy-haul truck shipments in Nevada.
c. Totals might differ from sums due to rounding.
Table J-54. Radiological and noru-adiological inopacts from incident-free Nevada transportation for the
mostly rail scenario - Modules 1 and 2.'
Category
Legal-weight
truck shipments
Rail shipments
Total"
Module 1
Involved worker
Collective dose (person-rem)
Estimated latent cancer fatalities
Public
Collective dose (person-rem)
Estimated latent cancer fatalities
Estimated vehicle emission-related fatalities
370
280 - 460
650 - 830
0.15
0.11-0.18
0.26 - 0.33
430
190 - 270
620 -700
0.22
0.09-0.14
0.31-0.36
0.00019
0.004
0.0042
a. Impacts are totals for 38 years (2010 to 2048).
b. Totals might differ from sums due to rounding.
was also performed for the commercial sites. That study found that the collective worker doses that could
be incurred during similar inbound and outbound transfer operations of a single loaded (with commercial
spent nuclear fuel) and unloaded cask were approximately 0.027 and 0.001 person-rem per cask,
respectively, as listed in Table J-55.
The analysis used these inbound and outbound collective dose factors to calculate the involved worker
impacts listed in Table J-56 for Module 1 and Module 2 inventories in the same manner it used for
J-99
Transportation
Table J-55. Collective worker doses (person-rem) from transportation of a single cask.^''
Inbound
Inbound
CD"
Outbound
Outbound
CD
Receive transport vehicle and loaded cask. 6.3x10"'
Monitor, inspect, unhook offsite drive unit,
and attach onsite drive unit.
Move cask to parking area and wait for 1 .4x10"'
wash down station. Attach to carrier puller
when ready.
Move cask to receiving and handling area. 9.2x10"'
Remove cask from carrier and place on 4.3x10"'
cask cart.
Connect onsite drive unit and move cask to 7.0x10"'*
inspection area; disconnect onsite drive
unit.
Hook up offsite drive unit, move to 1.4x10'^
gatehouse, perform final monitoring and
inspection of cask.
Notify appropriate organizations of the 0.0
shipment's departure.
Total 2.7x10
:-5
Receive transport vehicle and empty cask. 0.0
Monitor, inspect, unhook offsite drive unit,
and attach onsite drive unit.
Move cask to parking area and wait for 5.4x10"^
wash down station. Attach to carrier puller
when ready.
Move cask to receiving and handling area. 8.0x10"'
Remove cask from carrier and place on 2.2x10"*
cask cart.
Connect onsite drive unit and move cask to 3.3x10"'
inspection area; disconnect onsite drive
unit.
Hook up offsite drive unit, move to 8.3x10"'
gatehouse, perform final monitoring and
inspection of cask.
Notify appropriate organizations of the 0.0
shipment's departure.
Total 8.8x10'
a. Adapted from Smith, Daling and Faletti (1992, Table 4.2).
b. Values are rounded to two significant figures; therefore, totals might differ from sums of values.
c. CD = collective dose (person-millirem per cask).
Table J-56. Doses and radiological
operations - Modules 1 and 2.^^
health impacts to involved workers from intermodal transfer station
Module 1
Module 2
Group
Dose
Latent cancer fatality Dose
Latent cancer fatality
Maximally exposed individual worker*^
Involved worker population**
12
530
0.005 12
0.21 550
0.005
0.22
b.
c.
d.
Includes estimated impacts from handling 300 shipments of U.S. Navy fuel that would be shipped by rail under the mostly
legal-weight tmck transportation scenario. DOE estimated the impacts from these shipments by adjusting the impacts from
the approximately 19,300 shipments (9,650 x 2) that would pass through the intermodal transfer station under the mostly i
scenario.
Totals for 24 years of operations.
The estimated probability of a latent cancer fatality in an exposed individual.
The estimated number of latent cancer fatalities in an exposed involved worker population.
commercial power reactor spent nuclear fiiel impacts. The number of inbound and outbound shipments
for Module 1 and Module 2 inventories is from Section J. 1.2. The worker impacts reflect two-way
operations.
Incident-Free Transportation. Table J-57 lists the range of estimated incident-free impacts in Nevada
for the use of heavy-haul trucks to ship the materials included in Inventory Modules 1 and 2. It lists
impacts that would result from operations on each of the five possible highway routes in Nevada DOE is
evaluating. These include impacts of about 2,600 legal-weight truck shipments from commercial sites
that could not ship spent nuclear fuel using rail casks.
J-100
Transportation
Table J-57. Radiological and nonradiological health impacts from incident-free transportation for the
heavy-haul truck implementing alternatives - Modules 1 and 2.'
Legal-weight truck
Rail and heavy-haul
Category
shipments
truck shipments'"
Total'
Involved worker
Collective dose (person-rem)
370
830-1,000
1,200-1,400
Estimated latent cancer fatalities
0.15
0.33 - 0.40
0.48 - 0.55
Public
Collective dose (person-rem)
430
1,200-3,200
1,600-3,700
Estimated latent cancer fatalities
0.22
0.60-1.6
0.82-1.8
Estimated vehicle emission-related fatalities
0.00019
0.03
0.05
a. Impacts are totals for 38 years (2010 to 2048).
b. Includes impacts to workers at an intermodal transfer station.
c. Totals might differ from sums due to rounding.
J.3.5 IMPACTS IN NEVADA FROM TRANSPORTATION ACCIDENTS FOR INVENTORY
MODULES 1 AND 2
The analysis assumed that the routes, population densities, and shipment characteristics (for example,
assumed radioactive material contents of shipping casks) for the Proposed Action and Inventory Modules
1 and 2 would be the same. The only difference would be the projected number of shipments that would
travel to the repository. As listed in Table J-1, Module 2 would include about 3 percent more shipments
than Module 1.
J.3.5.1 Mostly Legal-Weight Truck Scenario
Radiological Impacts
The analysis estimated the radiological impacts of accidents in Nevada for the mostly legal-weight truck
scenario for shipments of the materials included in Inventory Modules 1 and 2. The radiological health
impacts associated with Module 1 would be 0.86 person-rem and for Module 2 would be 0.88 person-rem
(see Table J-58). These impacts would occur over 34 years in a population of more than 1 million people
who lived within 80 kilometers (50 miles) of the Nevada routes that DOE would use. This dose risk
would lead to about 1 chance in 1,(XX) of an additional cancer fatality in the exposed population. For
comparison, about 220,(XX) in a population of 1 million people would suffer fatal cancers from other
causes (ACS 1998, page 10).
Traffic Fatalities
The analysis estimated traffic fatalities from accidents involving the transport of spent nuclear fuel and
high-level radioactive waste by legal-weight trucks in Nevada for the mostly legal-weight truck scenario
for shipments of the materials included in Inventory Modules 1 and 2. It estimated that there would be
0.9 fatality over 34 years for Module 1 and 0.93 fatality for Module 2 (see Table J-58). The estimate of
traffic fatalities includes the risk of fatalities from 3(X) shipments of naval spent nuclear fuel.
J.3.5.2 Nevada Rail Implementing Alternatives
Industrial Safety Impacts
Table J-59 lists the estimated industrial safety impacts in Nevada for the operation of a branch rail line to
ship the materials included in Inventory Modules 1 and 2. The table lists impacts that would result from
operations for a branch line in each of the five possible rail corridors in Nevada that DOE is evaluating.
The representative workplace loss incidence rate for each impact parameter (as compiled by the Bureau of
Labor Statistics) was used as a multiplier to convert the operations crew level of effort to expected
J-101
Transportation
Table J-58. Accident radiological
health impacts
for Modules 1 and 2 -
- Nevada transportation."
Dose risk
(person-
Latent cancer
Traffic
Transportation scenario
rem)
fatalities
fatalities
Legal-weight truck
0.88"
0.0004
0.9
Legal-weight truck for the mostly rail
scenario
0.1
0.00006
0.1
Mostly rail (Nevada rail implementing alternatives)
Caliente
0.02
8.7x10'
0.13
Carlin
0.03
1.6x10"'
0.17
Sloan/Jean
0.11
5.3x10"'
0.10
Af)ex/Dry Lake
0.01
7.0x10"*
0.08
Caliente-Chalk Mountain
0.01
e.QxlO"*
0.09
Mostly rail (Nevada heavy-haul implementing alternatives)
Caliente
0.34
1.7x10-*
1.2
Caliente-Chalk Mountain
0.28
1.4x10-''
0.65
Caliente-Las Vegas
1.02
5.1x10"*
0.90
Apex/Dry Lake
0.94
4.7x10"*
0.46
Jean
6.5
3.2x10"'
0.49
a. Impacts over 38 years.
b. Estimates of dose risk are for the transportation of the materials included in
Module 2. Estimates of dose risk for
transportation of the materials in Module 1 would be slightly (about 3 percent) lower.
Table J-59. Rail corridor operation worker physical trauma impacts (Modules 1 and 2).
Worker group and _
Corridor
impact category
Caliente
Carlin
Chalk Mountain Jean
Valley Modified
Involved workers
TRC
200
200
200
150
150
LWC"
110
110
110
82
82
Fatalities
0.4
0.4
0.4
0.3
0.3
Noninvolved workers'^
TRC
9
9
9
7
7
LWC
5
5
5
3
3
Fatalities
0.01
0.01
0.01
0.01
0.01
All workers (totals)''
TRC
210
210
210
160
160
LWC
120
120
120
85
85
Fatalities
0.4
0.4
0.4
0.3
0.3
Traffic fatalities'
1.1
1.1
1.1
0.8
0.8
a. TRC = total recordable cases (injury and illness).
b. LWC = lost workday cases.
c. Noninvolved worker impacts are based on 25 percent of the involved worker level of effort.
d. Totals might differ from sums due to rounding.
e. Fatalities from accidents during commutes to and from jobs for involved and noninvolved workers.
industrial safety losses. The involved worker full-time equivalent multiples that DOE would assign to
operate each rail corridor each year was estimated to be 36 to 47 full-time equivalents, depending on the
corridor for the period of operations (scaled from cost data in TRW 1996, Appendix E). Noninvolved
worker full-time equivalent multiples were unavailable, so DOE assumed that the noninvolved worker
level of effort would be similar to that for the repository operations work force — about 25 percent of that
for involved workers. The Bureau of Labor Statistics loss incidence rate for each total recordable case,
lost workday, and fatality trauma category (for example, the number of total recordable cases per
full-time equivalent) was multiplied by the involved and noninvolved worker full-time equivalent
multiples to project the associated trauma incidence.
J- 102
Transportation
The involved worker total recordable case incidence rate, 170,000 total recordable cases in a workforce of
1,620,000 workers (0.1 1 total recordable case per full-time equivalent) reflects losses in the Trucking and
Warehousing sector during 1996. The same Bureau of Labor Statistics period of record and industry
sector was used to select the involved worker lost workday case incidence rate [96,000 lost workday cases
in a workforce of 1,620,000 workers (0.06 lost workday case per full-time equivalent)]. The involved
worker fatality incidence rate, 22 fatalities in a workforce of 1(X),000 workers (0.0(X)2 fatality per full-
time equivalent) reflects losses in the Transportation and Material Moving Occupations sector during the
Bureau of Labor Statistics 1994-to-1995 period of record.
The noninvolved worker incidence rate of 53,(X)0 total recordable cases in a workforce of 2,870,(XX)
workers (0.02 total recordable case per full-time equivalent) reflects losses in the Engineering and
Management Services sector during the Bureau of Labor Statistics 1996 period of record. DOE used the
same period of record and industry sector to select the noninvolved worker lost workday case incidence
rate [22,(XX) lost workday cases in a workforce of 2,870,(XX) workers (0.01 lost workday case per full-time
equivalent)]. The noninvolved worker fatality incidence rate, 1.5 fatalities in a workforce of 100,(XX)
workers (0.00002 fatality per full-time equivalent) reflects losses in the Managerial and Professional
Specialties sector during the 1994-to-1995 period of record.
Table J-59 lists the results of these industrial safety calculations for the five candidate corridors under
Inventory Modules 1 and 2. The table also lists estimates of the number of traffic fatalities that would
occur in the course of commuting by workers to and from their construction and operations jobs. These
estimates used national statistics for average commute distances [18.5 kilometers (11.5 miles) one-way
(ORNL 1999, all)] and fatality rates for automobile traffic [1 per 1(X) million kilometers (1.5 per
100 million miles) (BTS 1998, all)].
Radiological Impacts of Accidents
The analysis estimated the radiological impacts of accident scenarios in Nevada for the Nevada rail
implementing alternatives for shipments of the materials included in Inventory Modules 1 and 2. Table
J-58 lists the radiological dose-risk and associated risk of latent cancer fatalities. The risks include
accident risks in Nevada from approximately 2,6(X) legal-weight truck shipments from commercial sites
that could not ship spent nuclear fuel in rail casks. The risks would occur over 34 years.
Traffic Fatalities
Traffic fatalities from accidents involving transport of spent nuclear fiiel and high-level radioactive waste
by rail in Nevada were estimated for the Nevada rail implementing alternatives for shipments of materials
included in Inventory Modules 1 and 2. Table J-58 lists the estimated number of fatalities that would
occur over 34 years for a branch rail line along each of the five possible rail corridors. These estimates
include the risk of fatalities from about 2,6(X) legal-weight truck shipments from commercial generators
that could not ship spent nuclear fuel in rail casks.
J.3.5.3 Nevada Heavy-Haul Truck Implementing Alternatives
Industrial Safety Impacts
Tables J-60 and J-61 list the estimated industrial safety impacts in Nevada for operations of heavy -haul
trucks (principally highway maintenance safety impacts) and operation of an intermodal transfer station
that would transfer loaded and unloaded rail casks between rail cars and heavy-haul trucks for shipments
of the materials included in Inventory Modules 1 and 2. Table J-60 lists the estimated industrial safety
impacts in Nevada for the operation of a heavy-haul route to the Yucca Mountain site. Table J-61 lists
impacts that would result from the operation of an intermodal transfer station for any of the five possible
routes DOE is evaluating that heavy-haul trucks could use in Nevada.
J- 103
Transportation
Table J-60. Industrial health
impacts from
heavy-haul truck route operations (Modules 1 and 2).
Corridor
Worker group and
Caliente-Chalk
Caliente-
Sloan/
impact category
Caliente
Mountain
Las Vegas
Jean
Apex/Dry Lake
Involved workers
TRC
460
460
420
250
250
LWC""
250
250
230
140
140
Fatalities
0.8
0.8
0.8
0.5
0.5
Noninvolved workers'^
TRC
21
21
19
11
11
LWC
11
11
10
6
6
Fatalities
0.02
0.02
0.02
0.01
0.01
All workers (totals)
TRC
480
480
440
260
260
LWC
260
260
240
150
150
Fatalities
0.82
0.82
0.82
0.5
0.5
Traffic fatalities"
2.0
2.0
1.9
1.3
1.3
a. TRC = total recordable cases (injury and illness).
b. LWC = lost workday cases.
c. Noninvolved worker impacts are based on 25 percent of the involved worker level of effort.
d. Totals might differ from sums due to rounding.
e. Fatalities from accidents during commutes to and from jobs for involved and noninvolved workers.
Table J-61. Annual physical trauma impacts to workers from intermodal transfer station operations
(Module 1 or 2).
Involved workers
Noninvolved workers*
All workers
TRC" LWC Fatalities
TRC LWC Fatalities
TRC
LWC Fatalities
112 60 0.2
5 2 0.0
116
62 0.2
a. The noninvolved worker impacts are based on 25 percent of the involved worker level of effort.
b. TRC = total recordable cases of injury and illness.
c. LWC = lost workday cases.
Radiological Impacts of Accidents
The analysis estimated the radiological impacts of accidents in Nevada for the Nevada heavy -haul truck
implementing alternatives for shipments of the materials included in Inventory Modules 1 and 2.
Table J-58 lists the radiological dose-risk and associated risk of latent cancer fatalities. The risks include
accident risks in Nevada from approximately 2,600 legal-weight truck shipments from commercial
generating sites that could not ship spent nuclear fuel in rail casks. The risk would occur over 34 years.
Traffic Fatalities
The analysis estimated traffic fatalities from accidents involving the transport of spent nuclear fuel and
high-level radioactive waste (including the rail portion of transportation to and from an intermodal
transfer station) in Nevada for the heavy-haul truck implementing alternatives for shipments of the
materials included in Inventory Modules 1 and 2. Table J-58 lists the estimated number of fatalities that
would occur over 34 years for a branch rail line and for each of the five possible routes for heavy -haul
trucks. The estimate for traffic fatalities includes the risk of fatalities from about 2,600 legal-weight truck
shipments from commercial generators that could not ship spent nuclear fuel in rail casks.
J-104
Transportation
J.3.6 IMPACTS FROM TRANSPORTATION OF OTHER MATERIALS
Other types of transportation activities associated with the Proposed Action would involve shipments of
materials other than the spent nuclear fuel and high-level radioactive waste discussed in previous sections.
These activities would include the transportation of people. This section evaluates occupational and
public health and safety and air quality impacts from the shipment of:
• Construction materials, consumables, and personnel for repository construction and operation,
including disposal containers
Waste including low-level waste, construction and demolition debris, sanitary and industrial solid
waste, and hazardous waste
• Office and laboratory supplies, mail, and laboratory samples
The analysis includes potential impacts of transporting these materials for the case in which DOE would
not build a rail line to the proposed repository, because the larger number of truck shipments would lead
to higher impacts than those for rail shipments, as discussed above. In addition, because the construction
schedule for a new rail line would coincide with the schedule for the construction of repository facilities,
trucks would deliver materials for repository construction.
Rail service would benefit the delivery of 10,000 disposal containers from manufacturers. Two 33,000-
kilogram (about 75,000-pound) disposal containers and their 700-kilogram (about l,5(X)-pound) lids
(TRW 1999b, Request #027) would be delivered on a railcar — a total of 5,000 railcar deliveries over the
24-year period of the Proposed Action. These containers would be delivered to the repository along with
shipments of spent nuclear fuel and high-level radioactive waste or separately on supply trains along with
shipments of materials and equipment.
If rail service was not available, disposal container components that would weigh as much as 34 metric
tons (37.5 tons) would be transported to Nevada by rail and transferred to overweight trucks for shipment
to the repository site. In this event, 10,000 overweight truck shipments would move the containers from a
railhead to the site. The State of Nevada routinely provides permits to motor carriers for overweight,
overdimension loads if the gross vehicle weight does not exceed 58.5 metric tons (64.5 tons) (TRW
1999b, Request #046).
J.3.6.1 Transportation of Personnel and Materials to Repository
The following paragraphs describe impacts that would result from the transportation of construction
materials, consumables, disposal containers, supplies, mail, laboratory samples, and personnel to the
repository site during the construction, operation and monitoring, and closure phases.
Human Health and Safety
Most construction materials, construction equipment, and consumables would be transported to the Yucca
Mountain site on legal-weight trucks. Heavy and overdimensional construction equipment would be
delivered by trucks under permits issued by the Nevada Department of Transportation. DOE estimates
that about 42,000 truck shipments over 5 years would be necessary to transport materials, supplies, and
equipment to the site during the construction phase.
In addition to construction materials, supplies, equipment, and disposal containers, trucks would deliver
consumables to the repository site. These would include diesel fuel, cement, and other materials that
would be consumed in daily operations. About 13,000 semitrailer truck shipments would occur during
J-105
Transportation
each year of operation. Similarly, there would be an estimated 1,000 semitrailer truck shipments during
each year of monitoring and 1,200 each year during closure operations.
Over the 24-year period of the Proposed Action, the repository would receive about 300,000 truck
shipments of supplies, materials, equipment, disposal containers, and consumables, including cement and
other materials used in underground excavation. Most of these shipments would originate in the Las
Vegas metropolitan area. In addition, an estimated 54,000 shipments of office and laboratory supplies
and equipment, mail, and laboratory samples would occur during the 24 years of operation. A total of
about 21 million vehicle kilometers (13 million vehicle miles) of travel would be involved. Impacts
would include vehicle emissions, consumption of petroleum resources, increased truck traffic on regional
highways, and fatalities from accidents. Similarly, there would be about 76,000 shipments during the
76-year monitoring period after emplacement operations and 15,000 shipments during closure activities.
The number of shipments during shorter or longer monitoring periods would be proportionately fewer or
larger. Table J-62 summarizes these impacts.
Table J-62. Human health and safety impacts from shipments of material to the repository.'
Kilometers'" Fuel consumption Vehicle
traveled (thousands of emissions-
Phase (millions) Traffic fatalities liters)'^ related fatalities
Construction
8.2
-9.9
0.14
-0.17
1,900-
2,300
0.0006 - 0.0007
Operation and i
monitoring
Emplacement
and
development
29-66
0.5-
1.1
7,000-
15,000
0.002 - 0.005
Monitoring
26 years
6.5
0.1
1,500
0.0005
76 years
19
0.3
4,500
0.0014
276 years
69
1.2
16,000
0.005
Closure
4.1
0.1
1,000
0.0003
a. Impacts are totals for 24 years of operations.
b. To convert kilometers to miles, multiply by 0.62137.
c. To convert liters to gallons, multiply by 0.26418.
During the construction phase, many employees would use their personal automobiles to travel to
construction areas on the repository site and to highway or rail line construction sites. The estimated peak
level of direct employment during 5 years of repository construction would be 1,035 workers. Current
Nevada Test Site employees can ride DOE-provided buses to and from work; similarly, buses probably
would be available for repository construction workers, which would reduce the number of vehicles
traveling to the site each day by approximately a factor of 8. Table J-63 summarizes the anticipated
number of traffic -accident-related injuries and fatalities and the estimated consumption of gasoline that
would occur from this travel activity. The greatest impact of this traffic would be added congestion at the
northwestern Las Vegas Beltway interchange with U.S. Highway 95. Current estimates call for traffic at
this interchange during rush hours to be as high as 1,000 vehicles an hour (Clark County 1997,
Table 3-12, page 3-43). The additional traffic from repository construction, an estimated 500 vehicles per
hour, would add about 50 percent to traffic volume at peak rush hour and would contribute to congestion
although congestion in this area would be generally low.
The average level of employment during repository operations would be about 2,700 workers. As
mentioned above, DOE provides bus service from the Las Vegas area to and from the Nevada Test Site.
Table J-63 summarizes the anticipated number of traffic-accident-related fatalities and the estimated
consumption of gasoline that would occur from this travel activity. The greatest impact of this traffic
would be increased congestion at the northwestern Las Vegas Beltway interchange with U.S. 95. As
many as 500 vehicles an hour at peak rush hour would contribute to the congestion. Approximately
J- 106
Transportation
Table J-63. Health impacts from transportation of construction and operations workers."
Kilometers'" Vehicle
traveled Traffic Fuel consumption emissions-
Phase (in millions) fatalities (thousands of liters)*^ related fatalities
Construction 36.3 - 44.4 0.5 - 0.6
Operation and monitoring
Emplacement and development 240 -300 3.2 - 4.0
Monitoring (76 years) 62.2 . 0.8
Closure 20.2 - 42.7 0.3 - 0.6
400-500
0.0026 - 0.0032
2,600 - 3,3(X) 0.017 - 0.022
680 0.0045
220-470 0.0015-0.0031
a. Impacts are totals for 24 years for operations.
b. Toconvert kilometers to miles, multiply by 0.62137.
c. To convert liters to gallons, multiply by 0.26418.
150 people would be employed during monitoring and about 500 would be employed during closure. The
number of vehicles associated with these levels of employment would contribute negligibly to congestion.
Table J-64 lists the impacts associated with the delivery of fabricated disposal container components from
a manufacturing site to the repository. A total of 10,000 containers would be delivered; if a rail line to
Yucca Mountain was not available, the mode of transportation would be a combination of rail and
overweight truck. The analysis assumes that the capacity of each railcar would be two containers and that
the capacity of a truck would be one container, so there would be 5,000 railcar shipments to Nevada and
10,000 truck shipments to the Yucca Mountain site. The analysis estimated impacts for one national rail
route representing a potential route from a manufacturing facility to a Nevada rail siding. The analysis
estimated the impacts of transporting the containers from this siding over a single truck route — the
Apex/Dry Lake route analyzed for the transportation of spent nuclear fuel and high-level radioactive
waste by heavy-haul trucks. Although the actual mileage from a manufacturing facility could be shorter,
DOE decided to select a distance that represents a conservative estimate [4,439 kilometers (2,758 miles)].
The impacts are split into two subcategories — health effects from vehicle emissions and fatalities from
transportation accidents.
Table J-64. Impacts of disposal container shipments for Proposed Action."
Type of shipment Number of shipments Vehicle emissions-related health effects Traffic fatalities
Rail and truck
5.000 rail/1 0,000 truck
0.14
0.8
a. Impacts are totals for 24 years of operations.
Air Quality
The exhaust from vehicles involved in the transport of personnel and materials to the repository would
emit carbon monoxide, nitrogen dioxide, and particulate matter (PMio). Because carbon monoxide is the
principal pollutant of interest for evaluating impacts caused by motor vehicle emissions, the analysis
focused on it.
The analysis assumed that most of the personnel who would commute to the repository would reside in
the Las Vegas area and that most of the materials would travel to the repository from the Las Vegas area.
To estimate maximum potential emissions to the Las Vegas Valley airshed, which is in nonattainment for
carbon monoxide (FHWA 1996, pages 3-53 and 3-54), the analysis assumed that all personnel and
material would travel from the center of Las Vegas to the repository. Table J-65 lists the estimated
annual amount of carbon monoxide that would be emitted to the valley airshed during the phases of the
repository project and the percent of the corresponding threshold level.
As listed in Table J-65, the annual amount of carbon monoxide emitted to the nonattainment area would
be below the threshold level during all phases of the repository. In the operation phase, the estimated
annual amount of carbon monoxide emitted would be close (93 percent) to the threshold level. So, a more
J- 107
Transportation
Table J-65. Annual amount of carbon monoxide
emitted to Las Vegas Valley airshed from
transport of personnel and material to repository
(kilograms per year)^ for the Proposed Action.
Annual
GCR
emission
threshold
Phase
rate
level"
Construction
47,000
51
Operation and monitoring
Operation period
85,000
93
Monitoring period
6,700
7.4
Closure
17,000
19
a. To convert kilograms to tons, multiply by 0.00 11 023.
b. GCR = General Conformity Rule emission threshold
level for carbon monoxide is 91,000 kilograms
(100 tons) per year.
detailed analysis and conformity analysis might be
required to determine if mitigation would be needed
to ensure that the additional emissions did not
impede efforts in Nevada to bring the Las Vegas
area into attainment for carbon monoxide.
For areas that are in attainment, pollutant
concentrations in the ambient air probably would
increase due to the additional traffic but, given the
relatively small amount of traffic that passes
through these areas, the additional traffic would be
unlikely to cause the ambient air quality standards
to be exceeded.
Noise
Traffic -related noise on major transportation routes
used by the workforce would likely increase. The
analysis of impacts from traffic noise assumed that the workforce would come from Nye County (20
percent) and Clark County (80 percent). During the period of maximum employment in 2015, an
estimated daily maximum of 576 vehicles would pass through the Gate 100 entrance at Mercury during
rush hour (DOE 1996c, page 4-45), compared to a baseline of 232 vehicles per hour. This would result in
an increase in rush hour noise from 65.5 dBA to 69.5 dBA for the communities of Mercury and hidian
Springs. The 4.4-dBA increase could be perceptible to the communities but, because of the short
duration, would be unlikely to result in an adverse response.
J.3.6.2 Impacts of Transporting Wastes from the Repository
During repository construction and operations, DOE would ship waste and sample material from the
repository. The waste would include hazardous, mixed, and low-level radioactive waste. Samples would
include radioactive and nonradioactive hazardous materials shipped to laboratories for analysis. In
addition, nonhazardous solid waste could be shipped from the repository site to the Nevada Test Site for
disposal. However, as noted in Chapter 2, DOE proposes to include an industrial landfill on the
repository site. Table J-66 summarizes the maximum quantities of waste (generally from the uncanistered
packaging scenario and the low thermal load scenario) that DOE would ship from the repository and the
number of truck shipments.
Occupational and Pubiic Healtti and Safety
The quantities of hazardous waste that DOE would ship to approved facilities off the Nevada Test Site
would be relatively small and would present little risk to public health and safety. This waste could be
shipped by rail (if DOE built a rail line to the repository site) or by legal-weight truck to permitted
disposal facilities. The principal risks associated with shipments of these materials would be related to
traffic accidents. These risks would include 0.01 fatality for the combined construction, operation and
monitoring, and closure phases for hazardous wastes.
DOE probably would ship low-level radioactive waste by truck to existing disposal facilities on the
Nevada Test Site. Although these shipments would not use public highways, DOE estimated their risks.
As with shipments of hazardous waste, the principal risk in transporting low-level radioactive waste
would be related to traffic accidents. Because traffic on the Nevada Test Site is regulated by the Nye
County Sheriffs Department, DOE assumed that accident rates on the site are similar to those of
secondary highways in Nevada. Low-level radioactive waste would not be present during the
construction of the repository. Therefore, accidents involving such waste could occur only during the
J- 108
Transportation
Table J-66. Shipments of waste from the Yucca Mountain Repository.'
Construction
Operation and
monitoring
Closure
Volume
Number of
Volume
Number of
Volume
Number of
Waste
(cubic meters)''
shipments
(cubic meters)
shipments
(cubic meters)
shipments
Hazardous'
990
60
6,100
340
630
8
Low-level
0
0
68,000
1,800
3,500
2
radioactive''
Dual-purpose
0
0
30,000
6.600
0
0
canisters'
Mixed'
0
0
23
2
0
0
Nonhazardous solid^'^
13,000
120
90,000
810
160,000
1,400
a. Source: Chapter 4, Section 4. 1 . 1 2.
b. To convert cubic meters to cubic yards, multiply by 1 .3079.
c. Shipment numbers based on 1 6.64 cubic meters per shipment.
d. Shipment numbers based on 38 cubic meters per shipment.
e. Shipment numbers based on 23 metric tons per shipment.
f. Shipment numbers based on cubic meters per shipment.
g. Includes constmction and demolition debris and sanitary and industrial solid waste.
operation and monitoring and the closure phases, although most of this waste would be generated during
the operation and monitoring phase. DOE estimates 0.05 traffic fatality from the transportation of low-
level radioactive waste during the repository operation and monitoring and closure phases.
Air Quality
The quantities of hazardous waste that DOE would ship to approved facilities off the Nevada Test Site
would be relatively small. Vehicle emissions due to these shipments would present little risk to public
health and safety.
Bioiogical Resources and Soils
The transportation of people, materials, and wastes during the construction, operation and monitoring, and
closure phases of the repository would involve more than 1.6 billion vehicle-kilometers (1 billion vehicle-
miles) of travel on highways in southern Nevada. This travel would use existing highways that pass
through desert tortoise habitat. Individual desert tortoises probably would be killed. However, because
populations of the species are low in the vicinity of the routes (Bury and Germano 1994, pages 57 to 72),
few would be lost. Thus, the loss of individual desert tortoises due to repository traffic would not be
likely to be a threat to the conservation of this species. In accordance with requirements of Section 7 of
the Endangered Species Act, DOE would consult with the Fish and Wildlife Service and would comply
with mitigation measures resulting from that consultation to limit losses of desert tortoises from
repository traffic.
J.3.6.3 Impacts from Transporting Other Materials and People in Nevada for Inventory
Modules 1 and 2
The analysis evaluated impacts to occupational and public health and safety in Nevada from the transport
of materials, wastes, and workers (including repository-related commuter travel) for construction,
operation and monitoring, and closure of the repository that would occur for the receipt and emplacement
of materials in Inventory Modules 1 and 2. The analysis assumed that the routes and transportation
characteristics (for example, accident rates) for transportation associated with the Proposed Action and
Inventory Modules 1 and 2 would be the same. The only difference would be the projected number of
trips for materials, wastes, and workers traveling to the repository.
J-109
Transportation
Table J-67 lists estimated incident-free (vehicle emissions) impacts and traffic (accident) fatality impacts
in Nevada for the transportation of materials, wastes, and workers (including repository-related commuter
travel) for the construction, operation and monitoring, and closure of the repository that would occur for
the receipt and emplacement of the materials in Inventory Modules 1 and 2.
Table J-67.
and 2/
Impacts from transportation of materials, consumables, personnel, and waste for Modules 1
Category
Kilometers traveled''
Fatalities
Emission-related health effects
Materials
90 - 160
1.7-2.9
0.07 - 0.01
Personnel
490 - 650
4.9 - 6.5
0.04 - 0.05
Waste material (Module 1/Module 2)
Hazardous
0.17/0.20
0.018/0.021
0.00001/0.00001
Low-level radioactive
0.75/0.86
0.10/0.12
0.001
Nonhazardous solid
0.66
0.066
0.00005
Dual-purpose canisters
35
1.5
0.24
a. Numbers are rounded.
b. To convert kilometers to miles, multiply by 0.62137.
Even with the increased transportation of the other materials included in Module 1 or 2, DOE expects that
the transportation of materials, consumables, personnel, and waste to and from the repository would be
minor contributors to all transportation on a local, state, and national level. Public and worker health
impacts would be small from transportation accidents involving nonradioactive hazardous materials. On
average, in the United States there is about 1 fatality caused by the hazardous material being transported
for each 30 million shipments by all modes (DOT 1998a, page 1; DOT undated. Exhibit 2b).
J.3.6.4 Environmental Justice
The impacts of transporting people and materials other than spent nuclear fuel and high-level radioactive
waste would be small and random. Because the number of shipments and commuter trips would be small
in comparison to other commercial and commuter travel in southern Nevada and would use existing
transportation facilities in the area, impacts to land use; air quality; hydrology; biological resources and
soils; occupational and public health and safety; cultural resources; socioeconomics; noise; aesthetics;
utilities, energy, and materials; and waste management would be small. In addition, due to the nearly
random nature of accidents that would involve the transportation of materials and people, the probability
of such an accident would be small in any location, minimizing the risk at a specific location.
Furthermore, because potential accidents would be nearly random, impacts to minority or low-income
populations and to Native Americans along the routes in Nevada would be unlikely to be
disproportionately high and adverse.
Because there would be no adverse or disproportionate impacts from transportation of people and
materials, a detailed environmental justice study is not required.
J.3.6.5 Summary of Impacts of Transporting Other IVIaterials
Table J-68 summarizes the impacts of transporting other materials to the repository site for the Proposed
Action.
J-110
Transportation
Table J-68. Health impacts from transportation of materials, consumables, personnel, and waste for the
Proposed Action."
Category
Distance traveled
(kilometers)''
Impact
Human health and safety
Construction
Materials
Personnel
Waste
Hazardous
Low-level waste
Nonhazardous
Canisters
Operation and monitoring
Materials
Personnel
Waste
Hazardous
Low-level waste
Nonhazardous
Canisters
Closure
Materials
Personnel
Waste
Hazardous
Low-level waste
Nonhazardous
Canisters
Air quality
Construction traffic
Operation and monitoring traffic
Operations
Monitoring
Closure traffic
Biological resources
Noise
Environmental justice
8,200,000 - 9,900,000
36,300,000 - 44,400,000
14,500
C
29,000
57,000,000 - 94,000,000
300,000,000 - 360,000,000
90,000
435,000
196,000
1,590,000
4,400,000
20,200,000 - 42,700,000
9,200
22,200
338,000
0
74,000,000
860,000,000
170,000,000
1,000,000,000
1,000,000,000
0.14 -0.17 fatality
0.5 - 0.6 fatality
0.002 fatality
0.003 fatality
1.0- 1.6 fatalities
4.0 - 4.8 fatalities'*
0.002 fatality
0.008 fatality
0.003 fatality
0.028 fatality
0.1 fatality
0.3 - 0.6 fatality
0.001 fatality
0.002 fatality
0.04 fatality
75 percent of Air Quality General
Conformity Rule threshold for PM|o
170 percent of carbon monoxide threshold
9 percent of carbon monoxide threshold
30 percent of carbon monoxide threshold
Individual desert tortoises would be killed
but kills would not be likely to be a threat
to conservation of species
Small impacts unlikely to affect
communities
Traffic impacts unlikely to be high and
disproportionate for minority or low
income populations or populations of
Native Americans
a. Numbers are rounded.
b. To convert kilometers to miles, multiply by 0.62137.
c. - = none.
d. Monitoring for 76 years.
J-IU
Transportation
ACS 1998
Ardila-Coulson 1989
Battelle 1998
Best 1998
Biweretal. 1997
BTS 1996
BTS 1998
BTS 1999
Bury and Germano 1994
Caltrans 1997
Cashwell et al. 1986
Cerocke 1998
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€^xr<Tr^f/l^lrtm
Appendix K
Long-Term Radiological Impact
Analysis for the No-Action
Alternative
Long-Term Radiological Impact Analysis for the No-Action Alternative
TABLE OF CONTENTS
Section Page
K.l Intrcxiuction K-1
K.2 Analytical Methods K-3
K.2.1 General Methodology K-3
K.2.1.I Concrete Storage Module Degradation K-4
K.2.1.2 Storage Canister Degradation K-8
K.2.1.3 Infiltration K-8
K.2.1.4 Cladding K-9
K.2.1.4.1 Commercial Spent Nuclear Fuel Cladding K-9
K.2.1.4.2 DOE Spent Nuclear Fuel Cladding K-11
K.2. 1.5 Dissolution of Spent Nuclear Fuel and High-Level Radioactive Waste K-U
K.2.1.5.1 Commercial Spent Nuclear Fuel Dissolution K-U
K.2.1.5.2 DOE Spent Nuclear Fuel Dissolution K-12
K.2.1.5.3 High-Level Radioactive Waste Dissolution K-12
K.2. 1.6 Regionalization of Sites for Analysis K-12
K.2.2 Radionuclide Release K-12
K.2.3 Environmental Transport of Radioactive Materials K-14
K.2.3.1 Groundwater Transport K-16
K.2.3.2 Surface-Water Transport K-18
K.2.3.3 Atmospheric Transport K-19
K.2.4 Human Exposure, Dose, and Risk Calculations K-19
K.2.4.1 Gardener Impacts K-20
K.2.4.2 Direct Exposure K-23
K.2.5 Accident Methodology K-24
K.2.5.1 Aircraft Crash K-25
K.2.5.2 Criticality K-26
K.3 Results K-27
K.3.1 Radiological Impacts K-27
K.3.2 Unusual Events K-33
K.3.2.1 Accident Scenarios K-33
K.3.2.2 Sabotage K-34
K.4 Uncertainties K-34
K.4.1 Societal Values, Natural Events, and Improvements in Technology K-35
K.4.1.1 Societal Values K-35
K.4.1.2 Changes In Natural Events K-35
K.4.1.3 Improvements in Technology K-36
K.4.2 Changes in Human Behavior K-36
K.4.3 Mathematical Representations of Physical Processes and of the Data Input K-36
K.4.3.1 Waste Package and Material Degradation K-37
K.4.3.2 Consequences of Radionuclide Release K-37
K.4.3.3 Accidents and Their Uncertainty K-41
K.4.4 Uncertainty Summary K-41
References K-42
K-iii
Long-Term Radiological Impact Analysis for the No-Action Alternative
LIST OF TABLES
Table Page
K-1 Time after the assumed loss of effective institutional control at which first failures
would occur and radioactive materials could reach the accessible environment K-9
K-2 Average regional precipitation K-9
K-3 Proposed Action and Module 1 quantities of spent nuclear fuel and canisters of
high-level radioactive waste in each geographic region K-13
K-4 Radionuclides and relative contributions over 1 0,000 years to Scenario 2 impacts K- 1 4
K-5 Multimedia Environmental Pollutant Assessment System default elemental
equilibrium adsorption coefficients for soil pH between 5 and 9 K-17
K-6 Regional source terms and environmental transport data for important isotopes
used for collective drinking water radiological impact analysis K-18
K-7 Transport and population data for drinking water pathway impact analysis K-18
K-8 Multimedia Environmental Pollutant Assessment System regional groundwater
input parameters K-19
K-9 Multimedia Environmental Pollutant Assessment System human exposure input
parameters for determination of all pathways radiological impacts sensitivity
analysis K-2I
K-10 Multimedia Environmental Pollutant Assessment System groundwater transport
input parameters for estimating radiological impacts to the onsite and near-site
gardener K-23
K-1 1 Estimated collective radiological impacts to the public from continued storage of
Proposed Action and Module 1 inventories of spent nuclear fuel and high-level
radioactive waste at commercial and DOE storage facilities - Scenario 2 K-28
K-1 2 Estimated internal dose rates and year of peak exposure for the onsite and near-site
gardeners - Scenario 2 K-32
K-13 Estimated external peak dose rates for the onsite and near-site gardeners -
Scenario 2 K-33
K-14 Consequences of aircraft crash onto degraded spent nuclear fuel concrete storage
module K-34
K-15 Review of approaches, assumptions, and related uncertainties K-38
LIST OF FIGURES
Figure Page
K-1 Primary steps and processes involved in the degradation of the engineered barrier
system K-5
K-2 No-Action Alternative analysis regions K-6
K-3 Failure times for above-ground concrete storage modules K-7
K-4 Precipitation ranges for regions with existing spent nuclear fuel and high-level
radioactive waste storage facilities K-10
K-5 Percent of commercial spent nuclear fuel exposed over time due to new failures K-1 1
K-6 Potential exposure pathways associated with degradation of spent nuclear fuel and
high-level radioactive waste K-15
K-7 Major waterways near commercial and DOE sites K-29
K-8 Regional collective dose from the Proposed Action inventory under No-Action
Scenario 2 K-31
K-9 Total potential latent cancer fatalities throughout the United States from the
Proposed Action inventory under No-Action Scenario 2 K-31
K-iv
Long-Term Radiological Impact Analysis for the No-Action Alternative
APPENDIX K. LONG-TERM RADIOLOGICAL IMPACT ANALYSIS
FOR THE NO-ACTION ALTERNATIVE
K.1 Introduction
This appendix provides detailed information related to the radiological impact analysis for No-Action
Alternative Scenario 2, including descriptions of the conceptual models used for facility degradation,
spent nuclear fuel and high-level radioactive waste material degradation, and data input parameters. In
addition, this appendix discusses the computer programs and exposure calculations used. The methods
described include summaries of models and programs used for radioactive material release,
environmental transport, radiation dose, and radiological human health impact assessment. Although the
appendix describes No- Action Scenario 1, it focuses primarily on the long-term (100 to 10,000 years)
radiological impacts associated with Scenario 2.
NO-ACTION ALTERNATIVE SCENARIOS 1 AND 2
Under the Nuclear Waste Policy Act, the Federal Government has the responsibility to provide
permanent disposal of spent nuclear fuel and high-level radioactive waste to protect the public's
health and safety and the environment. DOE intends to comply with the terms of existing consent
orders and compliance agreements on the management of spent nuclear fuel and high-level
radioactive waste. However, the course that Congress, DOE, and the commercial nuclear utilities
would take if there was no recommendation to use Yucca Mountain as a repository is highly
uncertain.
In light of these uncertainties, it would be speculative to attempt to predict precise consequences. To
illustrate one set of possibilities, however, DOE decided to focus the analysis of the No-Action
Alternative on the potential impacts of two scenarios:
Scenario 1: Long-term storage of spent nuclear fuel and high-level radioactive waste at the current
storage sites, with effective institutional control for at least 10,000 years.
Scenario 2: Long-term storage of spent nuclear fuel and high-level radioactive waste, with the
assumption of no effective institutional control after approximately 100 years.
DOE recognizes that neither of these scenarios is likely to occur if there was a decision to not
develop a repository at Yucca Mountain. However, the Department selected these two scenarios for
analysis because they provide a baseline for comparison to the impacts from the Proposed Action
and because they reflect a range of the potential impacts that could occur.
To permit a comparison of the impacts between the construction, operation and monitoring, and eventual
closure of a proposed repository at Yucca Mountain and No-Action Scenario 2, the U.S. Department of
Energy (DOE) took care to maintain consistency, where possible, with the modeling techniques used to
conduct the Viability Assessment of a Repository at Yucca Mountain (DOE 1998, all) and in the Total
System Performance Assessment - Viability Assessment (TSPA-VA) Analyses Technical Basis Document
(TRW 1998a,b,c,d,e,f,g,h,i,j,k, all) for the proposed repository (see Appendix I, Section LI, for details).
In pursuit of this goal, DOE structured this analysis to facilitate an impact comparison with the repository
impact analysis. Important consistencies include the following:
• Identical evaluation periods (1(X) years and 10,(XX) years)
K-1
Long-Term Radiological Impact Analysis for the No-Action Alternative
• Identical spent nuclear fuel and high-level
radioactive waste inventories at the reference
repository:
- Proposed Action: 63,000 metric tons of
heavy metal (MTHM) of commercial spent
nuclear fuel; 2,333 MTHM of DOE spent
nuclear fuel; 8,315 canisters of high-level
radioactive waste; and 50 MTHM of
surplus weapons-usable plutonium
- Module 1: All Proposed Action materials,
plus an additional 42,000 MTHM of
commercial spent nuclear fuel; 167 MTHM
of DOE spent nuclear fuel; and 13,965
canisters of high-level radioactive waste.
This would result in a total of
approximately 105,000 MTHM of
commercial spent nuclear fuel; 2,500
MTHM of DOE spent nuclear fuel; and 22,280
MTHM of surplus weapons-usable plutonium (
DEFINITION OF
METRIC TONS OF HEAVY METAL
Quantities of spent nuclear fuel are
traditionally expressed in terms of metric
tons of heavy metal (typically uranium),
without the inclusion of other materials such
as cladding (the tubes containing the fuel)
and structural materials. A metric ton is
1,000 kilograms (1.1 tons or 2,200 pounds).
Uranium and other metals in spent nuclear
fuel (such as thorium and plutonium) are
called heavy metals because they are
extremely dense; that is, they have high
weights per unit volume. One metric ton of
heavy metal disposed of as spent nuclear
fuel would fill a space approximately the size
of a typical household refrigerator.
canisters of high-level radioactive waste, plus 50
see Appendix A, Figure A-2).
• Consistent spent nuclear fuel and high-level radioactive waste corrosion and dissolution models
• Identical radiation dose and risk conversion factors
• Similar assumptions regarding the future habits and behaviors of population groups (that is, that they
will not be much different from those of populations today)
For commercial facilities, the No-Action analysis estimated short- and long-term radiological impacts for
Scenario 1 and short-term impacts for Scenario 2 during the first 100 years for facility workers and the
public based on values provided by the U.S. Nuclear Regulatory Commission (NRC 1991a, page 21). For
DOE facilities, radiological impacts for these p)eriods under Scenarios 1 and 2 were estimated based on
analysis by Orthen (1999, all). To ensure consistency with the repository impact analysis, the long-term
facility degradation and environmental releases of radioactive materials were estimated by adapting Total
System Performance Assessment process models developed to predict the behavior of spent nuclear fuel
and high-level radioactive waste in the repository (Battelle 1998, pages 2.4 to 2.9).
Because DOE did not want to unduly influence the results to favor the repository, it used assumptions
were that generally resulted in lower predicted impacts (rather than applying the bounding assumptions
used in many of the repository impact analyses) if Total System Performance Assessment models were
not available or not appropriate for this continuous storage analysis. For example, the No-Action
Scenario 2 analysis took into account the protectiveness of the stainless-steel waste canister when
estimating releases of radioactive material from the vitrified high-level radioactive waste; the Total
System Performance Assessment assumed no credit for material protection or radionuclide retardation by
the intact canister. This approach dramatically reduced the release rate of high-level radioactive waste
materials to the environment, thereby resulting in lower estimated total doses and dose rates to the
exposed populations. Conversely, in many instances the Total System Performance Assessment selected
values for input parameters that defined ranges to ensure that there would be no underestimation of the
associated impacts. Section K.4 discusses other consistencies and inconsistencies between the Total
System Performance Assessment and the No-Action analysis.
K-2
Long-Term Radiological Impact Analysis for the No-Action Alternative
The long-term impact analysis used recent climate and meteorological data, assuming they would remain
constant throughout the evaluation period (Poe and Wise 1998, all). DOE recognizes that there could be
considerable changes in the climate over 10,000
years (precipitation patterns, ice ages, global
warming, etc.) but, to simplify the analysis, did not
attempt to quantify climate changes. Section
K.4.1.2 discusses the difficulties of modeling these
changes and the potential effect on outcomes
resulting from uncertainties associated with
predicting potential future climatic conditions.
Although the repository Total System Performance
Assessment used probabilistic process models to
evaluate the transport of radioactive materials
within Yucca Mountain and underlying
groundwater aquifers, DOE used the deterministic
computer program Multimedia Environmental
Pollutant Assessment System (MEPAS; Buck et
al. 1995, all) for the No- Action Scenario 2 analysis
because of the need to model the transport of
radioactive material. In addition, it discusses
environmental pathways not present at the
repository (for example, the movement of
contaminants through surface water). The
MEPAS program has been accepted and used by
DOE and the Environmental Protection Agency
for long-term performance assessments (Rollins
1998a, pages 1, 10, and 19).
PROBABILISTIC AND DETERMINISTIC
ANALYSES
A probabilistic analysis represents data input
to a model as a range of values that
represents the uncertainty associated with the
actual or true value. The probabilistic model
randomly samples these input parameter
distributions many times to develop a possible
range of results. The range of results provides
a quantitative estimate of the uncertainty of the
results.
A deterministic analysis uses a best estimate
single value for each model input and
produces a single result. The deterministic
analysis will usually include a separate
analysis that addresses the uncertainty
associated with each input and provides an
assessment of impact these uncertainties
could have on the model results.
Analyses can use both approaches to provide
similar information regarding the uncertainty of
the results.
K.2 Analytical Methods
This section describes the methodology used to evaluate the long-term degradation of the concrete
facilities, steel storage containers, and spent nuclear fuel and high-level radioactive waste materials. In
addition, it discusses the eventual release and transport of radioactive materials under Scenario 2. The
institutional control assumed under Scenario 1 would ensure ongoing maintenance, repair and
replacement of storage facilities, and containment of spent nuclear fuel and high-level radioactive waste.
For this reason, assuming the degradation of engineered barriers and the release and transport of
radioactive materials is not appropriate for Scenario 1. The Scenario 2 analysis assumed that the
degradation process would begin at the time when there was no effective institutional control (that is,
after approximately 100 years) and the facilities would no longer be maintained. This section also
describes the models and assumptions used to evaluate human exposures and potential health effects, and
cost impacts.
K.2.1 GENERAL METHODOLOGY
For the No-Action analysis, the facilities, dry storage canisters, cladding, spent nuclear fuel, and high-
level radioactive waste material, collectively known as the engineered barrier system, were modeled
using an approach consistent (to the extent possible) with that developed for the Viability Assessment
(DOE 1998, Volume 3). These process models were developed to evaluate, among other things, the
performance of the repository engineered barrier system in the underground repository environment. In
this analysis, the process models were adapted whenever feasible to evaluate surface environmental
conditions at commercial and DOE sites. These models are described below.
K-3
Long-Term Radiological Impact Analysis for the No-Action Alternative
Figure K-1 shows the modeling of the degradation of spent nuclear fuel and high-level radioactive waste
and the release of radioactive materials over long periods. Five steps describe the process of spent
nuclear fuel and high-level radioactive waste degradation; a sixth step, facility radioactive material
release, describes the amount and rate of precipitation that would transport the radioactive material or
dissolution products to the environment. This section describes each process and the results. Additional
details are provided in reference documents (Poe 1998a, all; Battelle 1998, all).
Environmental parameters important to the degradation processes include temperature, relative humidity,
precipitation chemistry (pH and chemical composition), precipitation rates, number of rain-days, and
freeze/thaw cycles. Other parameters considered in the degradation process describe the characteristics
and behavior of the engineered barrier system, including barrier material composition and thickness. To
simplify the analysis, the United States was divided into five regions (as shown in Figure K-2) for the
purposes of estimating degradation rates and human health impacts (see Section K.2.1.6 for additional
details).
Under the No-Action Alternative, commercial utilities would manage their spent nuclear fuel at
72 nuclear power generating facilities. DOE would manage its spent nuclear fuel and high-level
radioactive waste at five DOE facilities [the Hanford Site (Region 5), the Idaho National Engineering and
Environmental Laboratory (Region 5), Fort St. Vrain (Region 5), the West Valley Demonstration Project
(Region 1), and the Savannah River Site (Region 2)]. The No- Action analysis evaluated DOE spent
nuclear fuel and high-level radioactive waste at the commercial and DOE sites or at locations where
Records of Decision have placed or will place these materials (for example, West Valley Demonstration
Project spent nuclear fuel was evaluated at the Idaho National Engineering and Environmental Laboratory
(60 FR 28680, June 1, 1995). Therefore, the No- Action analysis evaluated DOE aluminum-clad spent
nuclear fuel at the Savannah River Site and DOE non-aluminum-clad fuel at the Idaho National
Engineering and Environmental Laboratory. DOE evaluated most of the Fort St. Vrain spent nuclear fuel
at the Colorado site. In addition, the analysis evaluated high-level radioactive waste at the West Valley
Demonstration Project, the Idaho National Engineering and Environmental Laboratory, the Hanford Site,
and the Savannah River Site.
K.2.1.1 Concrete Storage Module Degradation
The first process model analyzed degradation mechanisms related to failure of the concrete storage
module. Failure is defined as the time when precipitation would infiltrate the concrete and reach the
spent nuclear fuel or high-level radioactive waste storage canister. The analysis (Poe 1998a, Section 2.0)
considered degradation due to exposure to the surrounding environment.
The primary cause of failure of surface-mounted concrete structures is freeze/thaw cycles that cause the
concrete to crack and spall (break off in layers), which allows precipitation to enter the concrete, causing
more freeze damage. Freeze/thaw failure is defined as the time when half of the thickness of the concrete
is cracked and spalled. Some regions (coastal California, Texas, Florida, etc.) are essentially without the
freeze/thaw cycle. In these locations the primary failure mechanism is chlorides in precipitation, which
decompose the chemical constituents of the concrete into sand-like materials. This process progresses
more slowly than the freeze/thaw process. Figure K-3 shows estimated concrete storage module failure
times.
Below-grade concrete structures, such as those used to store some of the DOE spent nuclear fuel and most
of the high-level radioactive waste, would be affected by the same concrete degradation mechanisms as
surface facilities. Below grade, the freeze/thaw degradation would not be as great because the soil would
moderate temperature fluctuations. The primary failure mechanism for below-grade facilities would be
the loss of the above-grade roof, which would result in precipitation seeping around shield plugs. The
K-4
Long-Term Radiological Impact Analysis for the No- Action Alternative
^^^ ~~-~-^^^
f Environmental conditions^
Concrete storage module 1
X^^ pH, etc.) ^^
Dry storage canister exposed
Dry storage canister 1
>'
Infiltration 1
Water contacts spent nuclear fuel and
high-level radioactive waste material
Cladding 1
> r
Dissolution 1
^
Spent nuclear fuel and high-level
radioactive waste forms degrade
Facility ra
material
idioactive 1 ^ Cnrforp-watPr
flux
release 1 ► Surface-water
Percolation
groundwater
flux
Source: Adapted from Banelte (1998, page 2.4).
Figure K-1. Primary steps and processes involved in the degradation of the engineered barrier system.
K-5
Long-Term Radiological Impact Analysis for the No-Action Alternative
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K-6
Long-Term Radiological Impact Analysis for the No-Action Alternative
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K-7
Long-Term Radiological Impact Analysis for the No-Action Alternative
analysis assumed that this would occur 50 years after the end of facility maintenance, and that this would
be the reasonable life expectancy of a facility without maintenance and periodic repair (Poe 1998a,
pages 4-6 to 4-19).
K.2.1.2 Storage Canister Degradation
The second process analyzed was spent nuclear fuel and high-level radioactive waste storage canister
degradation. For commercial and DOE spent nuclear fuel, the analysis defined failure of the stainless-
steel dry storage canister as the time at which precipitation penetrated the canister and wet the spent
nuclear fuel. The analysis defined failure for the high-level radioactive waste as the time at which
precipitation penetrated the canister. This is consistent with the repository definition that failure of the
waste package would occur when water penetrated the package and came in contact with the contents.
The stainless-steel model used for the No-Action analysis was consistent with the waste package inner
layer corrosion model used for the repository Total System Performance Assessment (DOE 1998,
Volume 3, Section 3.4) with the functional parameters modified to incorporate stainless-steel corrosion
data (Section K.4.3.1 discusses the sensitivity of outcome to carbon-steel dry storage containers). In
addition, the analysis used parameters appropriate for above-ground conditions, including temperature,
meteorological data, and chemical constituents in the atmosphere and precipitation. Although
inconsistent with the assumptions used for the Total System Performance Assessment, the analysis took
credit for the protectiveness of the high-level radioactive waste canister because (1) it is the only
container between the waste material and the environment and, (2) to ignore the protectiveness of this
barrier would have resulted in a considerable overestimation of impacts. This approach is consistent with
the decision, in the case of the No-Action Scenario 2 analysis, to provide a realistic radionuclide release
rate where possible and to preclude the overestimation of the associated radiological human health
impacts.
The primary determinants of stainless-steel corrosion for the different regions are the amount, the acidity,
and the chloride concentration of the precipitation. The storage canisters degrade faster in the below-
grade storage configuration than on the surface due to the higher humidity in the below-grade
environment. The storage canisters degrade faster in the below-grade storage configuration than on the
surface due to the higher humidity in the below-grade environment. The high-level radioactive waste
canisters degrade faster than the spent nuclear fuel canisters because they are not as thick. The analysis
evaluated three corrosion mechanisms — general corrosion, pitting corrosion, and crevice corrosion
(Battelle 1998, Appendix A). Of the three, crevice corrosion would be the dominant failure mechanism
for the regions analyzed. Corrosion rates and penetration times vary among the different regions of the
country. The analysis calculated regional penetration times from the time at which it assumed that
precipitation first would come in contact with the stainless steel. Table K-1 lists the results.
K.2.1.3 Infiltration
The third process analyzes infiltration of water to the spent nuclear fuel and high-level radioactive waste.
The amount of water in contact with these materials would be directly related to the size of the dry
storage canister footprint and the mean (average) annual precipitation at each storage site. The rate of
precipitation varies throughout the Unites States from extremely low (less than 25 centimeters [10 inches]
per year) in the arid portions of the west to high (more than 150 centimeters [60 inches] per year) along
the Gulf Coast in the southeast (Table K-2, Figure K-4). Local precipitation rates were used to determine
the amount of water available that could cause dry storage canister and cladding failure, and spent nuclear
fuel and high-level radioactive waste material dissolution.
K-8
Long-Term Radiological ImpacI Analysis for the No-Action Alternative
Table K-1. Time (years) after the assumed loss of effective institutional control at which first failures
would occur and radioactive materials could reach the accessible environment.
Weather' Canister'' breached
Material Region Storage facility protection lost (initial material release)
Commercial spent nuclear fuel
DOE spent nuclear fiiel
High-level radioactive waste
1
Surface
100
1,400
2
Surface
700
1,500
3
Surface
170
1,100
4
Surface
750
1,600
5
Surface
3,500
5,400
2
Surface
700
1,400
5
Surface
50
1,400
5
Below grade
50
800
1
Surface
100
uoo
2
Below grade
50
500
5
Below grade
50
700
a. Source; Adapted from Poe (1998b, Appendix A).
b. Source: Battelle (1998, data files, all); spent nuclear fuel dry storage or high-level radioactive waste canister.
Table K-2. Average regional precipitation
Annual
precipitation
Percent of days with
Region
(centimeters)''
precipitation
1
110
30
2
130
29
3
80
33
4
110
31
5
30
24
a. Source: Adapted from Poe (1998b, Appendix A, pages A-13 to A-16).
b. To convert centimeters to inches, multiply by 0.3937.
K.2.1.4 Cladding
The fourth process analyzed was failure of the cladding, which is a protective barrier, usually metal
(aluminum, zirconium alloy, stainless steel, nickel-chromium, Hastalloy, tantalum, or graphite),
surrounding the spent nuclear fuel material to contain radioactive materials. For spent nuclear fuel,
cladding is the last engineered barrier to be breached before the radioactive material can begin to be
released to the environment.
K.2.1 .4.1 Commercial Spent Nuclear Fuel Cladding
The principal cladding material used on commercial spent nuclear fuel is zirconium alloy. About
1.2 percent (of MTHM) of commercial spent nuclear fuel is stainless-steel clad (Appendix A, Section
A.2. 1.5.3). To be consistent with the Total System Performance Assessment, this analysis evaluated two
cladding failure mechanisms: (1) so-caA\ed juvenile failures (failures existing at the start of the analysis
period), and (2) new failures (failures that occur during the analysis period due to conditions in the
storage container). The analysis assumed that juvenile failures existed in 0. 1 percent of the zirconium
alloy-clad spent nuclear fuel and in all of the stainless-steel-clad fuel at the beginning of the analysis
period, and that after failure the cladding would offer no further protection to the radioactive material
[this is consistent with the Viability Assessment assumption (DOE 1998, Volume 3, page 3-97)].
Figure K-5 shows new failures (expressed as percent of commercial spent nuclear fuel over time) of
zirconium alloy cladding, which were modeled using the median value assumed in the Total System
Performance Assessment-Viability Assessment cladding abstraction (TRW 1998f, pages 6-19 to 6-54)
K-9
Long-Term Radiological Impact Analysis for the No-Action Alternative
K-10
Long-Term Radiological Impact Analysis for the No-Action Alternative
100-
10-
0.1
S 0.01
0.001
for zirconium alloy corrosion. The
Viability Assessment (DOE 1998, Volume
3, all) defines this information as a
"fractional multiplier," which is calculated
from the fraction of the failed fuel pin
surface area. In the No-Action analysis,
this corrosion is assumed to commence
when weather protection afforded by the
waste package is lost and the cladding is
exposed to environmental precipitation.
The Total System Performance
Assessment- Viability Assessment also
considers cladding failure from creep
strain, delayed hydride cracking, and
mechanical failure from rock falls. These
additional mechanisms normally occur
after the 10,000-year analysis period and
are therefore not considered in the No-
Action analysis. As shown in Figure K-5,
during the 10,000-year analysis period, less than 0.01 percent of the zirconium alloy-clad spent nuclear
fuel would be expected to fail. If the upper limit curve from Figure 4 of the Total System Performance
Assessment-Viability Assessment cladding abstraction (TRW 1998f, pages 6-19 to 6-54) was used, the
value could be as high as 0.5 percent of the zirconium alloy-clad spent nuclear fuel. The lower limit
value from the Total System Performance Assessment-Viability Assessment cladding abstraction curve
would be much less than 0.001 percent.
0.0001 ■
100
1,000
10,000
100,000
1,000,000
Years
Source: Adapted from TRW (19d8t, Figure 6-5).
Figure K-5. Percent of commercial spent nuclear fuel
exposed over time due to new failures.
K.2.1 .4.2 DOE Spent Nuclear Fuel Cladding
The composition and cladding materials of DOE spent nuclear fuel vary widely. The cladding
assumption for the surrogate material used in this analysis is identical (no cladding credit) to the
assumption used in the Total System Performance Assessment analysis (see Section K.4.3.2 for the
discussion of uncertainty in relation to cladding).
K.2.1. 5 Dissolution of Spent Nuclear Fuel and High-Level Radioactive Waste
The fifth process analyzed was the dissolution of the spent nuclear fuel and high-level radioactive waste.
The rate of release of radionuclides from these materials would be related directly to the amount of
surface area exposed to moisture, the quantity and chemistry of available water, and temperature. The
Total System Performance Assessment process model, modified to reflect surface environmental
conditions (temperature, relative humidity, etc.), was used to estimate release rates from the exposed
spent nuclear fuel and high-level radioactive waste. The model and application to surface conditions is
described in detail in Battelle (1998, pages 2.9 to 2.1 1).
K.2.1 .5.1 Commercial Spent Nuclear Fuel Dissolution
Consistent with the repository impact analysis, this analysis estimated that new zirconium alloy failures
would begin late in the 10,000-year period (see Figure K-5). As discussed in Section K.2.1. 4.1, only
0.01 percent of the zirconium alloy-clad spent nuclear fuel would be likely to fail during the 10,000-year
analysis period. Therefore, most of the exposed material considered in this analysis would result from
juvenile failures of zirconium alloy- and stainless-steel-clad spent nuclear fuel.
K-11
Long-Term Radiological Impact Analysis for the No-Action Alternative
K.2.1 .5.2 DOE Spent Nuclear Fuel Dissolution
The analysis assumed that DOE spent nuclear fuel would be a metallic uranium fuel with zirconium alloy
cladding (a representative or surrogate fuel that consisted primarily of N-Reactor fuel). Consistent with
the repository input analysis, the No-Action Scenario 2 analysis takes no credit for the cladding. The
analysis used the Total System Performance Assessment model for metallic uranium fuel, modified for
surface environmental conditions, to predict releases of the DOE spent nuclear fuel.
K.2.1 .5.3 High-Level Radioactive Waste Dissolution
Most high-level radioactive waste would be stored in below-grade concrete vaults. As discussed in
Section K.2.1.1, these vaults would be exposed to precipitation as soon as weather protection was lost
(the model assumed this would occur 50 years after loss of institutional control). After the loss of
weather protection and failure of the stainless-steel canisters, the high-level radioactive waste would be
exposed to precipitation. The environment in the underground vault would be humid and deterioration
would occur. Thus, the material would be exposed to either standing water or humid conditions in the
degrading vaults after the canister failed. The borosilicate glass deterioration model used in this analysis
was the same as the Total System Performance Assessment model modified to reflect surface conditions
(temperature and precipitation chemistry).
K.2.1 .6 Regionalization of Sites for Analysis
The climate of the contiguous United States varies considerably across the country. The release rate of
the radionuclide inventory would depend primarily on the interactions between environmental conditions
(rainfall, freeze-thaw cycles) and engineered barriers. To simplify the analysis, DOE divided the country
into five regions (see Figure K-2) (Poe 1998b, page 2).
The analysis assumed that a single hypothetical site in each region would store all the spent nuclear fuel
and high-level radioactive waste existing in that region. Such a site does not exist but is a mathematical
construct for analytical purposes. To ensure that the calculated results for the regional analyses reflect
appropriate inventory, facility and material degradation, and radionuclide transport, the spent nuclear fuel
and high-level radioactive waste inventories, engineered barriers, and environmental conditions for the
hypothetical sites were developed from data for each of the existing sites in the given region. Weighting
criteria to account for the amount and types of spent nuclear fuel and high-level radioactive waste at each
site were used in the development of the environmental data for the regional site, such that the results of
the analyses for the hypothetical site were representative of the sum of the results of each actual site if
they had been modeled independently (Poe 1998b, page 1). If there are no storage facilities in a particular
area of the country, the environmental parameters of that area were not evaluated.
Table K-3 lists the Proposed Action and Module 1 quantities of commercial spent nuclear fuel, DOE
spent nuclear fuel, and high-level radioactive waste in each of the five regions. The values in Table K-1
are the calculated results of failures of the various components of the protective engineered barriers and
release of radioactive material in each region.
K.2.2 RADIONUCLIDE RELEASE
The sixth and final step in the process is the release of radioactive materials to the environment. The
anticipated release rates (fluxes) were estimated in terms of grams per 70-year period (typical human life
expectancy in the United States) of uranium dioxide, uranium metal, or borosilicate glass for commercial
spent nuclear fuel, DOE spent nuclear fuel, and high-level radioactive waste, respectively. To assess
potential lifetime impacts on human receptors, the amount of fission products and transuranics associated
K-12
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-3. Proposed Action and Module 1 quantities of spent nuclear fuel (metric tons of heavy metal)
and canisters of high-level radioactive waste in each geographic region.''''
Commercial spent
nuclear fuel'
DOE spent
High
With juvenile cladding
Stainless-
-level
Region
total"
failure
steel cladding
nuclear fuel'
radioactive waste'
Proposed
Proposed
Proposed Action
Proposed
Proposed
Action
Module 1
Action
Module 1
and Module 1*
Action
Module 1
Action
Module 1
Region
(MTHM)
(MTHM)
(MTHM)
(MTHM)
(MTHM)
(MTHM)
(MTHM)
(canisters) (canisters)
1
17,000
27,000
16
27
410
300
300
2
19,000
32,000
19
32
0
30
45
6,000
6,200
3
15,000
23,000
15
23
170
4
7,200
14,000
7
14
0
5
5,400
10,000
5
9
140
2,300
2,455
2,000
15,500
Totals
63,600
106,000
62
105
720
2300
2,500
8300
22,000
a. Source: Appendix A.
b. Totals might differ from sums due to rounding.
c. All analyzed as stored on surface as shown on Chapter 2, Figures 2-36, 2-37, and 2-38.
d. Includes plutonium in mixed-oxide spent nuclear fuel, which is assumed to behave like other commercial spent nuclear fuel.
e. A representative or surrogate fuel that consisted primarily of N-reactor fuel.
f. Includes plutonium in can-in-canister.
g. Assumes failure of 100 percent of stainless-steel-clad when placed into dry storage.
with gram quantities of uranium dioxide, uranium metal, and borosilicate glass were calculated for
approximately 140 consecutive 70-year average human lifetimes to determine releases from the
10,000-year analysis period. Weighting criteria were used to ensure appropriate contributions by the
different types of spent nuclear fuel and the high-level radioactive waste in each region, as appropriate.
The result was a single release rate for each region that accounted for the different materials (uranium
dioxide, uranium metal, and borosilicate glass).
The radionuclide distributions in the spent nuclear fiiel and high-level radioactive waste (Appendix A)
were used for these analyses. These were expressed as radionuclide-specific curies for storage packages
(assembly or canister). The curies per storage
package were converted to curies per gram of
uranium dioxide, uranium metal, or borosilicate
glass (as described above for each spent nuclear
fuel and high-level radioactive waste material).
This radionuclide distribution was multiplied by
release flux (curies of spent nuclear fuel and
high-level radioactive waste material per
70-year period) after being corrected for decay
and the ingrowth of decay products for various
times after disposal. These corrections were
determined using the ORIGEN computer
program (ORNL 1991, all) for each of the
approximately 140 consecutive 70-year human
lifetimes to determine the release over the
10,000-year period. The results of the ORIGEN
runs were used as input to the environmental
transport program.
In addition to the 53 isotopes important to the
repository long-term impact analysis specified
in Appendix A, the No-Action Scenario 2
analysis considered 167 other isotopes in the
DEFINITIONS
Fission products: Elements produced when
uranium atoms split in a nuclear reactor, some
of which are radioactive. Examples are cesium,
iodine, and strontium.
Transuranics: Radioactive elements, heavier
than uranium, that are produced in a nuclear
reactor when uranium atoms absorb neutrons
rather than splitting. Examples of transuranics
include plutonium, americium, and neptunium.
Curie: The basic unit of radioactivity. It is
equal to the quantity of any radionuclide in
which 37 billion atoms are decaying per second.
Specific activity: An expression of the number
of curies of activity per gram of a given
radionuclide. It is dependent on the half life and
molecular weight of the nuclide.
K-13
Long-Term Radiological Impact Analysis for the No-Action Alternative
light-water reactor radiological database (DOE 1992, Page 1.1-1). Of the 220 isotopes evaluated, six
would contribute more than 99.5 percent of the total dose. Table K-4 lists these six isotopes along with
technetium-99, which individually would contribute less than 0.003 percent of the total dose. Plutonium-
239 and -240 would contribute more than 96 percent of the radiological impacts during the 10,000-year
analysis period because of their very large dose conversion factors. Americium-24rand -243 would be
minor contributors to the dose. Neptunium-237 and technetium-99 were of tertiary importance
(Table K-4).
Table K-4. Radionuclides and relative contributions
over 10,000 years to Scenario 2 impacts.'
Isotope Percent of total dose
Americium-241 3.2
Americium-243 0.86
Neptunium-237 0.29
Plutonium-238 0.2
Plutonium-239 49.0
Plutonium-240 47.0
Technetium-99 < 0.003
a. Source: Toblin (1998a, page 6).
K.2.3 ENVIRONMENTAL TRANSPORT OF RADIOACTIVE MATERIALS
Radioactive materials in degraded spent nuclear fuel and high-level radioactive waste could be
transported to the environment surrounding each storage facility by three pathways: groundwater,
surface-water runoff, and atmosphere. Figure K-6 shows the potential exposure pathways. The analysis
assumed that existing local climates would persist throughout the time of exposure of the spent nuclear
fuel and high-level radioactive waste to the environment. The assumed configuration for the degraded
storage facilities would have debris covering the radioactive material, which would remain inside the dry
storage canisters. While the dry storage canisters could fail sufficiently to permit water to enter, they
probably would retain their structural characteristics, thereby minimizing the dispersion of radioactive
particulate material to the atmosphere (Mishima 1998, page 4). Based on this analysis, the airborne
particulate pathway generally would not be an important source of human exposure. The assumption is
that after radionuclides dissolved in the precipitation they would reach the environment either through
groundwater or surface-water transport.
The analysis performed environmental fate and transport pathway modeling using the Multimedia
Environmental Pollutant Assessment System program (Buck et al. 1995, all). The Multimedia
Environmental Pollutant Assessment System is an integrated system of analytical, semianalytical, and
empirically based mathematical models that simulate the transport and fate of radioactive materials
through various environmental media and calculate concentrations, doses, and health effects at designated
receptor locations.
The Multimedia Environmental Pollutant Assessment System was originally developed by Pacific
Northwest National Laboratory to enable DOE to prioritize the investigation and remediation of the
Department's hazardous, radioactive, and mixed waste sites in a scientific and objective manner based on
readily available site information. The Multimedia Environmental Pollutant Assessment System has
evolved into a widely accepted (by Federal and international agencies) computational tool for calculating
the magnitude of environmental concentrations and public health impacts caused by releases of
radioactive material from various sources.
The following sections discuss the assumptions and methods used to determine radioactive material
transport for groundwater and surface-water pathways. Environmental parameters defined for input to the
K-14
Long-Term Radiological Impact Analysis for the No-Action Alternative
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K-15
Long-Term Radiological Impact Analysis for the No-Action Alternative
Multimedia Environmental Pollutant Assessment System program were collected from various sources
for specific sites (Sinkowski 1998, page 2) and regionalized parameters were developed (Poe and Wise
1998, all). The analysis used long-term averages to represent environmental conditions, and assumed that
these parameters would remain constant over the 10,000-year analysis period. The following sections
discuss the method for each pathwav.
K.2.3.1 Groundwater Transport
Precipitation falling on degrading spent nuclear fuel and high-level radioactive waste material would form
a radioactive solution (leachate) that could migrate through the vadose zone (the unsaturated upper layer
of soil) to the underlying water table, which would dilute, disperse, and transport the material
downgradient through the local aquifer system. As a result, there is a potential for human exposure
through the groundwater pathway to downgradient well users and to populations along surface-water
bodies where groundwater feeds into surface water.
The groundwater component of the radioactive material fluxes (infiltration) averaged over 70-year
(lifetime) increments was entered in the Multimedia Environmental Pollutant Assessment System
program. The infiltration would carry the contaminated leachate down through the vadose zone to the
saturated zone (aquifer). The contaminants would be diluted and dispersed as they traveled through the
aquifer. Radioactive material retardation would occur in both the unsaturated (above the water table) and
saturated (below the water table) zones. A distribution adsorption (that is, surface retention) coefficient,
K<i, (the amount of material adsorbed to soil particles relative to that in the water) modeled this retardation
(Toblin 1998a, page 2). This coefficient is radioactive material-specific and varies for each material
based on such factors as soil pH and clay content.
Table K-5 lists the adsorption coefficients, K<i, for the elements explicitly modeled for groundwater
transport. The coefficients are expressed as a function of the clay content of the soil through which the
elements are being transported; the analyses assumed a soil pH between 5 and 9. Note that the Ka values
of all isotopes of a given element (for example, plutonium-238, -239, and -240) are the same, because
adsorption is a chemical rather than nuclear process.
The time required to traverse the groundwater was determined for each radionuclide and 70-year period
(Toblin 1998a, page 4). Tables K-6 and K-7 list the range of nuclide groundwater transport times, from
source to receptor, for each of the five regions. Times are listed for the important nuclides (see
Table K-4). The analysis assumed that the vadose/aquifer flow fields were steady-state, so that the
nuclide travel times at a particular site would be constant over the 10,000-year analysis period, although
the nuclide release rates were not. Table K-6 lists parameters describing the total (over the analysis
period) and maximum nuclide release rates for the same important nuclides. Region 5, dominated by two
large DOE sites, is seen to result in the largest nuclide releases of all of the regions.
Table K-7 also lists the number of water systems and people that would obtain water from the affected
waterways. Many of these people would be subject to impacts from more than one site because they
would obtain their water from affected waterways downstream from multiple sites.
When the groundwater reached the point where it outcropped to surface water, radioactive material
transport would be subject to further dilution and dispersion. For most of the regions analyzed, the
distance between the storage location and the downgradient surface-water body would be inside the site
boundary; therefore, offsite wells generally would not be affected. However, the analysis calculated
groundwater concentrations for hypothetical onsite and offsite receptors. The Multimedia Environmental
Pollutant Assessment System program calculated groundwater and surface-water concentrations at each
receptor location for consecutive 70-year lifetimes in the 10,000-year analysis period.
K-16
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-5. Multimedia Environmental Pollutant Assessment System default
elemental equilibrium adsorption coefficients (Kj) for soil pH between 5 and 9.'
Clay content by weight
Element
< 10 percent
10 to 30 percent
> 30 percent
Actinium
228
538
4,600
Americium
82
200
1,000
Californium
0
0
0
Carbon
0
0
0
Cesium
51
249
270
Chlorine
0
0
0
Cobalt
2
9
200
Curium
82
200
1,000
Iodine
0
0
0
Krypton
0
0
0
Lead
234
597
1,830
Neptunium
3
3
3
Nickel
12
59
650
Niobium
50
100
100
Palladium
0
4
40
Plutonium
10
100
250
Protactinium
0
50
500
Radium
24
100
124
Ruthenium
274
351
690
Samarium
228
538
4,600
Selenium
6
15
15
Strontium
24
100
124
Technetium
3
20
20
Thorium
100
500
2,700
Tin
5
10
10
Tritium
0
0
0
Uranium
0
50
500
Zirconium
50
500
1,000
a. Source: Toblin (1998a, page 2).
The parameters necessary for the spent nuclear fuel and high-level radioactive waste storage sites for the
Multimedia Environmental Pollutant Assessment System were defined. Pertinent hydrologic and
hydrogeologic information was derived from the site-specific Updated Final Safety Analysis Reports for
commercial nuclear sites and site-specific data provided by the various DOE sites (Jenkins 1998, page 1).
Table K-8 lists the range (over the individual sites) in each region of the important hydrogeologic
parameters that would affect the transport of the radionuclides through the groundwater. These
parameters form the basis for the nuclide transport times listed in Table K-7.
A simplifying analytical assumption was that radioactive material transport would occur only through the
shallowest aquifer beneath the site. Because this assumption limits the interchange of groundwater with
underlying aquifers, less radioactive material dilution would occur, and groundwater pathway impacts
could be slightly overestimated. However, because impacts from the groundwater pathway would be
minor in comparison to surface-water pathways, the total estimated impacts would not be affected by this
assumption.
K-17
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-6. Regional source terms and environmental transport data for important isotopes used for
collective drinking water radiological impact analysis.'
Plutonium-
Parameter
239/240
Plutonium-238 Americium-241
I Americium-243
Neptunium-237 Technetium-99
Nuclide released in 10,000 years (curies)
Region 1
4,200
20
660
115
8.9
98
Region 2
17,000
97
1,500
240
32
1,200
Region 3
130,000
660
31,000
3,300
260
2,600
Region 4
4,300
17
450
110
9.0
89
Region 5
570,000
180
42,000
1,700
720
6,500
Maximum annual nuclide release (curies per year)
Region 1
19
0.020
1.2
0.053
0.0031
0.034
Region 2
53
0.035
2.2
0.11
0.0083
0.19
Region 3
60
0.71
56
1.6
0.092
1.0
Region 4
0.20
0.016
0.78
0.054
0.0034
0.035
Region 5
140
0.22
66
0.47
0.14
1.4
Years (from 2016) of maximum <
annual nuclide release
Region 1
1,435
1,435
1,435
1,435
1,435
1,435
Region 2
1,575
1,575
1,575
1,575
1,575
1,575
Region 3
1,155
1,155
1,155
1,155
1,155
1,155
Region 4
1,715
1,715
1,715
1,715
1,715
1,715
Region 5
875
875
875
875
875
875
Nuclide reaching receptors in 10,000 year (curies)
Region 1
3,600
11
130
43
8.8
95
Region 2
13,000
10
1.4
39
31
1,100
Region 3
110,000
250
380
510
250
2,500
Region 4
2,000
3.6
0.66
24
6.0
59
Region 5
180,000
2.6
0.020
1.2
630
5,600
Nuclide transport time" (years)
Region 1
10-5,500
10-5,500
10-45,000
10-45,000
10-1,700
10-1,700
Region 2
460-9,000
460-9,000
2,000-36,000
2,000-36,000
43-860
140-1,500
Region 3
65-45,000
65-45,000
410-260,000
410-260,000
31-9,800
31-9,800
Region 4
850-520,000
850-520,000
3,000-1,000,0003,000-1,000,000
59-16,000
130-100,000
Region 5
1,400-26,000
1,400-26,000
2,700-220,000
2,700-220,000
44-8,000
280-8,000
a. Source: Toblin (1998a, page 4).
b. Time from source to receptor.
Table K-7.
Transport and population data for drinking water pathway
impact analysis.
Parameter
Region 1
Region 2
Region 3 Regior
1 4 Region 5
Groundwater flow time (years)'
2.0 to 59
4.6 to 37
1.8 to 420 4.6 to 960 2.9 to 190
Number of people that would obtain domestic water 6.7
5.3
13.1
5.3
0.16
supply from affected waterways (millions)''
Affected drinking water systems
,C
>
112
147
137
64
23
a. From source to outcrop; source: adapted from Jenkins (1998, Table 2).
b. Source: Poe (1998b, page 12).
c. Source: Adapted from Sinkowski (1998, all).
K.2.3.2 Surface-Water Transport
The amount of leachate from degraded spent nuclear fuel and high-level radioactive waste in the surface-
water pathway would depend on soil characteristics and the local climate. The Multimedia
Environmental Pollutant Assessment System considers precipitation rates (Table K-2), soil infiltration,
evapotranspiration, and erosion management practices to determine the amount of leachate that would run
K-18
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-8. Multimedia Environmental Pollutant Assessment System regional groundwater input
parameters."
Parameter
Region 1
Region 2
Region 3
Region 4
Region 5
Vadose zone
Contaminated liquid infiltration
3.1-3.5
4.4
2.7-3.1
2.7 - 4.4
0.88-3.1
rate (vertical Darcy velocity) (feet
per year)''
Clay content (percent)
0-15
1 -47
1-47
3-15
1-15
pH of f)ore water
5-9
5-9
5-9
5-9
5-9
Thickness (feet)
6-50
10-50
4-160
2-80
23 - 250
Bulk density (grams per cubic
1.4- 1.9
1.4-1.6
1.4-1.6
1.4- 1.6
1.4-1.7
centimeter)
Total porosity (percent)
5-46
38-49
38-49
38-46
38-49
Field capacity (percent)
2.5 - 28
9-42
9-42
9-28
9-28
Saturated hydraulic conductivity
210-6,800
27 - 6,800
27 - 6,800
210 - 6.800
72 - 6,800
(feet per year)
Aquifer
Clay content (percent)
0-3
0-47
0-15
0-15
0-10
pH of pore water
5-9
5-9
5-9
5-9
5-9
Thickness (feet)
7-100
10-85
7-160
20 - 150
25 - 250
Bulk density (grams per cubic
1.6-2.1
1.4-2.0
1.5-1.7
1.4-1.7
1.5-1.9
centimeter)
Total porosity (percent)
5-38
5-49
5-44
5-46
23-44
Effective porosity (percent)
2.9 - 22
2.9 - 28
2.9 - 25
22-27
13-25
Saturated hydraulic conductivity
210-6,800
27 - 6,800
27 - 6,800
210-6,800
72 - 6,800
(feet per year)
Darcy velocity (feet per year)
6.8 - 1,400
12 - 170
3.9 - 430
0.58 - 270
33-560
Travel distance (feet)
1,900-5,600
2,000 - 4,700
1,900-23,000
1,600-12,000
1,900-37,000
a. Source: Adapted from Jenkins (1998, Table 2).
b. Annual precipitation rate (through degraded structure).
off rather than percolate into the soil. The contaminated runoff would travel overland and eventually
enter nearby rivers and streams that would dilute it further.
To determine the impacts of the contaminated discharge to surface water on the downstream populations
using that water (affected populations), DOE calculated the surface water flow rate and the release rate of
contaminants (as curies per year) contributed by each storage location draining to the surface water.
Using these values, DOE determined surface-water radionuclide concentrations for each receptor
location. DOE applied these concentrations to the respective affected populations to estimate impacts for
each region.
K.2.3.3 Atmospheric Transport
If degraded spent nuclear fuel or high-level radioactive waste was exposed to the environment, small
particles could become suspended in the air and transported by wind. The Multimedia Environmental
Pollutant Assessment System methodology includes formulations for radioactive material (particulate)
suspension by wind, vehicular traffic, and other physical disturbances of the ground surface. The impacts
from the atmospheric pathways would be small in comparison to surface-water pathways because the
cover provided by the degraded structures and the relatively large particle size and density of the
materials (see Section K.2.3) would preclude suspension by wind. Therefore, impacts from the transport
of radioactive particulate materials were not included in the analysis.
K.2.4 HUMAN EXPOSURE, DOSE, AND RISK CALCULATIONS
This section describes methods used in the No-Action Scenario 2 analysis to estimate dose rates and
potential impacts (latent cancer fatalities) to individuals and population groups from exposures to
K-19
Long-Term Radiological Impact Analysis for the No-Action Alternative
radionuclide contaminants in groundwater and surface water and in the atmosphere. As discussed above,
these contaminated environmental media would result from the degradation of storage facilities (Sections
K.2.I.1), corroding dry storage canisters (Section K.2.1.2), cladding failure (Section K.2.1.4), spent
nuclear fuel and high-level radioactive waste dissolution (Section K.2.1.5), leachate percolation and
groundwater transport (Section K.2.3.1), surface-water runoff (Section K.2.3.2), and atmospheric
suspension and transport (Section K.2.3.3).
For Scenario 1 and the first 1(X) years of Scenario 2, the presence of effective institutional control would
ensure that radiological releases to the environment and radiation doses to workers and the public
remained within Federal limits and DOE Order requirements and were maintained as low as reasonably
achievable. As a result, impacts to members of the public would be very small. Potential radiological
human health impacts that could occur would be due primarily to occupational radiation exposure of
onsite workers. The analysts estimated these impacts based on actual operational data from commercial
nuclear powerplant sites (NRC 1991a, pages 22 - 25) and projected these impacts for the 100- and
10,(XX)-year analysis periods for Scenario 1.
For Scenario 2, impacts to onsite workers and the public during institutional control (approximately
1(X) years) would be the same as those for Scenario 1. However, because the assumption for Scenario 2 is
that there would be no effective institutional control after approximately 1(X) years, engineered barriers
would begin to degrade and eventually would not prevent radioactive materials from the spent nuclear
fuel and high-level radioactive waste from entering the environment. During the period of no effective
institutional control, there would be no workers at the site. Thus, impacts were calculated only for the
public.
For Scenario 2, the potential highest exposures and dose rates over a 70-year lifetime period were
evaluated for individuals and exposed populations, hi addition, the total integrated dose to the exposed
population for the 10,(X)0-year analysis period was estimated. Human exposure parameters (exposure
times, ingestion and inhalation rates, agricultural activities, food consumption rates, etc.) were developed
based on recommendations from Federal agencies (EPA 1988, pages 113 to 131; EPA 1991,
Attachment B; NRC 1977, pages 1.109-1 to 1.109-2; Shipers and Harlan 1989, all; NRC 1991b,
Chapter 6) and are reflected as Multimedia Environmental Pollutant Assessment System default values
(Buck et al. 1995, Section 1.0). Other parameters chosen for this analysis are summarized in supporting
documentation (Sinkowski 1998, all; Toblin 1998a,b,c, all). Table K-9 lists the exposure and usage
parameters for all of the pathways considered in the analysis (see Section K.3.1).
The Scenario 2 analysis evaluated long-term radiation doses and impacts to populations exposed through
the surface-water and groundwater pathways. This analysis estimated population impacts only for the
drinking water pathway using regionalized effective populations and surface-water dilution factors
discussed in Section K.2.3.2. Other pathways were evaluated to determine their potential contribution in
relation to drinking water doses. These analyses are discussed in Section K.3.1.
K.2.4.1 Gardener Impacts
To reasonably bound human health impacts resulting from human intrusion, two types of gardener were
evaluated — the onsite gardener (10 meters [33 feet]) from the degrading storage facility) and the near-site
gardener (5 kilometers [3 miles] from the degrading facility). The analysis had both of these hypothetical
gardeners residing on the flow path for groundwater. The gardeners would obtain all their drinking water
from contaminated groundwater, grow their subsistence gardens in contaminated soils, and irrigate them
with the contaminated groundwater. The contaminated garden soils, suspended by the wind, would
contaminate the surfaces of the vegetables consumed by the gardeners. The hypothetical onsite gardener
would be the maximally exposed individual.
K-20
I
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-9. Multimedia Environmental Pollutant Assessment System human exposure
input parameters for determination of all pathways radiological impacts sensitivity
analysis (page 1 of 2).'
Water source'' Surface water
Domestic water supply treatment' Yes
Fraction of plutonium removed by water treatment'' 0.3
Drinking water rate (liters per day per person)' 2
Irrigation rate (liters per square meter per month)*^ 100
Leafy vegetable consumption rate (kilograms per day per person)* 0.021
Other vegetable consumption rate (kilograms per day per person) 0. 1 3
Meat consumption rate (kilograms per day per person) 0.065
Milk consumption rate (kilograms per day per person) 0.075
Finfish consumption rate (kilograms per day per person) 0.0065
Shellfish consumption rate (kilograms per day per person) 0.0027
Shoreline contact (hours per day per person) 0.033
Americium ingestion dose conversion factor (rem per picocurie)'' 3.6x10*
Americium finfish bioaccumulation factor 250
Americium shellfish bioaccumulation factor 1,000
Americium meat transfer factor (days per kilogram) 3.5x10"'
Americium milk transfer factor (days per liter) 4.0x10''
Neptunium ingestion dose conversion factor (rem per picocurie) 4.4x10*
Neptunium finfish bioaccumulation factor 250
Neptunium shellfish bioaccumulation factor 400
Neptunium meat transfer factor (days per kilogram) 5.5x10"'
Neptunium milk transfer factor (days per liter) 5.0x10"*
Technetium ingestion dose conversion factor (rem per picocurie) 1 .5x10"'
Technetium finfish bioaccumulation factor 15
Technetium shellfish bioaccumulation factor 5
Technetium meat transfer factor (days per kilogram) 8.5x10"'
Technetium milk transfer factor (days per liter) 1.2x10"^
Plutonium ingestion dose conversion factor (rem per picocurie) 3.5x10"*
Plutonium finfish bioaccumulation factor 250
Plutonium shellfish bioaccumulation factor 100
Plutonium meat transfer factor (days per kilogram) 5.0x10"'
Plutonium milk transfer factor (days per liter) 1x10"'
Yield of leafy vegetables [kilograms (wet) per square meter] 2.0
Yield of vegetables [kilograms (wet) per square meter] 2.0
Yield of meat feed crops [kilograms (wet) per square meter] 0.7
Yield of milk animal feed crops [kilograms (wet) per square meter] 0.7
Meat animal intake rate for feed (liters per day) 68
Milk animal intake rate for feed (liters per day) 55
Meat animal intake rate for water (liters per day) 50
Milk animal intake rate for water (liters per day) 60
Agricultural areal soil density (kilograms per square meter) 240
Retention fraction of activity on plants 0.25
Translocation factor for leafy vegetables 1 .0
Translocation factor for other vegetables 0. 1
Translocation factor for meat animal 0. 1
Translocation factor for milk animal 1 .0
Fraction of meat feed contaminated 1 .0
Fraction of milk feed contaminated 1 .0
Fraction of meat water contaminated 1 .0
Fraction of milk water contaminated 1 .0
Meat animal soil intake rate (kilograms per day) 0.5
K-21
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-9. Multimedia Environmental Pollutant Assessment System human exposure
input parameters for determination of all pathways radiological impacts sensitivity
analysis (page 2 of 2).'
Water source'' Surface water
Milk animal soil intake rate (kilograms per day) 0.5
Leafy vegetable growing period (days) 60
Other vegetable growing period (days) 60
Beef animal feed growing period (days) 30
Milk animal feed growing period (days) 30
Water intake rate while showering (liters per hour) 0.06
Duration of shower exposure (hours per shower) 0.167
Shower frequency (per day) 1 .0
Thickness of shoreline sediment (meters) 0.04
Density of shoreline sediments (grams per cubic meter) 1 .5
Shore width factor for shoreline external exposure 0^2
a. Source: Bucket al. (1995, MEP AS default settings).
b. Groundwater for gardener.
c. No for gardener.
d. Zero for gardener.
e. To convert liters to gallons, multiply by 0.26418.
f To convert liters per square meter to gallons per square foot, multiply by 0.00025.
g. To covert kilograms to pounds, multiply by 2.2046.
h. Sediment ingestion = 0. 1 grams per hour (0.000022 pounds per hour) during contact,
i. For plutonium-239/240.
HUMAN INTRUSION
Spent nuclear fuel and high-level radioactive waste in surface or below^-grade storage facilities would
be readily accessible in the absence of institutional control. For this reason, DOE anticipates that
both planned and inadvertent intrusions could occur. An example of the former would be the
scavenger who searches through the area seeking articles of value; an example of the latter would
be the farmer who settles on the site and grows agricultural crops with no knowledge of the storage
structure beneath the soil. Intrusions into contaminated areas also could occur through activities
such as building excavations, road construction, and pipeline or utility replacement.
Under the conditions of Scenario 2, intruders could receive external exposures from stored spent
nuclear fuel and high-level radioactive waste that would grossly exceed current regulatory limits and,
in some cases, could be sufficiently high to cause prompt fatalities. In addition, long-term and
repeated intrusions, such as those caused by residential construction or agricultural activities near
storage sites, could result in long-term chronic exposures that could produce increased numbers of
latent cancer fatalities. These intrusions could also result in the spread of contamination to remote
locations, which could increase the total number of individuals potentially exposed.
Calculations were performed using transport models described by Buck et al. (1995, all) for gardeners in
each of the five analysis regions using regionalized source terms and environmental parameters.
Therefore, calculated impacts to the regional gardener (maximally exposed individual) would not
represent the highest impacts possible from a single site in a given region, but rather would reflect an
average impact for the region. Details of the analysis are provided in Toblin (1998c, all). The regional
hydrogeologic parameters listed in Table K-10, together with transient nuclide release rates (the
maximum of which is indicated in the table), were used to determine the radiological impacts to the
regional gardener as a result of groundwater transport. The regional parameters were based on a curie-
weighting of the individual site parameters for plutonium and americium. The exposure parameters in
K-22
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-10. Multimedia Environmental Pollutant Assessment System groundwater transport input
parameters for estimating radiological impacts to the onsite and near-site gardener.'
Parameter
Region 1 Region 2 Region 3 Region 4 Region 5
1
10
12
11
2
5-9
5-9
5-9
5-9
5-9
11
44
7.1
43
180
0.11
0.44
0.071
0.43
1.8
1.6
1.5
1.5
1.5
1.6
38
42
44
45
41
9.3
15
23
21
12
6,500
660
1,700
1,000
5,900
1.8
6.5
1.2
4.4
0.69
5-9
5-9
5-9
5-9
5-9
45
50
37
64
210
1.6
1.8
1.6
1.6
1.7
38
40
38
35
30
22
23
22
20
17
340
62
69
51
300
f(xr
f(x)
f(x)
f(x)
f(x)
f(x)^3
f(x)^3
f(x)^3
f(x)-3
f(x)-r3
f(x) -r 400
f(x) H- 400
f(x) ^ 400
f(x) ^ 400
f(x) -r 400
4.9
0.24
3.8
0.32
2.1
Vadose zone
Contaminated liquid infiltration rate (vertical Darcy 3.5 4.4 2.7 3.5 0.88
velocity) (feet per year)''''
Clay content (percent)
pH of pore water
Thickness (feet)
Longitudinal dispersivity (feet)
Bulk density (grams per cubic meter)*"
Total porosity (percent)
Field capacity (percent)
Saturated hydraulic conductivity (feet per year)
Aquifer
Clay content (percent)
pH of poK water
Thickness (feet)
Bulk density (grams per cubic meter)
Total porosity (percent)
Effective porosity (percent)
Darcy velocity (feet per year)
Longitudinal dispersivity (feet)
Lateral dispersivity (feet)
Vertical dispersivity (feet)
Maximum annual plutonium-239 and -240 release
(curies per year)
Years (from 2016) of maximum annual plutonium 1,365 1,575 1,155 1,715 875
release ^__^
a. Source: Toblin (1998c, page 2-4).
b. Annual precipitation rate (through degraded structure).
c. To convert feet to meters, multiply by 0.3048.
d. To convert grams per cubic meter to pounds per cubic foot, multiply by 0.0000624.
e. f(x) = 2.72 X (logioO.3048 x x)^"'*, where x = downgradient distance.
Table K-9 describe the radionuclide exposure to the gardener where applicable (for example, exposure
parameters related to the fish are not appUcable to the gardener).
K.2.4.2 Direct Exposure
The analysis evaluated potential external radiation dose rates to the maximally exposed individual for a
commercial independent spent fuel storage installation because this type of facility would provide the
highest external exposures of all the facilities analyzed in this appendix. Maximum dose rates over the
10,000-year analysis period were evaluated for each region. The maximally exposed individual was
assumed to be 10 meters (about 33 feet) from an array of concrete storage modules containing 1,000
MTHM of commercial spent nuclear fuel. The maximum dose rate varied between regions depending on
how long the concrete shielding would remain intact (Table K-1).
The direct gamma radiation levels were calculated (Davis 1998, page 1). To ensure consistency between
this analysis and the Total System Performance Assessment, the same radionuclides were used for the
design of the Yucca Mountain Repository surface facility shielding (TRW 1995, Attachment 9.5).
Radionuclide decay and radioactive decay product ingrowth over the 10,000-year analysis period were
calculated using the ORIGEN computer program (ORNL 1991, all).
K-23
Long-Term Radiological Impact Analysis for the No-Action Alternative
Neutron emissions were not included because worst-case impacts (death within a short period of
exposure) would be the same with or without the neutron component. Details of these calculations and
analyses are provided in supporting documentation (Rollins 1998b, all).
K.2.5 ACCIDENT METHODOLOGY
Spent nuclear fuel and high-level radioactive waste stored in above-ground dry storage facilities would be
protected initially by the robust surrounding structure (either metal or concrete) and by a steel storage
container that contained the material. Normal storage facility operations would be primarily passive
because the facilities would be designed for cooling via natural convection. DOE evaluated potential
accident and criticality impacts for both Scenario 1 (institutional control for 10,000 years) and Scenario 2
(assumption of no effective institutional control after approximately 1(X) years with deterioration of the
engineered barriers initially protecting the spent nuclear fuel or high-level radioactive waste).
For Scenario 1, human activities at each facility would include surveillance, inspection, maintenance, and
equipment replacement when required. The facilities and the associated systems, which would be
licensed by the Nuclear Regulatory Commission, would have certain required features. License
requirements would include isolation of the stored material from the environment and its protection from
severe accident conditions (10 CFR 50.34). The Nuclear Regulatory Commission requires an extensive
safety analysis that considers the impacts of plausible accident-initiating events such as earthquakes, fires,
high winds, and tornadoes. No plausible accident scenarios have been identified that result in the release
of radioactive material from the storage facilities (PGE 1996, all; CP&L 1989, all). In addition, the
license would specify that facility design requirements include features to provide protection from the
impacts of severe natural events. This analysis assumed maintenance of these features indefinitely for the
storage facilities.
DOE performed a scoping analysis to identify the kinds of events that could lead to releases of radioactive
material to the environment prior to degradation of concrete storage modules and found none. The two
events determined to be the most challenging to the integrity of the concrete storage modules would be
the crash of an aircraft into the storage facility and a severe seismic event.
• Davis, Strenge, and Mishima (1998, all) concluded that the postulated aircraft crash would be
potentially more severe than a postulated seismic event because storage facility damage from an
aircraft crash probably would be accompanied by a fire that could heat the spent nuclear fuel or high-
level radioactive waste and increase the quantity of material released to the environment. The
analysis showed that hurtling aircraft components produced by such an event would not penetrate the
storage facility and that a subsequent fire would not result in a release of radioactive materials.
• For the seismic event, meaningful damage would be unlikely because storage facilities would be
designed to withstand severe earthquakes. Even if such an event caused damage, no immediate
release would occur because no mechanism has been identified that would cause meaningful fuel
pellet damage to create respirable airborne particles. If this damage did not occur, the source term
would be limited to gaseous fission products, carbon-14, and a very small amount of preexisting fuel
pellet dust. Subsequent repairs to damaged facilities or concrete storage modules would preclude the
long-term release of radionuclides.
Criticality events are not plausible for Scenario 1 because water, which is required for criticality, could
not enter the dry storage canister. The water would have to penetrate several independent barriers, all of
which would be maintained and replaced as necessary under Scenario 1 .
K-24
Long-Term Radiological Impact Analysis for the No-Action Alternative
Under Scenario 2, facilities would degrade over time and the structures would gradually deteriorate and
lose their integrity. The analysis determined that two events, an aircraft crash and inadvertent criticality,
would be likely to dominate the impacts from accidents, as described in the following paragraphs.
K.2.5.1 Aircraft Crash
DOE determined that an aircraft crash into a degraded concrete storage module would be the largest
plausible accident-initiating event that could occur at the storage sites. This event would provide the
potential for the airborne dispersion of radioactive material to the environment and, as a result, the
potential for exposure of individuals who lived in the vicinity of the site. The aircraft crash could result in
mechanical damage to the storage casks and the fuel assemblies they contained, and a fire could result.
The fire would provide an additional mechanism for dispersion of the radioactive material. The
frequency and consequences of this event are described in detail in Davis, Strenge and Mishima (1998,
all).
The aircraft assumed for the analysis is a midsize twin-engine commercial jet (Davis, Strenge, and
Mishima 1998, page 2). The area affected by a crash was computed using the DOE standard formula
(DOE 1996, Chapter 6) in which the aircraft could crash directly into the side or top of the concrete
storage modules, or could strike the ground in the immediate vicinity of the facility and skid into the
concrete storage modules. Using this formula, the dimensions of a typical storage facility as shown in
Chapter 2, Figure 2-37, and the aircraft configuration would result in an estimated aircraft crash
frequency of 0.(XXXX)32 (3 in 1 million) crashes per year (Davis, Strenge, and Mishima 1998, page 5).
This frequency is within the range that DOE typically considers the design basis, which is defined by
DOE as 0.000001 or greater per year (DOE 1993, page 28).
The analysis estimated the consequences of the aircraft crash on degraded concrete storage modules. The
twin-engine jet was assumed to crash into an independent spent fuel storage installation that contained
100 concrete storage modules, each containing 24 pressurized-water reactor fuel assemblies. Using the
penetration methodology from DOE (1996, Chapter 6), an aircraft crash onto these concrete storage
modules could penetrate 0.8 meter (2.6 feet). Because the concrete storage modules have 1.2-meter
(3.9-foot) thick walls, the crash projectiles would not penetrate the reinforced concrete in the as-
constructed form. Thus, DOE determined that the aircraft crash would not cause meaningful
consequences until the concrete storage modules were considerably degraded, when an aircraft projectile
could penetrate a concrete storage module and damage a storage cask (Davis, Strenge, and Mishima 1998,
page 7). The degradation process is highly location-dependent, as noted in Section K.2.1.1. For sites in
northern climates, the degradation would be relatively rapid due to the freeze/thaw cycling that would
expedite concrete breakup; considerable degradation could occur in 2(X) to 3(X) years. For southern
climates, the degradation would be much slower. Thus, an aircraft crash probably would not result in
meaningful consequences for a few hundred to a few thousand years, depending on location. The timing
is of some importance because the radioactive materials in the fuel would decay over time, and the
potential for radiation exposure would decline with the decay.
The analysis assumed that the aircraft crash occurred 1,000 years after the termination of institutional
control at a facility where the concrete had degraded sufficiently to allow breach of the dry storage
canister. Computing public impacts from the air crash event requires estimating the population to a
distance of 80 kilometers (50 miles) from a hypothetical site (the distance beyond which impacts from an
airborne release would be very small). This analysis considered two such sites, one in an area of a high
population site and one in an area of low population. The average population around all of the sites in
each of the five regions defined in Figure K-2 was computed based on 1990 census data. The average
ranged from a high of 330 persons per square mile in region 1 (high population) to a low of 77 persons
K-25
Long-Term Radiological Impact Analysis for the No-Action Alternative
per square mile in region 4 (low population). Both of these population densities (assumed to be uniform
around the hypothetical sites) were used in the consequence calculation.
Estimating the amount of airborne respirable particles that would result from a crash requires assumptions
about the impact and resulting fire. The impact of the jet engines probably would cause extensive damage
to the fuel assemblies in the degraded concrete storage module, and would scatter fuel pins around the
immediate area. The fuel tanks in the aircraft would rupture, and fuel would disperse around the site and
collect in pools. These pools would ignite, and an intense fire [hotter than 5(X)°C (approximately 930°F)]
(Davis, Strenge, and Mishima 1998, page 8) would result. The fire would heat the fuel pins to the point
of cladding rupture. The ruptured fuel pins would cause fuel pellets to be exposed to the fire. As the fire
burned, the fuel pools would recede, exposing additional fuel pellets to the air. This would cause
oxidation of the hot uranium dioxide fuel pellets, converting them to UsOg (another form of uranium
oxide), which would produce a large amount of fuel pellet dust, including small particles that could
become airborne and inhaled into the lungs. The estimated fraction of the fuel converted to respirable
airborne dust would be 0.12 percent (Davis, Strenge and Mishima 1998, page 9). The fire would cause a
thermal updraft that could loft the fuel pellet dust into the atmosphere.
The consequences from the event were computed with the MACCS2 program (Rollstin, Chanin, and Jow
1990, all). This model has been used extensively by the Nuclear Regulatory Commission and DOE to
estimate impacts from accident scenarios involving releases of radioactive materials. The model
computes dose to the public from the direct radiation by the cloud of radioactive particles released during
the accident, from inhaling particles, and from consuming food produced from crops and grazing land that
could be contaminated as the particles are deposited on the ground from the passing cloud. The food
production and consumption rates are based on generic U.S. values (Kennedy and Strenge 1992, pages
6.19 to 6.28; Chanin and Young 1998, all). The program computes the dispersion of the particles as the
cloud moves downwind. The dispersion would depend on the weather conditions (primarily wind speed,
stability, and direction) that existed at the time of the accident. This calculation assumed median weather
conditions and used annual weather data from airports near the centers of the regions.
K.2.5.2 Criticality
DOE evaluated the potential for nuclear criticality accidents involving stored spent nuclear fuel. A
criticality accident is not possible in high-level radioactive waste because most of the fissionable atoms
were removed or the density of fissionable atoms was reduced by the addition of glass matrix. Nuclear
criticality is the generation of energy by the fissioning (splitting) of atoms as a result of collisions with
neutrons. The energy release rate from the criticality event can be very low or very high, depending on
several factors, including the concentration of fissionable atoms, the availability of moderating materials
to slow the neutrons to a speed that enables them to collide with the fissionable atoms, and the presence of
materials that can absorb neutrons, thus reducing the number of fission events.
Criticality events are of concern because under some conditions they could result in an abrupt release of
radioactive material to the environment. If the event were energetic enough, the dry storage canister
could split open, fuel cladding failure could occur, and fragmentation of the uranium dioxide fuel pellets
could occur.
The designs of existing dry storage systems for spent nuclear fuel, in accordance with Nuclear Regulatory
Commission regulations (10 CFR Part 72) preclude criticality events by various measures, including
primarily the prevention of water entering the dry storage canister. If water is excluded, a criticality
cannot occur.
K-26
Long-Term Radiological Impact Analysis for the No-Action Alternative
If institutional control was maintained at the dry storage facilities (Scenario 1), a criticality is not
plausible because the casks would be monitored and maintained such that introduction of water into the
canister would not be possible. However, under Scenario 2, eventual degradation (corrosion) of the dry
storage canisters could lead to the entry of water from precipitation, at which point criticality could be
possible if other conditions were met simultaneously.
The analysis considered three separate criticality events:
• A low-energy event that involved a criticality lasting over an intermediate period (minutes or more).
This event would not produce high temperatures or generate large additional quantities of
radionuclides. Thus, no fuel cladding failures and no meaningful increase in consequences would be
likely.
• An event in which a system went critical but at a slow enough rate so the energy release would not be
large enough to produce steam, which would terminate the event. This event could continue over a
relatively long period (minutes to hours), and would differ from the low-energy event in that the total
number of fissions could be very large, and a large increase in radionuclide inventory could result.
This increase could double the fission product content of the spent nuclear fuel. No fuel cladding
failures would be likely in this event, so no abrupt release of radionuclides would occur.
• An energetic event in which a system went critical and produced considerable fission energy. This
event could occur if seriously degraded fuel elements collapsed abruptly to the bottom of the canister
in the presence of water that had penetrated the canister. This event would produce high fuel
temperatures that could lead to cladding rupture and fuel pellet oxidation. The radiotoxicity of the
radionuclide inventory produced by the fission process would be comparable to the inventory in the
fuel before the event.
The probability of a criticality occurring as described in these scenarios is highly uncertain. However,
DOE expects the probability would be higher for the first two events, and much lower for the third
(energetic energy release). Several conditions would have to be met for any of the three events to occur.
The concrete storage module and dry storage canister must have degraded such that water could enter but
not drain out. The fuel would have to contain sufficient fissionable atoms (uranium-235, plutonium 239)
to allow criticality. This would depend on initial enrichment (initial concentration of uranium-235) and
bumup of the fuel in the reactor before storage (which would reduce the uranium-235 concentration).
Because a small amount of spent nuclear fuel would be likely to have appropriate enrichment bumup
combinations that could enable criticality to occur, none of the criticality events can be completely ruled
out. The energetic criticality event is the only one with the potential to produce large impacts. Such an
event would be possible, but would be highly unlikely; its consequences would be uncertain. The event
could cause a prompt release of radionuclides. However, the amount released would not be likely to
exceed that released by the aircraft crash event evaluated above. Thus, this analysis did not evaluate
specific consequences of a criticality event.
K.3 Results
K.3.1 RADIOLOGICAL IMPACTS
Impacts to human health from long-term environmental releases and human intrusion were estimated
using the methods described in Section K.2 and in supporting technical documents (Sinkowski 1998, all;
Jenkins 1998, all; Battelle 1998, all; Poe I998a,b, all; Poe and Wise 1998, all; Toblin 1998a,b,c, all). The
radiological impacts on human health would include internal exposures due to the intake of radioactive
materials released to surface water and groundwater.
K-27
Long-Term Radiological Impact Analysis for the No-Action Alternative
Six of the seven radionuclides listed in Table
K-4 would contribute more than 99 percent of
the total dose. Table K-1 1 lists the estimated
radiological impacts by region during the last
9,900 years under Scenario 2 fcM" the Pr<qx)sed
Action and Module 1 inventories of spent
nuclear fuel and high-level radioactive waste.
As noted above, these inlets would be to the
public firom drinking water from the major
waterways contaminated by surface-water
runoff of radioactive materials from degraded
spent nuclear fuel and high-level radioactive
waste storage facilities (Toblin 1998a,b, all).
Figure K-7 shows the locations of all
commercial nuclear and DOE waste storage
sites in the United States and more than 20
potentially affected major waterways. At
present, 30.5 million people are served by
municipal water systems with intakes al(Mig
the potentially affected pcations of these
waterways. Over the 9,900-year analysis
period, about 140 generations would be potentially
to occur during about the first 1,000 years for most
high as 3.9 billion.
SCENARIO 2 IMPACTS
The principal tong-term human health
consequences from the storage of spent nuclear
fuel and high-level radioactive waste would result
from rainwater flowing through degraded storage
facilities where it would dissolve the material. The
dissolved material would travel through
groundwater and surface-water runoff to rivers
and streams where people could use it for
domestic purposes such as drinking water and
crop irrigation. The Scenario 2 analysis estimated
population impacts resulting only from the
consumptkxi of contaminated drinking water and
exposures resulting from land contamination due
to periodic flooding, although other pathways,
such as eating contaminated fish, could contribute
additional impacts larger than those from drinking
water for selected indivkluals in tfie exposed
populatk>n.
affected. However, because releases are not estimated
regions, the potential affected population could be as
Table K-11. Estimated collective radiological impacts to the public fixjm continued stwage of Proposed
Action and Module 1 inventcmes of spent nuclear fiiel and high-level radioactive waste at commercial and
DOE storage facilities - Scenario 2.*
9,900-year population dose*"
(person-
rem)
9,900-year
LCFs'
Years until peak impact''
Region
Proposed Action
Module 1
Proposed Action
Module 1
Proposed Action Module 1
1
1,800.000
1,820,000
900
900
1,400 1,400
2
760,000
1,260,000
380
630
5,100 8,300
3
3,500,000
3,650,000
1,800
1,830
3.400* 3,400*
4
70.000
138,000
30
69
3.900 3,900
5
460,000
461,000
230
230
7,100 7,000
Totals
6^90,000
7330J)00
3340
3,700
a. Total population (collective) dose from drinking water pathway over 9,900 years.
b. LCF = latent cancer fatality; additional number of latent cancer fatalities for the exposed population group based on an
assumed risk of 0.0005 latent cancer fatality per person-rem of collective dose (NCRP 1993a, page 1 12).
c. Yearsafter21i6 when the maximum doses would occur.
d. Year of combined U.S. peak impact would be the same as for Region 3 peak impact because the predominant impact would
be in Region 3.
Table K-11 indicates the variability of individual doses and potential inlets in the five regions analyzed
(see Section K.2.1.6). The variability among regions is due to differences in types and quantities of spent
nuclear fuel and high-level radioactive waste, annual precipitation, size of affected populations, and
surface-water bodies available to transport the radioactive material.
Table K-11 also indicates that the Proposed Action inventory would produce a collective drinking water
dose of 6.6 million person-rem over 9,900 years, which could result in an additional 3,300 latent cancer
K-28
Long-Term Radiological Impact Analysis for the No-Action Alternative
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K-29
Long-Term Radiological Impact Analysis for the No-Action Alternative
fatalities in the total potentially exposed population of 3.9 billion, in which about 900 million fatal
cancers [using the lifetime fatal cancer risk of 24 percent (NCHS 1993, page 5)] would be likely to occur
from all other causes. Figures K-8 and K-9 show the Proposed Action inventory regional collective doses
and potential latent cancer fatalities, respectively, for approximately 140 consecutive 70-year lifetimes
that would occur during the 9,900-year analysis period. The peaks shown in Figures K-8 and K-9 would
result from the combination of the sites that drain to the Mississippi River and the relatively large
populations potentially affected along these waterways. These values include impacts for the Proposed
Action inventory only. Similar curves for the Module 1 inventory are not shown because of their
similarity to those for the Proposed Action inventory. As listed in Table K-1 1, the impacts from the
Module 1 inventory would be approximately 20 percent greater than for the Proposed Action inventory.
The additional 3,300 Proposed Action latent cancer fatalities (or 3,700 Module 1 latent cancer fatalities)
over the 10,000-year analysis period would not be the only negative impact. Under Scenario 2, more than
20 major waterways of the United States (for example, the Great Lakes, the Mississippi, Ohio, and
Columbia rivers, and many smaller rivers along the Eastern Seaboard) that currently supply domestic
water to 30.5 million people would be contaminated with radioactive material. The shorelines of these
waterways would be contaminated with long-lived radioactive materials (plutonium, uranium, americium,
etc.) that would result in exposures to individuals who came into contact with the sediments, potentially
increasing the number of latent cancer fatalities. Each of the 72 commercial and 5 DOE sites throughout
the United States would have potentially hundreds of acres of land and underlying groundwater systems
contaminated with radioactive materials at concentrations that would be potentially lethal to anyone who
settled near the degraded storage facilities. The radioactive materials at the degraded facilities and in the
floodplains and sediments would persist for hundreds of thousands of years.
As mentioned above, DOE only estimated potential collective impacts resulting from the consumption of
contaminated surface water. However, other pathways (food consumption, contaminated floodplains,
etc.) that could contribute to collective dose were evaluated (Toblin 1998b, all; Rollins 1998c, all) to
determine their relative importance to the drinking water pathway. These pathways included the
following:
• Consumption of vegetables irrigated with contaminated water
• Consumption of meat and milk from animals that drank contaminated water or were fed with
contaminated feed
• Consumption of contaminated fmfish and shellfish
• Direct exposure to contaminated shoreline sediments
• Exposures resulting from contamination of floodplains during periods of high stream (river) flow
These analyses determined that an individual living in a contaminated floodplain and consuming
vegetables irrigated with contaminated surface water could receive a radiation exposure dose three times
higher than that from the consumption of contaminated surface water only (Toblin 1998b, page 3). In
addition, the analysis determined that impacts to 30 million individuals potentially living in contaminated
floodplains would be less than 10 percent of the collective impacts shown in Figure K-9 and, therefore,
did not include them in the estimates because DOE did not want to overestimate the impacts from
Scenario 2.
DOE evaluated airborne pathways (Mishima 1998, all) and judged that potential impacts from those
pathways would be very small in comparison to impacts from liquid pathways because the degraded
facility structures would protect the radioactive material from winds. To simplify the analysis, impacts to
K-30
Long-Term Radiological Impact Analysis for the No-Action Alternative
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K-31
Long-Term Radiological Impact Analysis for the No-Action Alternative
the public from radiation emanating from the degraded storage facilities were not included. Those
impacts were judged to represent a small fraction of the impacts calculated for the liquid pathways
(Table K-11).
Estimates of localized impacts (Toblin 1998c, page 1) assumed that individuals (onsite and near-site
gardeners) would take up residence near the degraded storage facilities and would consume vegetables
from their gardens irrigated with groundwater withdrawn from the contaminated aquifer directly below
their locations. In addition, the onsite gardener would be exposed to external radiation emanating from
the exposed dry storage canisters; therefore, the onsite gardener would be the maximally exposed
individual.
Table K-12 lists the internal estimated dose rates (see Section K.2.4.1 for details) and the times for peak
exposure for each of the five regions.
Table K-12. Estimated internal dose rates (rem per year) and year of peak exposure* (in parentheses) for
the onsite and near-site gardeners - Scenario 2}
Maximally exposed individual distances (meters)*^ from storage facilities
Region 10^ 150 1,000 5,000
1
3,100(1,800)
670 (2,200)
51 (2,000)
12 (2,600)
2
100 (2,700)
96 (2,000)
12 (2,900)
2 (7,100)
3
3,100(1,800)
1,800 (2,000)
150 (2,600)
31 (6,000)
4
140 (3,200)
130 (3,900)
14 (4,800)
2 (9,300)
5
3,300 (4,600)
180(5,300)
59 (5,300)
2 (6,100)
a. Years after facility maintenance ended.
b. Source: Adapted from Toblin (1998c, Table 4, page 5).
c. To convert meters to feet, multiply by 3.2808.
d. The maximally exposed individual would be the onsite gardener.
The regional dose rates listed in Table K-12 would depend on the concentration of contaminants
(primarily plutonium) in the underlying aquifer from which water was extracted and used by the gardener
for consumption and crop irrigation. These aquifer concentrations, in turn, would be affected by the type
and location of stored materials (spent nuclear fuel and high-level radioactive waste) in each region, the
rate at which the contaminants were leached from the stored material, the amount of water (precipitation)
available for dilution, and the thickness of the aquifer. For example, releases in Region 5 would probably
be smaller and would occur later than those in other regions because of the region's lack of precipitation.
This is indeed the case for commercial fuel, which is stored in above-grade concrete storage modules,
stainless-steel dry storage canisters, and mostly intact corrosion-resistant zirconium alloy cladding.
However, early releases would occur in Region 5 because most DOE spent nuclear fuel is stored in
below-grade vaults (see Appendix A, page A-25) that would stop providing rain protection after 50 years
(see Section K.2.1.1 for details). In addition, the analysis assumed no credit for the protectiveness of the
DOE spent nuclear fuel cladding (see Section K.2. 1.4.2 for details), which would result in releases that
began early (about 8(X) years after weather protection was lost) and persist at a nearly constant rate for
more that 6,000 years (Toblin 1998c, page 3).
The 10-meter (33-foot) doses listed in Table K-12 would be due to leachate concentrations from the
storage area with no groundwater dilution. Downgradient doses decrease more rapidly in Regions 1 and 5
than in other regions because of greater groundwater dilution. The downgradient decrease in Region 5
would also be due to the relatively thick aquifer, which results in greater vertical plume spread and
increases plume attenuation (Toblin 1998c, pages 4-6).
As shown in Table K-12, an onsite gardener in Region 5 could receive an internal committed dose as high
as 3,300 rem for each year of ingestion of plutonium-239 and -240. However, the individual actually
K-32
Long-Term Radiological Impact Analysis for the No-Action Alternative
would receive only about 70 rem the first year, 140 rem the second year, 210 rem the third year, and so on
until reaching an equilibrium annual dose (in approximately 50 years) of 3,300 rem per year. The
individual would continue to receive this equilibrium dose as long as the radioactive material uptake
remained constant.
If the annual doses are added, in less than 10 years the individual would have received more than 2,000
rem. If the International Commission on Radiological Protection risk conversion factor were applied to
this dose, a probability of fatal cancer induction of 1 could be calculated. In other words, the use of this
risk conversion would predict that the individual would contract a fatal cancer after 10 years of exposure.
This calculated risk is approximately 4 times greater than the lifetime risk of contracting a fatal cancer
from all other causes (24 percent).
Table K-13 shows that the direct radiation dose rate to the onsite gardener could be as high as 7,3(X) rem
per year. Unlike internal dose, this dose would actually be delivered during the year of exposure. This
maximum value assumes a complete loss of shielding normally provided by the concrete storage module
at the same time as the loss of weather protection (see Table K-1). Assuming a dose of 7,300 rem per
year, the individual probably would die from acute radiation exposure. This dose would probably cause
extensive cell damage in the individual that would result in severe acute adverse health conditions and
death within weeks or months (NRC 1996, page 8.29-5). However, these higher radiation dose rates are
based on an early estimated time to structural failure of the concrete storage module. If these failure
times were extended by as little as 1(X) years, the associated dose rates would decrease by a factor of
10 because the levels of radiation emanating from the degraded facilities would have decreased by about a
factor of 10 due to radioactive decay (Rollins 1998c, page 12).
Table K-13. Estimated extemal peak dose rates (rem per year) for the onsite and near-site gardeners -
Scenario 2.
Year of peak exposure""
Maximally exposed individual distances (meters)'
from
storage facilities
Region
10'
150
1,000
5,000
1
190
7,200
4
0.001
0.0
2
800
28
0.04
0.0
0.0
3
170
7,300
4
0.001
0.0
4
850
31
0.04
0.0
0.0
5
3,600
32
0.05
0.0
0.0
a To convert meters to feet, multiply by 3.2808.
b. Years after 21 16; source: adapted from Poe (1998a, all).
c. Source: Adapted from Davis (1998, all); the maximally exposed individual would be the onsite gardener.
The internal and extemal dose rates are presented separately because they would occur at different times
and are therefore not additive.
K.3.2 UNUSUAL EVENTS
This section includes a quantitative assessment of potential accident impacts and a qualitative discussion
of the impacts of sabotage.
K.3.2.1 Accident Scenarios
The analysis examined the impacts of accident scenarios that could occiu- during the above-ground
storage of spent nuclear fuel and high-level radioactive waste and concluded that the most severe accident
scenarios would be an aircraft crash into concrete storage modules or a severe seismic event. In Scenario
1, where storage would be in strong rigid concrete storage modules that had not degraded, the accident
would not be expected to release radioactive material.
K-33
Long-Term Radiological Impact Analysis for the No-Action Alternative
In Scenario 2, the concrete storage modules would deteriorate with time. DOE concluded that an aircraft
crash into degraded concrete storage modules would dominate the consequences. The analysis evaluated
the potential for criticality accidents and concluded that an event severe enough to produce meaningful
consequences would be extremely unlikely, and that the consequences would be bounded by the aircraft
crash consequences. Table K-14 lists the consequences of an aircraft crash on a degraded spent fuel
concrete storage module.
Table K-14. Consequences of aircraft crash onto degraded spent nuclear fuel concrete storage module.'
Impact High-population site** Low-population site*^
Collective population dose (person-rem) 26,000 6,000
Latent cancer fatalities 13 3
a. Source: Davis, Strenge, and Mishima (1998, page 11).
b. 330 persons per square mile.
c. 77 persons per square mile.
K.3.2.2 Sabotage
Storage of spent nuclear fiiel and high-level radioactive waste over 10,000 years would entail a continued
risk of intruder access at each of the 77 sites. Sabotage could result in a release of radionuclides to the
environment around the facility. In addition, intruders could attempt to remove fissile material, which
could result in releases of radioactive material to the environment. For Scenario 1, the analysis assumed
that safeguards and security measures currently in place would remain in effect during the 10,000-year
analysis period at the 77 sites. Therefore, the risk of sabotage would continue to be low. However, the
difficulty of maintaining absolute control over 77 sites for 10,000 years would suggest that the cumulative
risk of intruder attempts would increase.
For Scenario 2, the analysis assumed that safeguards and security measures would not be maintained at
the 77 sites after approximately the first 100 years. For the remaining 9,900 years of the analysis period,
the cumulative risk of intruder attempts would increase. Therefore, the risk of sabotage would increase
substantially under this scenario.
K.4 Uncertainties
Section K.3 contains estimates of the radiological impacts of the No-Action Alternative, which assumes
continued above-ground storage of spent nuclear fuel and high-level radioactive waste at sites across the
United States. Associated with the impact estimates are uncertainties typical of predictions of the
outcome of complex physical and biological phenomena and of the future state of society and societal
institutions over long periods. DOE recognized this fact from the onset of the analysis; however, the
predictions will be valuable in the decisionmaking process because they provide insight based on the best
information and scientific judgments available.
This analysis considered five aspects of uncertainty:
• Uncertainties about the nature of changes in society and its institutions and values, in the physical
environment, and of technology as technology progresses
• Uncertainties associated with future human activities and lifestyles
• Uncertainties associated with the mathematical representation of the physical processes and with the
data in the computer models
K-34
Long-Term Radiological Impact Analysis for the No-Action Alternative
• Uncertainties associated with the mathematical representation of the biological processes involving
the uptake and metabolism of radionuclides and the data in the computer models
• Uncertainties associated with accident scenario analysis
The following sections discuss these uncertainties in the context of possible effects on the impact
estimates reported in Chapter 7 and Section K.3.
K.4.1 SOCIETAL VALUES, NATURAL EVENTS, AND IMPROVEMENTS IN TECHNOLOGY
K.4.1.1 Societal Values
History is marked by periods of great social upheaval and anarchy followed by periods of relative
political stability and peace. Throughout history, governments have ended abruptly, resulting in social
instability, including some level of lawlessness and anarchy. The Scenario 1 assumption is that political
stability would exist to the extent necessary to ensure adequate institutional control to monitor and
maintain the spent nuclear fuel and high-level radioactive waste to protect the workers and the public for
10,000 years. The Scenario 2 assumption is that in the United States political stability would exist for
100 years into the future and that the spent nuclear fuel and high-level radioactive waste would be
properly monitored and maintained and the public would be protected for this length of time. If a
political upheaval, such as the one that recently occurred in the former Soviet Union, were to occur in the
United States, the government could have difficulty protecting and maintaining the storage facilities, and
the degradation processes could begin earlier than postulated in Scenario 2. If institutional control were
not maintained for at least 100 years, radioactive materials from the spent nuclear fuel and high-level
radioactive waste could enter the environment earlier, which would result in higher estimated impacts due
to the higher radiotoxicity of the materials. However, this scenario would probably increase overall
impacts by no more than a factor of 2.
K.4.1. 2 Changes In Natural Events
Because of the difficulty of predicting impacts of climate change (glaciation, precipitation, global
warming), DOE decided to evaluate facility degradation and environmental transport mechanisms based
on current climate conditions. For example, glaciation, which many scientists agree will occur again
within 10,000 years, probably would cover the northeastern United States with a sheet of ice. The ice
would crush all structures including spent nuclear fuel and high-level radioactive waste storage facilities
and could either disperse the radioactive materials in the accessible environment or trap the materials in
the ice sheet. In addition, large populations would migrate from the northeastern United States to warmer
climates, thus changing the population distribution and densities throughout the United States (the
coastline could move 100 miles out from its current position due to the reduced water in the oceans).
Other scientists predict that global warming could lead to extensive flooding of low-lying coastal areas
throughout the world. Such changes would have to be known with some degree of certainty to make
accurate estimates of potential impacts associated with the release of spent nuclear fuel and high-level
radioactive waste materials to the environment. To simplify the analysis, DOE has chosen not to attempt
to quantify the impacts resulting from the almost certain climate changes that will occur during the
analysis period.
K-35
Long-Term Radiological Impact Analysis for the No-Action Alternative
K.4.1 .3 Improvements in Technology
We are living in a time of unparalleled technical advancement. It is possible that cures for many common
cancers will be found in the coming decades. In this regard, the National Council on Radiation Protection
and Measurements (NCRP 1995, page 51) states that:
One of the most important factors likely to affect the significance of radiation dose in the centuries
and millennia to come is the effect of progress in medical technology. At some future time, it is
possible that a greater proportion of somatic [cancer] diseases caused by radiation will be treated
successfully. If, in fact, an increased proportion of the adverse health effects of radiation prove to
be either preventable or curable by advances in medical science, the estimates of long-term
detriments may need to be revised as the consequences (risks) of doses to future populations could
be very different.
Effective cures for cancer would affect the fundamental premise on which the No-Action Alternative
impact analysis is based. However, this technology change was not included in the impact analyses.
Other advancements in technology could include advancements in water purification that could reduce the
concentration of contaminants in drinking water supplies. Improved corrosion-resistant materials could
reduce package degradation rates, which could reduce the release of contaminants and the resultant
impacts. In addition, future technology could enable the detoxification of the spent nuclear fuel and high-
level radioactive waste materials, thereby removing the risks associated with human exposure.
K.4.2 CHANGES IN HUMAN BEHAVIOR
General guidance for the prediction of the evolution of society has been provided by the National
Research Council in Technical Bases for Yucca Mountain Standards (National Research Council 1995,
pages 28 and 70), in which the Committee on Technical Bases for Yucca Mountain Standards concluded
that there is no scientific basis for predicting future human behavior. The study recommends policy
decisions that specify the use of default (or reference) scenarios to incorporate future human behaviors
into compliance assessment calculations. This No-Action Alternative analysis followed this approach,
based on societal conditions as they exist today. In doing so, the analysis assumed that populations would
remain at their present locations and that population densities would remain at the current levels. This
assumption is appropriate when estimating impacts for comparison with other proposed actions; however,
it does not reflect reality. Populations are constantly moving and changing in size. If, for example,
populations were to move closer to and increase in size in areas near the storage facilities, the radiation
dose and resultant adverse impacts could increase substantially. However, DOE has no way to predict
such changes accurately and, therefore, did not attempt to quantify the resultant effects on overall
impacts.
Another lifestyle change that could affect the overall impacts would involve food consumption patterns.
For example, people might curtail their use of public water supplies derived from rivers if they learned
that the river water carried carcinogens. Widespread adoption of such practices could reduce the impacts
associated with the drinking water pathway.
K.4.3 MATHEMATICAL REPRESENTATIONS OF PHYSICAL PROCESSES AND OF THE
DATA INPUT
The DOE approach for the No-Action Altemative was to be as comparable as possible to the approach
used for the predictions of impacts from the proposed Yucca Mountain Repository to enable direct
comparisons of the impact estimates for the two cases. Therefore, the analysis either used the process
models developed for the Total System Performance Assessment directly or adapted them for the
K-36
I
Long-Term Radiological Impact Analysis for the No-Action Alternative
No-Action Alternative impact calculations. For processes that were different from those treated in the
Total System Performance Assessment, DOE developed analytical approaches.
hi a general sense, the Total System Performance Assessment calculations used a stochastic (random)
approach to develop radiological impact estimates. Existing process models were used to generate a set
of responses for a particular process, hi the Total System Performance Assessment process, the impact
calculations sample each set of process responses and calculate a particular impact result. A large number
of calculations were performed. From the set of variable results, an expected value can be identified, as
can a distribution of results that is an indication of the uncertainties in the calculated expected values.
For the No- Action Alternative analysis, the calculations were based on only a single set of best estimate
parameters. No statistical distribution of results was generated as a basis for the quantification of
uncertainties. This section describes the uncertainties associated with the input data and modeling used to
evaluate the rates of degradation of the materials considered in this document and to estimate the impacts
of the resulting releases. It describes the key assumptions, shows where the assumptions are consistent
with Total System Performance Assessment assumptions, and qualitatively assesses the magnitude of the
uncertainties caused by the assumptions.
Calculating the radiological impacts to human receptors required a mathematical representation of
physical processes (for example, water movement) and data input (for example, material porosity). There
are uncertainties in both the mathematical representations and in the values of data. The Total System
Performance Assessment accommodates these uncertainties by using a probabilistic approach to
incorporate the uncertainties, whereas the No-Action analysis uses a deterministic approach in
combination with an uncertainty analysis. When done correctly, both approaches yield the same
information, although, as in the case of the Total System Performance Assessment, the probabilistic
approach provides quantitative information.
K.4.3.1 Waste Package and Material Degradation
The major approaches and assumptions used for the No- Action Scenario 2 analysis are listed in
Table K-15. The table indicates where the continued storage calculations followed the basic methods
developed for the Total System Performance Assessment. It also indicates the processes for which
models other than those used in the Total System Performance Assessment were applied.
DOE analyzed surface storage of commercial spent nuclear fuel in horizontal stainless-steel canisters
inside concrete storage modules. There are other probable forms of storage, including horizontal and
vertical casks made of materials ranging from stainless steel to carbon steel. Degradation and releases
from vertical carbon-steel casks were evaluated qualitatively. Such storage units would be likely to fail
from corrosion earlier than concrete and stainless steel. The concrete and stainless-steel units were
calculated to fail and begin releasing their contents at about 1,000 years after the assumed loss of
institutional control. The less-resistant carbon-steel units could begin releasing their contents earlier and
their use would result in a longer period of release and increased impacts. This difference is likely to be
an increase of 10 to 30 percent in population dose commitment and resultant latent cancer fatalities.
K.4.3.2 Consequences of Radionuclide Release
The dose-to-risk conversion factors typically used to estimate adverse human health impacts resulting
from radiation exposures contain considerable uncertainty. The risk conversion factor of 0.0(X)5 latent
cancer fatality per person-rem of collective dose for the general public typically used in DOE National
Environmental Policy Act documents is based on recommendations of the International Commission on
Radiological Protection (ICRP 1991, page 22) and the National Council on Radiation Protection and
Measurements (NCRP 1993a, page 1 12). The factor is based on health effects observed in the high dose
K-37
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-15. Review of approaches, assumptions, and related uncertainties" (page 1 of 2).
Consistent with
repository analysis Sensitivity of impacts
Approach or assumption assumptions to approach or assumption''
Period of analysis - 10,000 years
Commercial spent nuclear fuel, DOE spent
nuclear fuel, and high-level radioactive
waste quantities equivalent to NWPA
specified 70,000 MTHM and Module 1
No credit for stainless-steel cladding on
commercial spent nuclear fuel
0.1 percent of zirconium alloy cladding is
initially failed
Concrete storage module weather protection
Concrete base pad degradation
Credit for stainless-steel canister on high-
level radioactive waste
DOE spent nuclear fuel evaluated by a
representative surrogate that is based mostly
on DOE N-Reactor spent nuclear fuel (other
spent nuclear fuel types not evaluated)
No credit given for zirconium alloy
cladding on N-Reactor spent nuclear fuel
Stainless steel deterioration
Yes
Yes
Yes
Yes
This is a primary protective
barrier for the No-Action
analysis and is not applicable
to TSPA
Not applicable
No; TSPA does not take credit
for stainless-steel container
Yes
Yes
Zirconium alloy cladding deterioration
Zirconium alloy cladding credit
Deterioration of spent nuclear fuel and
high-level radioactive waste core materials
Model paralleled TSPA
approach for Alloy-22
Yes, very slow corrosion rate.
Yes
Yes
None
None
If credit were taken for stainless-steel
cladding, LCFs" could decrease by as much
as a factor of 10.
If energetic events (that is, concrete
collapse) had been considered in the No-
Action analysis, impacts could have been
slightly smaller (additional protection from
winds) to a factor of 1(X) higher.
If weather protection from the concrete
storage module had not been assumed in the
No-Action analysis, LCFs could be higher
by less than a factor of 10.
Used NRC recommended values (probably
overestimated degradation and reduced
consequences in the No-Action analysis);
increase in LCFs by several factors but less
than a factor of 10
If the No-Action analysis had not taken
credit for the stainless-steel canister, LCFs
would change very little (slight increase)
because of the intrinsic stability of the
borosilicate glass.
If actual fuel types were evaluated, LCFs
could either increase or decrease by less
than a factor of 2.
If credit was given for the N-Reactor
zirconium alloy cladding, the LCFs would
decrease by less than a factor of 2.
Model based on best information; if
incorrect and corrosion proceeds more
rapidly and stainless steel offers no
protection, LCFs could increase by as much
as a factor of 100
If the No- Action analysis had assumed
larger or smaller deterioration rates, LCFs
could have increased by several orders of
magnitude or decreased by less than a factor
of 2.
If the No- Action analysis had not taken
credit for zirconium alloy cladding, LCFs
could have increased by as much as 2 orders
of magnitude.
None
K-38
Long-Term Radiological Impact Analysis for the No-Action Alternative
Table K-15. Review of approaches, assumptions, and related uncertainties^ (page 2 of 2).
Consistent with Sensitivity of impacts
Approach or assumption
repository analysis assumptions
to approach or assumption
Use of recent regional climate conditions
to determine deterioration (temperature,
precipitation, etc.)
Surface transport by precipitation
Regional binning of sites - not specific
site parameters
Atmospheric dose consequences judged to
be small when compared to liquid
pathways.
Drinking water doses
Used the Multimedia Environmental
Pollutant Assessment System*^ (Buck et al.
1995, all (Leigh et al. 1993, all) modeling
approach for calculating population
uptake/ingestion
ICRP' approach to calculate dose
commitment from ingested radionuclides
Human health impacts calculated as LCFs
with NCR?' conversion factors
No; No-Action analysis used
constant "effective" regional
weather parameters weighted
for material inventories and
potentially affected
downstream populations; TSPA
used actual weather patterns
measured at Yucca Mountain.
The TSPA also assumed long-
term climate changes would
occur in the form of increased
precipitation.
Not applicable; TSPA only
considered groundwater
transport because there is no
surface-water transport
pathway possible for the
repository.
Not applicable; TSPA
considered only a single site;
the No-Action analysis
evaluated potential impacts
from 77 sites on a regional
basis.
Yes
Yes; primary pathway
evaluated
No; TSPA uses GENII-S."
GENIl-S uses local survey
data; the Multimedia
Environmental Pollutant
Assessment System uses
EPA/NRC exposure/uptake
default and actual population
data
Yes
NA; TSPA does not estimate
LCFs.
If actual site climate data and projected
future potential climate changes had been
considered in the No-Action analysis, LCFs
could have increased or decreased by as
much as a factor of 10. Climate change
assumptions such as a glacier covering most
of the northeastern seaboard of the United
States would have made estimating impacts
from continued storage virtually impossible.
If the No- Action analysis had not
considered the groundwater transport
pathway, LCFs could have been as much as
a factor of 10 higher.
None, the No- Action analysis binned sites
into categories and developed "effective"
regional climate conditions such that
calculated impacts would be comparable to
those which could be calculated by a site-
specific analysis.
Small impact on LCFs
Use of drinking-water-only pathway
underestimates total collective LCFs by less
than afactor of 3.
No impact. The two programs yield
comparable results as used in these analyses.
No impact.
Use of other than the linear no-threshold
model could result in a change in estimated
LCFs from 0.25 to 2 times the nominal
value.*
a. Abbreviations: NWPA = Nuclear Waste Policy Act; MTHM = metric tons of heavy metal; LCF = latent cancer fatality; TSPA = Total
System Performance Assessment; NRC = Nuclear Regulatory Commission; ICRP = International Commission on Radiological Protection;
EPA = Environmental Protection Agency.
b. Sensitivity of impacts to approach/assumption is based on professional judgement and, if applicable, the effects of the
approaches/assumptions on calculations.
c. Buck etal. (1995, all).
d. Leigh et al. (1993, al).
e. ICRP (1979, all).
f. NCRP (1993a, page 112).
g. NCRP (1997, page 75).
K-39
Long-Term Radiological Impact Analysis for the No-Action Alternative
and high dose rate region (20 to 50 rem per year). Health effects were extrapolated to the low-dose region
(less than 10 rem per year) using the linear no-threshold model. This model is generally recommended by
the International Commission on Radiological Protection and the National Council of Radiation
Protection and Measurements, and most radiation protection professionals believe this model produces a
conservative estimate (that is, an overestimate) of health effects in the low-dose region, which is the
exposure region associated with continued storage of spent nuclear fuel and high-level radioactive waste.
This report summarizes estimates of the impacts associated with very small chronic population doses to
enable comparison of alternatives in this EIS. These impact estimates should be viewed as conservatively
high; in fact, the uncertainties are such that the actual level of impact could be zero.
According to the National Council on Radiation Protection and Measurements, the results of an analysis
of the uncertainties in the risk coefficients "show a range (90 percent confidence intervals) of uncertainty
values for the lifetime risk for both a population of all ages and an adult worker population from about a
factor of 2.5 to 3 below and above the 50th percentile value" (NCRP 1997, page 74).
The National Council on Radiation Protection and Measurements states, "This work indicates that given
the sources of uncertainties considered here, together with an allowance for unspecified uncertainties, the
values of the lifetime risk can range from about one-fourth or so to about twice the nominal values"
(NCRP 1997, page 75).
Because of the large uncertainties that exist in the dose/effect relationship, the Health Physics Society has
recommended ". . .against quantitative estimation of health risks due to radiation exposure below a
lifetime dose of 10 rem . . ." (HPS 1996, page 1). In essence, the Society has recommended against the
quantification of risks due to individual radiation exposures comparable to those estimated in the No-
Action analysis. These uncertainties are due, in part, to the fact that epidemiological studies have been
unable to demonstrate that adverse health effects have occurred in individuals exposed to small doses
(less than 10 rem per year) over a period of many years (chronic exposures) and to the fact that the extent
to which cellular repair mechanisms reduce the likelihood of cancers is unknown.
Other areas of uncertainty in estimation of dose and risk include the following:
• Uncertainties Related to Plant and Human Uptake of Radionuclides. There are large
uncertainties related to the uptake (absorption) of radionuclides by agricultural plants, particularly in
the case where "regionalized," versus "site-specific" data are used. Also of importance are variations
in the absorption of specific radionuclides through the human gastrointestinal tract. Factors that
influence the absorption of radionuclides include their chemical or physical form, their
concentrations, and the presence of stable elements having similar chemical properties. In the case of
agricultural crops, many of these factors are site-specific.
• Uncertainties in Dose and Risk Conversion Factors. The magnitudes and sources of the
uncertainties in the various input parameters for the analytical models need to be recognized. In
addition to the factors cited above, these include those required for converting absorbed doses into
equivalent doses, for calculating committed doses, and for converting organ doses into effective
(whole body) doses. Although these various factors are commonly assigned point values for purposes
of dose and risk estimates, each of these factors has associated uncertainties.
• Conservatisms in Various Models and Parameters. In addition to recognizing uncertainties, one
must take into account the magnitudes and sources of the conservatisms in the parameters and models
being used. These include the fact that the values of the tissue weighting factors and the methods for
calculating committed and collective doses are based on the assumption of a linear no-threshold
relationship between dose and effect. As the International Commission on Radiological Protection
K-40
Long-Tenn Radiological Impact Analysis for the No-Action Alternative
and the National Council on Radiation Protection and Measurements have stated, the use of the linear
no-threshold hypothesis provides an upper bound on the associated risk (ICRP 1966, page 56). Also
to be considered is that the concept of committed dose could overestimate the actual dose by a factor
of 2 or more (NCRP 1993b, page 25).
K.4.3.3 Accidents and Their Uncertainty
The accident methodology used in this analysis is described in Section K.2.5 for Scenarios 1 and 2. It
states that for Scenario 1 an aircraft crash into the storage array would provide the most severe accident
scenario and its consequences would not cause a release from the rugged concrete storage module. The
analysis placed considerable weight on the quality and strength of the concrete storage module and dry
storage canister. For an analysis extending 10,000 years, more severe natural events can be postulated
than those used as the design basis for the dry storage canister, and they could cause failure of the
canister. This could exceed the consequences estimated for Scenario 1, but it would be unlikely to exceed
the consequences for the aircraft accident scenario evaluated for Scenario 2.
Section K.2.5. 1 concludes that the aircraft crash on the degraded concrete storage modules would be the
largest credible event that could occur. The best estimate impacts from this event ranged from 3 latent
cancer fatalities for a low-population site to 13 for a high-population site. The uncertainties in these
estimates are very large. As discussed above, the aircraft crash could cause a minimum of no latent
cancer fatalities given the uncertainty in the model that converts doses to cancers. The maximum impact
could be 50 times greater than the estimated values if an aircraft crash involving the largest commercial
jet occurred at the time of initial concrete storage module degradation at a northern site under adverse
weather conditions (conditions that would maximize the offsite doses) involving spent fuel with the
maximum expected inventory of radionuclides.
K.4.4 UNCERTAINTY SUIVIMARY
The sections above discuss qualitatively and semiquantitatively the uncertainties associated with impact
estimates resulting from the long-term storage of spent nuclear fuel and high-level radioactive waste at
multiple sites across the United States. As stated above, DOE has not attempted to quantify the
variability of estimated impacts related to possible changes in climate, societal values, technology, or
future lifestyles. Although uncertainties with these changes could undoubtedly affect the total
consequences reported in Section K.3 by several orders of magnitude, DOE did not attempt to quantify
these uncertainties to simplify the analysis.
DOE attempted to quantify a range of uncertainties associated with mathematical models and input data,
and estimated the potential effect these uncertainties could have on collective human health impacts. By
summing the uncertainties discussed in Sections K.4.1, K.4.2, and K.4.3 where appropriate, DOE
estimates that total collective impacts over 10,000 years could have been underestimated by as much as
3 or 4 orders of magnitude. However, because there are large uncertainties in the models used for
quantifying the relationship between low doses (that is, less than 10 rem) and the accompanying health
impacts, especially under conditions in which the majority of the populations would be exposed at a very
low dose rate, the actual collective impact could be zero.
On the other hand, impacts to individuals (human intruders) who could move to the storage sites and live
close to the degraded facilities could be severe. During the early period (2(X) to 4(X) years after the
assumed loss of institutional control), acute exposures to external radiation from the spent nuclear fuel
and high-level radioactive waste material could result in prompt fatalities. In addition, after a few
thousand years onsite shallow aquifers could be contaminated to such a degree that consumption of water
from these aquifers could result in severe adverse health effects, including premature death. Uncertainties
K-41
Long-Term Radiological Impact Analysis for the No-Action Alternative
related to these localized impacts are related primarily to the inability to predict accurately how many
individuals could be affected at each of the 77 sites over the 10,000-year analysis period. In addition, the
uncertainties associated with localized impacts would exist for potential consequences resulting from
unusual events, both manmade and natural.
Therefore, as listed in Table K-15, uncertainties resulting from future changes in natural phenomena and
human behavior that cannot be predicted, process model uncertainties, and dose-effect relationships,
taken together, could produce the results presented in Section K.3, overestimating or underestimating the
impacts by as much as several orders of magnitude. Uncertainties of this magnitude are typical of
predictions of the outcome of complex physical and biological phenomena over long periods. However,
these predictions (with their uncertainties) are valuable to the decisionmaking process because they
provide insight based on the best information available.
REFERENCES
Battelle 1998
Bucketal. 1995
Chanin and Young 1998
CP&L 1989
Davis 1998
Davis, Strenge, and Mishima
1998
DOE 1992
DOE 1993
Battelle (Battelle Pacific Northwest Division), 1998, Analytical
Approach for Estimating Releases of Spent Nuclear Fuel and High-Level
Waste for the Yucca Mountain Environmental Impact Statement No-
Action Alternative, Richland, Washington. [MOL. 199905 13.0039]
Buck, J. W., G. Whelan, J. G. Droppo, Jr., D. L. Strenge, K. J. Castleton,
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Long-Term Radiological Impact Analysis for the No-Action Alternative
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Orthen 1999
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Foe 1998b
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Long-Term Radiological Impact Analysis for the No-Action Alternative
Rollins 1998c
Rollstin, Chanin, and Jow
1990
Shipers and Harlan 1989
Sinkowski 1998
Toblin 1998a
Toblin 1998b
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TRW 1995
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[MOL.19981008.0001]
TRW (TRW Environmental Safety Systems Inc.), 1998b, "Chapter 2:
Unsaturated Zone Hydrology Model," Total System Performance
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TRW (TRW Environmental Safety Systems Inc.), 1998c, "Chapter 3:
Thermal Hydrology," Total System Performance Assessment - Viability
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[MOL. 19981008.0003]
K-46
Long-Term Radiological Impact Analysis for the No-Action Alternative
TRW 1998d
TRW 1998e
TRW 1998f
TRW 1998g
TRW 1998h
TRW 19981
TRW 1998J
TRW 1998k
TRW (TRW Environmental Safety Systems Inc.), 1998d, "Chapter 4:
Near-Field Geochemical Environmental," Total System Performance
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Nevada. [MOL. 1998 1008.0004]
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Waste Package Degradation Modeling Abstraction," Total System
Performance Assessment - Viability Assessment (TSPA-VA) Analyses
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01, Las Vegas, Nevada. [MOL. 1998 1008.0005]
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Waste Form Degradation, Radionuclide Mobilization and Transport
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Nevada. [MOL. 1998 1008.0006]
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Unsaturated Zbne Radionuclide Transport," Total System Performance
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Saturated Zone Flow Transport," Total System Performance Assessment
- Viability Assessment (TSPA-VA) Analyses Technical Basis Document,
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Biosphere," Total System Performance Assessment - Viability
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Disruptive Events," Total System Performance Assessment - Viability
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Viability Assessment (TSPA-VA) Analyses Technical Basis Document,
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[MOL. 19981008.0011]
K-47
Appendix L
Floodplain/Wetlands Assessment
for the Proposed Yucca Mountain
Geologic Repository
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
<,■• ■
TABLE OF CONTENTS
Section Page
L.l Introduction L-1
L.2 Project Description Lh6
L.2.1 Proposed Actions at Yucca Mountain L-7
L.2.1.1 Rail Access L-7 * v.
L.2.1.2 Heavy-Haul Truck Access L-7 '"V
L.2.1.3 Construction L-8
L.2. 2 Possible Actions Elsewhere in Nevada L-8
L.3 Existing Environment L-9
L.3.1 Existing Environment at Yucca Mountain L-9
L.3.L1 Flooding L-9
L.3.1.2 Wetlands L-9 - ^. •
L.3.L3 Biology L-9 - '
L.3.L4 Archaeology L-10
L.3.2 Existing Environment Elsewhere in Nevada L-10 : ". '
L.3.2.1 Caliente Rail Corridor L-10
L.3.2.2 Carlin Rail Corridor L-12
L.3.2.3 Caliente-Chalk Mountain Rail Corridor L-13
L.3.2.4 Jean Rail Corridor L-13
L.3.2.5 Valley-Modified Rail Corridor L-14
L.3.2.6 Caliente Intermodal Transfer Station L-14
L.3.2.7 Apex/Dry Lake Intermodal Transfer Station L-15
L.3.2.8 Sloan/Jean Intermodal Transfer Station L-15
L.4 FloodplainAVetlands Effects L-15
L.4.1 Floodplain/Wetlands Effects Near Yucca Mountain L-16
L.4.2 Floodplain/Wetlands Effects Elsewhere in Nevada L-18
L.4.2.1 Effects along Rail Corridors L-1,8 ^ ;.. '
L.4.2.2 Effects at Intermodal Transfer Stations L-18 '^^■
L.5 Mitigation Measures L-18 ■\yi'l._
L.6 Alternatives L-19 , 'i^-
L.6.1 Alternatives Near Yucca Mountain ••L-19 C ,-^V
L.6.2 Alternative Rail Corridors and Alternative Sites for an Intermodal Transfer ', * : . -^ ,
Station h-^ J^%
L.6.3 No-Action Alternative L-20 f-^''
L.7 Conclusions L-20 ' • V-;^', '
References L-20 ^i^iH"
LIST OF TABLES ^ ^K ^
Table Page ^ i^('M
L-1 Surface-water-related resources along candidate rail corridors L-11 " yu'-
L-2 Length of each rail corridor implementing alternative L-12 ■ C
LIST OF FIGURES ^ ^ T
Figure Page .'.n' .
L-1 Yucca Mountain site topography, floodplains, and potential rail corridors L-3 A," /"^
L-2 Potential Nevada rail corridors to Yucca Mountain L-4 .. vt '-^■_
L-3 Potential routes in Nevada for heavy-haul trucks L-5 .■;?;:•
L-iii
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
APPENDIX L. FLOODPLAIN/WETLANDS ASSESSMENT FOR THE
PROPOSED YUCCA MOUNTAIN GEOLOGIC REPOSITORY
L.1 Introduction
Pursuant to Executive Order 1 1988, Floodplain Management, each Federal agency is required, when
conducting activities in a floodplain, to take actions to reduce the risk of flood damage; minimize the
impact of floods on human safety, health, and welfare; and restore and preserve the natural and beneficial
values served by floodplains. Pursuant to Executive Order 1 1990, Protection of Wetlands, each Federal
agency is to avoid, to the extent practicable, the destruction or modification of wetlands, and to avoid
direct or indirect support of new construction in wetlands if a practicable alternative exists. Regulations
issued by the U.S. Department of Energy (DOE) that implement these Executive Orders are contained in
Title 10 of the Code of Federal Regulations (CFR) Part 1022, Compliance with Floodplain/Wetlands
Environmental Review Requirements.
In 1982, Congress enacted the Nuclear Waste Policy Act in recognition of the national problem created by
the accumulation of spent nuclear fuel and high-level radioactive waste at many commercial and DOE sites
throughout the country. The Act recognized the Federal government's responsibility to permanently
dispose of the Nation's spent nuclear fuel and high-level radioactive waste. By 1986, DOE narrowed the
number of potentially acceptable geologic repository sites to three. Then in 1987, Congress amended the
Act by redirecting DOE to determine the suitability of only Yucca Mountain in southern Nevada.
If, after a possible reconunendation by the Secretary of Energy, the President considers the site qualified
for an application to the U.S. Nuclear Regulatory Commission for a construction authorization, the
President will submit a recommendation of the site to Congress. If the site designation becomes effective,
the Secretary of Energy will submit to the Nuclear Regulatory Commission a License Application for a
construction authorization. DOE could then select a rail corridor or a site for an intermodal transfer
station, along with its associated route for heavy-haul trucks, among those considered for Nevada in the
EIS. Following such a decision, additional field surveys, environmental and engineering analyses, and
National Environmental Policy Act reviews would likely be needed regarding a specific rail alignment for
the selected corridor or the site for the intermodal transfer station and its associated route. When more
specific information becomes available about activities proposed to take place within floodplains and
wetlands, DOE will conduct further environmental review in accordance with 10 CFR 1022.
In 1989, DOE published a Notice of FloodplainAVetlands Involvement (54 FR 6318, February 9, 1989) for
site characterization studies at Yucca Mountain. These studies are designed to determine the suitability of
Yucca Mountain to isolate nuclear waste. A floodplain assessment was prepared (DOE 1991, all) and a
Statement of Findings was issued by DOE (56 FR 49765, October 1, 1991). In 1992, DOE prepared a
second floodplain assessment on locating part of the entry point to the subsurface Exploratory Studies
Facility in the 100-year floodplain of a wash at Yucca Mountain (DOE 1992, all). The Statement of
Findings for this assessment was published in the Federal Register (57 FR 48363, October 23, 1992).
Both Statements of Findings concluded that the benefits of locating activities and structures in the
floodplains outweigh the potential adverse impacts to the floodplains and that alternatives to these actions
were not reasonable.
The Nuclear Waste Policy Act, as amended, requires that a recommendation by the Secretary to the
President to construct a repository must be accompanied by a Final EIS. As part of the EIS process, and
following the requirements of 10 CFR Part 1022, DOE issued a Notice of Floodplain and Wetlands
Involvement in the Federal Register (64 FR 31554, June 1 1, 1999). The Notice requested comments from
L-1
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
the public regarding potential impacts on floodplains and wetlands associated with construction of a
potential rail line or a potential intermodal transfer station with its associated route for heavy-haul trucks to
and in the vicinity of Yucca Mountain, depending on the rail or intermodal alternative selected (Figure
L-1). As of July 2,1999, DOE had received no conunents from the public. This floodplain/wedands
assessment has been prepared in conjunction with the Notice of Floodplain and Wetlands Involvement, and
in accordance with 10 CFR Part 1022.
This assessment examines the effects of proposed repository construction and operation and potential
construction of a rail line or intermodal transfer station on:
1 . Floodplains near the Yucca Mountain site (Fortymile Wash, Busted Butte Wash, Drillhole Wash, and
Midway Valley Wash; there are no delineated wetlands near the Yucca Mountain site), and
2. Floodplains and areas that may have wetlands (for example, springs and riparian areas) along potential
rail corridors in Nevada and at intermodal transfer station locations associated with routes for heavy-
haul trucks. If DOE selects rail as the mode of spent nuclear fuel and high-level radioactive waste
transport in Nevada to the Yucca Mountain site, one of five rail corridors would be selected (Figure
L-2). If DOE selects heavy-haul as the mode of transport for spent nuclear fuel and high-level
radioactive waste to the Yucca Mountain site, one of five corridors and one of three intermodal transfer
station locations would be selected (Figure L-3). A more detailed floodplain/wetlands assessment of
the selected rail corridor or route for heavy-haul trucks would then be prepared. This assessment
compares what is known about the floodplains, springs, and riparian areas along the five possible rail
corridors and at the three intermodal transfer station locations. This assessment does not evaluate
potential floodplain or wetlands effects along routes because these existing roads should already be
designed to meet 100-year floodplain design specifications. If upgrades to existing roads are deemed
necessary, a more detailed floodplain/wetlands assessment would be prepared at that time.
Title 10 CFR Part 1022.4 defines a flood or flooding as ". . .a temporary condition of partial or complete
inundation of normally dry land areas from.... the unusual and rapid accumulation of runoff of surf ace
waters... " Title 10 CFR Part 1022.4 identifies floodplains that must be considered in a floodplain
assessment as the base floodplain and the critical-action floodplain. The base floodplain is the area
inundated by a flood having a 1 .0 percent chance of occurrence in any given year (referred to as the
100-year floodplain). The critical-action floodplain is the area inundated by a flood having a 0.2 percent
chance of occurrence in any given year (referred to as the 5(X)-year floodplain). Critical action is defined
as any activity for which even a slight chance of flooding would be too great. Such actions could include
the storage of highly volatile, toxic, or water-reactive materials. The critical-action floodplain was
considered because petroleum, oil, lubricants, and other hazardous materials could be used during the
construction of a rail line or road upgrades and because spent nuclear fuel and high-level radioactive waste
would be transported across the washes.
Title 10 CFR Part 1022. 1 1 requires DOE to use Flood Insurance Rate Maps or Flood Hazard Boundary
Maps to determine if a proposed action would be located in the base or critical-action floodplain. On
Federal or state lands where Flood Insurance Rate Maps or Flood Hazard Boundary Maps are not
available, DOE is required to seek flood information from the appropriate land-management agency or
from agencies with expertise in floodplain analysis. The U.S. Geological Survey was therefore asked by
DOE to complete a flood study of Fortymile Wash and its principal tributaries (which include Busted
Butte, Drillhole, and Midway Valley washes) and outline areas of inundation from l(X)-year and 500-year
floods (Squires and Young 1984, Plate 1).
L-2
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Figure L-1. Yucca Mountain site topography, floodplains, and potential rail corridors.
L-3
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Oregon I Idaho
Reno
Legend
Existing rail line
Highway
State line
County line
^^^s Potential rail corridor
Mesquite
Variation of potential
rail corridor
Carlin corridor 520 kilonneters
Caliente corridor 513 kilometers
Caliente-Chalk corridor 345 kilometers
Valley Modified corridor 159 kilometers
Jean corridor 1 81 kilometers
50
50 Kilometers
Sourca: Moditied from
DOE (1998a. all).
Figure L-2. Potential Nevada rail corridors to Yucca Mountain.
L-4
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
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L-5
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Title 10 CFR Part 1022 also requires DOE to determine whether wetlands would be affected by the
proposed action and, if necessary, to conduct a wetlands assessment. As required by 10 CFR Part
1022. 11 (c), DOE examined the following information with regard to possible wetlands in the vicinity of
the Yucca Mountain site:
• U.S. Fish ana Wildlife Service National Wetlands Inventory. Maps from the National Wetlands
Inventory do not identify any naturally occurring wetlands in the vicinity of the Yucca Mountain site
(FWS 1995, all).
• U.S. Department of Agriculture, Soil Conservation Service Local Identification Maps. The
Soils Conservation Service (now called Natural Resource Conservation Service) has not conducted a
soil survey of the Yucca Mountain site. However, DOE and other agencies have conducted
comprehensive surveys and studies of soils at the Yucca Mountain site and in the surrounding area.
These surveys are summarized in TRW (1999a, pages 2 to 6). The surveys indicate that there are no
naturally-occurring hydric soils at Yucca Mountain.
• U.S. Geological Survey Topographic Maps. Topographic maps of the vicinity (for example,
USGS 1983, all) do not show springs, permanent streams, or other indications of wetlands.
• State Wetlands Inventories. There are no State of Nevada wetlands inventories in the vicinity of
Yucca Mountain.
• Regional or Local Government-Sponsored Wetlands or Land-Use Inventories. DOE has
conducted a wetlands inventory of the Nevada Test Site (Hansen et al. 1997, page 1-161). The closest
naturally occurring wetlands to Yucca Mountain is on the upper west slope of Fortymile Canyon,
6 kilometers (3.7 miles) north of the North Portal, outside of the proposed repository construction area.
In addition, riparian vegetation occurs adjacent to four man-made well ponds east of Yucca Mountain
(TRW 1999b, page 2-14), but these are outside of areas where construction or other proposed actions
would occur.
Based on this information, DOE concluded that a wetlands assessment is not required to comply with
10 CFR Part 1022.
L.2 Project Description
If Yucca Mountain is selected as a site to construct a repository, DOE would ship spent nuclear fuel and
high-level radioactive waste to the site for a period of about 24 years. Under the current schedule spent
nuclear fuel and high-level radioactive waste emplacement would begin in 2010. One of five possible rail
corridors leading to the site could be selected in Nevada (Figure L-2). In the vicinity of the Yucca
Mountain site the five rail corridors converge to two possible routes. Alternatively, if heavy-haul transport
were selected, one intermodal transfer station and one associated route would be identified from the three
potential intermodal transfer station locations and five potential routes for heavy-haul trucks (Figure L-3).
In the vicinity of the Yucca Mountain site, the potential routes converge to two possible routes that may
require upgrades. At greater distances, routes would utilize public roads and existing Nevada Test Site
roads to the extent possible.
Some transportation-related actions associated with the DOE proposal would occur in floodplains on the
proposed repository site on land the Federal government would manage. Route construction and operation
could affect the 100-year and 500-year floodplains of Fortymile Wash, Busted Butte Wash, Drillhole
Wash, and Midway Valley Wash in the vicinity of the Yucca Mountain site. This assessment examines the
L-6
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
potential floodplain impacts to all four washes although all four might not be affected. The effects on
floodplains and areas that may contain wetlands elsewhere in Nevada along the five rail corridors and at
the three intermodal station locations associated with heavy-haul transport are examined using available
information. When DOE makes a decision whether to use rail or heavy-haul transport, more information
would be obtained to support further environmental review.
This section is divided into two parts. Section L.2.1 discusses the proposed action in the vicinity of the
Yucca Mountain site including rail access; heavy-haul truck access; and potential construction of an
associated rail line, bridge, and roads. Section L.2.2 discusses possible actions elsewhere in Nevada
including rail access and intermodal transfer station locations.
L.2.1 PROPOSED ACTIONS AT YUCCA MOUNTAIN
The preliminary layout of surface facilities at the repository is shown on Figure L-1 . Except for a possible
rail line and roads, no facilities are generally anticipated to be located within either the 100-year or
500-year floodplains of Fortymile Wash, Busted Butte Wash, Drillhole Wash, or Midway Valley Wash.
The paragraphs below describe the rail line and roads that could affect the floodplains of these washes in
the vicinity of the Yucca Mountain site.
L.2.1 .1 Rail Access
At this time, there is no rail access to the Yucca Mountain site. DOE has identified five potential rail
corridors in Nevada for transporting spent nuclear fuel and high-level radioactive waste to Yucca
Mountain.
If DOE selected a rail corridor leading to the Yucca Mountain site from the west and south (either the
Carlin or Caliente corridors), the rail line could cross Busted Butte Wash, Drillhole Wash just west of its
confluence with Fortymile Wash, and Midway Valley Wash (Figure L-1). Cut, fill, drainage culverts or
bridges could be used to cross Busted Butte, Drillhole, and Midway Valley washes. The widths of Busted
Butte Wash and Drillhole Wash (including their floodplains) are about 150 meters (500 feet) each where
they would be crossed by the rail line. The width of Midway Valley Wash (including its floodplain) is
about 300 meters (1,000 feet) where it could be crossed by the rail line.
If DOE selected a rail corridor leading to the Yucca Mountain site from the east (Caliente-Chalk Mountain,
Jean, or Valley-Modified corridors) the rail line could cross approximately 400 meters (1,300 feet) of
Fortymile Wash and its associated floodplains. In this case, the rail line could cross the wash on either a
bridge (with supports located in the wash) or on a raised rail line that could be constructed in the wash
(with appropriately-sized drainage culverts). After crossing Fortymile Wash, the rail line could continue
along the east side of Yucca Mountain and cross about 300 meters (1,000 feet) of Midway Valley Wash
before arriving at the repository.
L.2.1 .2 Heavy-Haul Truck Access
DOE has identified five potential routes for heavy-haul trucks in Nevada for transporting spent nuclear fuel
and high-level radioactive waste to the Yucca Mountain site.
If DOE selected a route leading to the Yucca Mountain site from the west and south, the route could cross
Busted Butte Wash, Drillhole Wash, and Midway Valley Wash (Figure L-1). Cut, fill, drainage culverts or
bridges could be used to cross Busted Butte, Drillhole, and Midway Valley washes.
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Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
If DOE selected a route leading to the Yucca Mountain site from the east, the route could cross Fortymile
Wash. The route could either cross through the wash or a bridge could be constructed over it. After
crossing Fortymile Wash, the route could continue along the east side of Yucca Mountain and could cross
Midway Valley Wash before arriving at the repository.
During potential repository operation, some spent nuclear fuel and high-level radioactive waste would be
transported to the Yucca Mountain site by legal-weight trucks. These trucks could access Yucca Mountain
from the east by crossing Fortymile Wash along the existing road or access Yucca Mountain along the
route used by heavy-haul trucks. The legal-weight trucks could then proceed along the east side of Yucca
Mountain and cross Midway Valley Wash along the route.
L.2.1.3 Construction
Construction of a potential rail line near Yucca Mountain as well as upgrading the existing roads for
heavy-haul and legal-weight trucks in the vicinity would take about one year to complete. Standard
construction practices would be used, including the use of explosives and heavy earth-moving equipment.
Standard measures would also be used to minimize erosion. Petroleum fuels, oils, lubricants and other
hazardous materials would be used during construction, although these materials would be stored outside
the 500-year floodplain.
Construction aggregate could be obtained from local borrow pits, but rail-bed ballast would need to be
obtained from outside sources. Concrete would be obtained from a nearby concrete batch plant or from a
new batch plant that may be built closer to the repository site. Neither the borrow pits nor the concrete
batch plant would be located in a floodplain or wetlands.
If a bridge were constructed across Fortymile Wash, it would be about 30 meters (100 feet) wide. Supports
for the bridge would be constructed in the floodplain of the wash. If a rail line were constructed across the
bottom of Fortymile Wash, extensive earthwork (cut and fill) would be required to maintain the less than
two percent grade required for the rail alignment.
L.2.2 POSSIBLE ACTIONS ELSEWHERE IN NEVADA
At this time there is no rail access to Yucca Mountain. This means that material traveling by rail would
have to continue to the repository on a new branch rail line or transfer to heavy-haul trucks at an
intermodal transfer station in Nevada and then travel on existing highways. DOE is considering
construction of either a new branch rail line or an intermodal transfer station and associated highway
improvements. The DOE has identified five possible rail corridors, each of which has alignment variations
(Figure L-2), and three possible locations for an intermodal transfer station associated with heavy-haul
trucks (Figure L-3).
For analytical purposes, it is assumed that construction of a rail line in Nevada would take approximately
two and one half years. If a decision were made to proceed with development of a repository, it is likely
that the DOE would decide at that time whether to build a rail line or to develop an intermodal transfer
station site for heavy-haul waste transport. Should the DOE decide to construct a rail line, standard
practices for construction of rail lines would be used, including minimizing steep grades, utilizing cut and
full earthwork techniques, and crossing flood prone areas using culverts or bridges. Should the DOE
decide to use a route for heavy-haul trucks, portions of the existing roads used for heavy-haul transport
may require upgrades to acconmiodate the heavy loads.
L-8
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
L.3 Existing Environment
L.3.1 EXISTING ENVIRONMENT AT YUCCA MOUNTAIN
Fortymile Wash is about 150 kilometers (93 miles) long and drains an area of about 810 square kilometers
(310 square miles) to the east and north of Yucca Mountain (Figure L-1). The wash continues southward
and connects to the Amargosa River. The Amargosa River drains an area of about 8,000 square kilometers
(3,100 square miles) by the time it reaches Tecopa, California. The mostly-dry river bed extends another
90 kilometers (56 miles) before ending in Death Valley.
Busted Butte and Drillhole washes drain the east side of Yucca Mountain and flow into Fortymile Wash
(Figure L-1; Midway Valley Wash is a tributary to Drillhole Wash). Busted Butte Wash drains an area of
17 square kilometers (6.6 square miles) and Drillhole Wash drains an area of 40 square kilometers
(15 square miles).
The existing environment at and near Yucca Mountain, including Fortymile Wash, Busted Butte Wash,
Drillhole Wash, and Midway Valley Wash is described in Chapter 3 of the EIS. The information below
summarizes several of the more important aspects of the environment that pertain to this floodplain
assessment.
L.3.1 .1 Flooding
Water flow in the four washes is rare. The arid climate and meager precipitation [about 10 to 25
centimeters (4 to 10 inches) per year at Yucca Mountain] result in quick percolation of surface water into
the ground and rapid evaporation. Flash floods, however, can occur after unusually strong summer
thunderstorms or during sustained winter precipitation. During these times, runoff from ridges, pediments,
and alluvial fans flows into the normally dry washes that are tributary to Fortymile Wash. Estimated peak
discharges in Fortymile Wash are 340 cubic meters per second (720,000 cubic feet per second) for the
100-year flood and 1,600 cubic meters per second (3,390,(X)0 cubic feet per second) for the 500-year
flood. Estimated peak discharges in Busted Butte Wash are 40 cubic meters per second (85,(X)0 cubic feet
per second) for the 100-year flood and 180 cubic meters per second (380,000 cubic feet per second) for the
500-year flood. Estimated peak discharges in Drillhole Wash are 65 cubic meters per second (140,000
cubic feet per second) for the l(X)-year flood and 280 cubic meters per second (590,000 cubic feet per
second) for the 500-year flood.
The nearest man-made structure within Fortymile Wash is U.S. Highway 95 more than 19 kilometers
(12 miles) south of the confluence of Drillhole and Fortymile washes. Lathrop Wells, the nearest
population center to Yucca Mountain, is also about 19 kilometers to the south along U.S. 95 and
3.2 kilometers (2 miles) east of Fortymile Wash.
L.3.1 .2 Wetlands
There are no springs, perennial streams, hydric soils, or naturally occurring wetlands at Yucca Mountain.
There are two man-made well ponds within Fortymile Wash, and two east of that wash, that have riparian
vegetation (TRW 1999a, pages 5 to 6; TRW 1999b, page 2-14).
L. 3.1.3 Biology
Vegetation at and near Fortymile Wash is typical of the Mojave Desert. The mix or association of
vegetation in Fortymile Wash, which is dominated by the shrubs white bursage {Ambrosia dumosa).
L-9
■',■*•■
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
€ i. creosotebush (Larrea tridentata), white burrobush {Hymenoclea salsola), and heathgoldenrod (Ericameria
}' "^ paniculata), differs somewhat from other vegetation association at Yucca Mountain (TRW 1998a, pages 5
'•■\^' to 7). No plant species are known to be restricted to the floodplains. In addition, none of the more than
:^ 1 ^80 plant species known to occur at Yucca Mountain is endemic to the area.
., V ' None of the 36 mammal, 27 reptile, or 120 bird species that have been documented at Yucca Mountain are
>; restricted to or dependent on the floodplain. These species all are widespread throughout the region. No
■. , ; \ amphibians have been found at Yucca Mountain.
The only plant or animal species that has been found at Yucca Mountain that is classified as threatened,
endangered, or proposed under the Endangered Species Act is the desert tortoise {Gopherus agassizii)
which is classified as threatened. Yucca Mountain is at the northern edge of the range of the desert tortoise
(Rautenstrauch, Brown, and Goodwin 1994, page 11). Desert tortoises are known to occur within the
floodplain of Fortymile Wash, but their abundance there and elsewhere at Yucca Mountain is low
compared to other parts of its range farther south and east (TRW 1997, pages 6 to 1 1). Information on the
ecology of the desert tortoise population at Yucca Mountain is summarized in TRW (1999b, page 2-8).
Four species classified as sensitive by the Bureau of Land Management occur at Yucca Mountain: two
species of bats [the long-legged myotis (Myotis volans) and the fringed myotis {Myotis thysanodes)] (TRW
1998b, page 11), the western chuckwalla (Sauromalus obesus obesus) (TRW 1998c, pages 22 to 23), and
the western burrowing owl (Speotyto cunicularia hypugaea) (Steen et al. 1997, pages 19 to 29). These
species may occur within the floodplain of Fortymile Wash, but they are not dependent upon habitat there
(TRW 1998b, page 8; TRW 1998c, pages 22 to 23; Steen et al. 1997, pages 19 to 29).
L.3.1.4 Archaeology
Archaeological surveys have been conducted in Fortymile Wash east of Yucca Mountain. Fortymile Wash
was an important crossroad where several trails converged from such distant places as Owens Valley,
Death Valley, and the Avawtz Mountains.
L.3.2 EXISTING ENVIRONMENT ELSEWHERE IN NEVADA
The following sections describe the environment along each of the five possible rail corridors (Figure L-2)
and at the three intermodal transfer station locations (Figure L-3). Table L-1 lists surface-water-related
resources along each of the five rail corridors. The corridors are about 0.4 kilometer (0.25 mile) wide, and
the length of each corridor varies (Table L-2). Details of each of the corridors and surface-water-related
resources are found in TRW (1999b, Appendixes E, F, G, H, and I).
More detail on each of the rail corridors is provided in Chapter 2, Section 2.1.3.3.2, and Chapter 3,
Section 3.2.2. Chapter 6, Section 6.3.2, describes the potential impacts of rail implementing alternatives
and Chapter 6, Section 6.3.3 describes the potential impacts of the construction and use of intermodal
transfer stations under the heavy-haul truck implementing alternatives.
L.3.2.1 Caliente Rail Corridor
Flooding: The Caliente rail corridor crosses 352 washes en route to the Yucca Mountain site (TRW
1999c, pages 3 to 4). Approximately 12 washes along this route are large enough that bridges would be
required to cross them. Floodplains associated with these washes have not been defined at this time.
Wetlands: At least four springs or groups of springs and three streams or riparian areas that may have
associated wetlands are within 0.4 kilometer (0.25 mile) of the Caliente rail corridor. However, no field
L-10
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Table L-1. Surface-water-related resources along candidate rail corridors."
Distance from
corridor
Rail corridor
(kilometers)''
Caliente
Caliente to Meadow Valley
0.5
Within
Meadow Valley to Sand Spring
Valley
1.0
0.05 - 2.6
Within
0.8
Sand Spring Valley to Mud Lake
Mud Lake to Yucca Mountain
0.02
Within - 2.5
Within
0.3 - L3
Within - 0.3
Carlin
Beowawe to Austin
0.5
0.8
0.9
0.4
0.8
1.0
Within
Within
Austin to Mud Lake
Mud Lake to Yucca Mountain
Caliente-Chalk Mountain
Caliente to Meadow Valley
Meadow Valley to Sand Spring
Valley
Sand Spring Valley to Yucca
Mountain
0.1
Within
0.7
0.8
0.6
0.6
0.3
1.0
0.8
Jean
Valley Modified
Feature
Springs - two unnamed springs, in Meadow Valley north of Caliente
Riparian area/stream - corridor crosses and is adjacent to stream and
riparian area in Meadow Valley Wash
Spring - Bennett Spring, 3.2 kilometers southeast of Bennett Pass
Springs - group of five springs (Deadman, Coal, Black Rock,
Hamilton, and one unnamed) east of White River
Riparian/river - corridor parallels (and crosses) the White River for
about 25 kilometers. August 1997 survey found river to be
mostly underground with ephemeral washes above ground.
Spring - McCutchen Spring, north of Worthington Mountains
Spring - Black Spring, south of Warm Springs
Springs - numerous springs and seeps along Amargosa River in
Oasis Valley
Riparian Area - designated area east of Oasis Valley, flowing into
Amargosa Valley
Springs - group of 13 unnamed springs in Oasis Valley north of
Beatty
Riparian area/stream - Amargosa River, with persistent water and
extensive wet meadows near springs and seeps
Spring - Tub Spring, northeast of Red Mountain
Spring - Red Mountain Spring, east of Red Mountain
Spring - Summit Spring, west of corridor and south of Red
Mountain
Spring - Dry Canyon Spring, west of Hot Springs Point
Spring - unnamed spring on eastern slope of Toiyabe Range,
southwest of Hot Springs Point
Riparian area - intermittent riparian area associated with Rosebush
Creek, in western Grass Valley, north of Mount Callaghan
Riparian/creek - corridor crosses Skull Creek, portions of which
have been designated riparian areas
Riparian/creek - corridor crosses intermittent Ox Corral Creek;
portions designated as riparian habitat. August, 1997 survey
found creek dry with no riparian vegetation present
Spring - Rye Patch Spring, at north entrance of Rye Patch Canyon,
west of Bates Mountain
Riparian area - corridor crosses and parallels riparian area in Rye
Patch Canyon
Spring - BuUrush Spring, east of Rye Patch Canyon
Springs - group of 35 unnamed springs, about 25 kilometers north of
Round Mountain on east side of Big Smokey Valley
Riparian area - marsh area formed from group of 35 springs
Spring - Mustang Spring, south of Seyler Reservoir
Riparian/reservoir - Seyler Reservoir, west of Manhattan
See Caliente corridor
See Caliente corridor
See Caliente corric*
Spring - Reitman's Seep, in eastern Yucca Flat, east of BJ Wye
Spring - Cane Spring, on north side of Skull Mountain on Nevada
Test Site
None identified
None identified
a. Source: TRW (1999b, Appendixes E, F, G, H, and I).
b. To convert kilometers to miles, multiply by 0.62137.
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Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Table L-2. Length of each rail corridor implementing alternative.
Rail corridor Length
Caliente 513 kilometers (319 miles)
Carlin 520 kilometers (323 miles)
Caliente-Chalk Mountain 345 kilometers (214 miles)
Jean 1 8 1 kilometers ( 1 1 2 miles)
Valley Modified 159 kilometers (99 miles)
searches or formal delineations of wetlands have been conducted along this route. Black Spring is near the
corridor at the north end of the Kawich Range and an unnamed spring is near the corridor at the north end
of the North Pahroc Range. An unnamed spring is 0.3 kilometer (0.2 mile) east of the corridor between
Mud Lake and the Yucca Mountain site. A group of springs is in the corridor near the Amargosa River in
Oasis Valley. The corridor crosses the Meadow Valley Wash south of Panaca. The corridor also crosses
the White River between U.S. Highway 93 and Sand Spring Valley and parallels the river for
approximately 26 kilometers (16 miles). That portion of the White River normally is dry. The corridor
crosses the Amargosa River in the north end of the Oasis Valley, in an area designated as riparian area by
the Bureau of Land Management (TRW 1999b, page 3-23).
Biology: The desert tortoise is the only threatened or endangered species found along the Caliente rail
corridor. The southern 50 kilometers (30 miles) of this corridor is within desert tortoise habitat. This area
is not designated as critical habitat and the abundance of tortoises in the area is low (TRW 1999b, page
3-23). Three other species (Meadow Valley Wash speckled dace [Rhinichthys osculus ssp.]. Meadow
Valley Wash desert sucker [Catostomus clarki ssp.], and Nevada sanddune beardtongue) classified as
sensitive by the Bureau of Land Management or as protected by Nevada have been found along the
Caliente rail corridor. This rail corridor crosses approximately 14 areas designated as game habitat and
one area classified as waterfowl habitat (TRW 1999b, page 3-23). Two of these species, the speckled dace
and desert sucker, are restricted to the floodplain of the Meadow Valley Wash. The designated waterfowl
habitat also is generally restricted to the floodplain of Meadow Valley Wash and adjacent wetlands.
Archaeology. There are 97 archaeological sites that have been recorded along the Caliente route.
L.3.2.2 Carlin Rail Corridor
Flooding: The Carlin rail corridor crosses 273 washes en route to the Yucca Mountain site (TRW 1999c,
pages 3 to 4). Approximately 10 washes along this route are large enough that bridges would be required
to cross them. Floodplains associated with these washes have not been defined at this time.
Wetlands: There are at least three springs or groups of springs, six streams designated as riparian areas by
the Bureau of Land Management, and one reservoir that may have associated wetlands within 0.4
kilometer (0.25 mile) of the Carlin rail corridor. However, no field searches or formal delineations of
wetlands have been conducted along this route. Rye Patch Spring is on the edge of the corridor at the
south end of the Simpson Park Mountains, an unnamed spring is 0.3 kilometer (0.2 mile) east of the
corridor between Mud Lake and Yucca Mountain, and a group of springs is in the corridor near the
Amargosa River in Oasis Valley. Seyler Reservoir is 0. 16 kilometer (0. 1 mile) from the corridor in the
south end of Big Smoky Valley. There are five riparian areas (Skull, Steiner, and Ox Corral creeks, and
Water and Rye Patch canyons) along the section of the route between Beowawe and Austin at the south
end of Grass Valley. Two of these (Steiner and Ox Corral creeks, both at the south end of Grass Valley)
are ephemeral and have little or no riparian vegetation where the route crosses them. The corridor crosses
the Amargosa River in the northern Oasis Valley, in an area designated as a riparian area by the Bureau of
Land Management (TRW 1999b, pages 3-25 to 3-26).
L-12
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Biology: The desert tortoise is the only threatened or endangered species found along the Carlin rail
corridor. The southern 50 kilometers (30 miles) of this corridor is within desert tortoise habitat. This area
is not designated as critical habitat and the abundance of tortoises in the area is low (TRW 1999b, page
3-25). Three other species (ferruginous hawk [Buteo regalis], San Antonio pocket gopher [Thomomys
umbrinus curtatus], and Nevada sand dune beardtongue [Penstemom arenarius]) classified as sensitive by
the Bureau of Land Management or as protected by the State of Nevada have been found along the Carlin
rail corridor. Additionally, the rail corridor crosses approximately 1 1 areas designated as game habitat by
the Bureau of Land Management (TRW 1999b, page 3-25). None of these species or game habitats are
restricted to floodplains or areas that may have wetlands.
Archaeology: There are 11 0 archaeological sites that have been recorded along the Carlin route.
L.3.2.3 Caliente-Chalk Mountain Rail Corridor
Flooding: The Caliente-Chalk Mountain rail corridor crosses 281 washes en route to the Yucca Mountain
site (TRW 1999c, pages 3 to 4). Approximately five washes along this route are large enough that bridges
would be required to cross them. Floodplains associated with these washes have not been defined at this
time.
Wetlands: One spring and two streams that may have associated wetlands occur within 0.4 kilometer
(0.25 mile) of the Caliente-Chalk Mountain rail corridor. However, no field searches or formal
delineations of wetlands have been conducted along this route. An unnamed spring is near the corridor at
the north end of the North Pahroc Range. The corridor crosses Meadow Valley Wash south of Panaca.
The corridor crosses the White River between U.S. 93 and Sand Spring Valley and parallels the river for
approximately 26 kilometers (16 miles). That portion of the White River normally is dry.
Biology: The desert tortoise is the only threatened or endangered species found along the Caliente-Chalk
Mountain rail corridor. The southern 40 kilometers (25 miles) of this corridor is within desert tortoise
habitat. This area is not designated as critical habitat and the abundance of tortoises in the area is low
(TRW 1999b, page 3-27). Six species (Meadow Valley Wash speckled dace. Meadow Valley Wash desert
sucker, Ripley's springparsley [Cymopterus ripleyi var. saniculoides], largeflower suncup [Camissonia
megalantha], Beatley's scorpionweed [Phacelia beatleyae], and long-legged myotis [Myotis volans])
classified as sensitive by the Bureau of Land Management or protected by Nevada have been found in the
Caliente-Chalk Mountain rail corridor. This rail corridor crosses approximately eight areas designated as
game habitat and one area of waterfowl habitat (TRW 1999b, page 3-27). Two of these sensitive species,
the speckled dace and desert sucker, are restricted to the floodplain of the Meadow Valley Wash. The
designated waterfowl habitat also is generally restricted to the floodplain of Meadow Valley Wash and
adjacent wetlands.
Archaeology: There are 100 archaeological sites that have been recorded along the Caliente-Chalk
Mountain route.
L.3.2.4 Jean Rail Corridor
Flooding: The Jean rail corridor crosses 89 washes en route to the Yucca Mountain site (TRW 1999c,
pages 3 to 4). Approximately five washes along this route are large enough that bridges would be required
to cross them. Floodplains associated with these washes have not been defined at this time.
Wetlands: No springs, perennial streams, or riparian areas that may have associated wetlands have been
identified within 0.4 kilometer (0.25 mile) of the Jean rail corridor (TRW 1999b, page 3-29). However, no
field searches or formal delineations of wetlands have been conducted along this route.
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Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Biology: The desert tortoise is the only threatened or endangered species found along the Jean rail
corridor. This entire corridor is within desert tortoise habitat, but does not cross any areas designated as
critical habitat. The abundance of desert tortoises is low along most of the rail corridor, although there is a
higher abundance along some portions in Ivanpah, Goodsprings, Mesquite, and Pahrump valleys (TRW
1999b, page 3-28). One species, the pinto beardtongue (Penstemon bicolor spp.) that is classified as
sensitive by the Bureau of Land Management has been found within the corridor. This rail corridor
crosses approximately 12 areas designated as game habitat by the Bureau of Land Management (TRW
1999b, page 3-28). None of these species or game habitats are restricted to floodplains or areas that may
have wetlands.
Archaeology: Six archaeological sites have been recorded along the Jean rail corridor.
L.3.2.5 Valley-Modified Rail Corridor
Flooding: The Valley-Modified rail corridor crosses 95 washes en route to the Yucca Mountain site
(TRW 1999c, pages 3 to 4). Approximately three washes along this route are large enough that bridges
would be required to cross them. Floodplains associated with these washes have not been defined at this
time.
Wetlands: No springs, perennial streams, or riparian areas that may have associated wetlands have been
identified within 0.4 kilometer (0.25 mile) of the Valley-Modified rail corridor (TRW 1999b, pages 3-29 to
3-30). However, no field searches or formal delineations have been conducted along this route.
Biology: The desert tortoise is the only threatened or endangered species found along the Valley-Modified
rail corridor. This entire corridor is within desert tortoise habitat, but does not cross any areas designated
as critical habitat. The abundance of desert tortoises is low along this rail corridor (TRW 1999b, page
3-29). Two plant species (Parish's scorpion weed [Phacelia parishii] and Ripley's springparsley) classified
as sensitive by the Bureau of Land Management have been found in the rail corridor. None of these
species are restricted to floodplains or areas that may have wetlands. The Valley-Modified rail corridor
does not cross any Bureau of Land Management-designated game habitat (TRW 1999b, page 3-29).
Archaeology: Nineteen archaeological sites have been recorded along the Valley-Modified rail corridor.
L.3.2.6 Caliente Intermodal Transfer Station
Flooding: The two proposed sites for the Caliente intermodal transfer station are located in the Meadow
Valley Wash south of Caliente. Both areas are outside the inundation boundary of the l(X)-year floodplain,
but within the boundary of the 500-year floodplain.
Wetlands: Part of the proposed station location is moist during at least some portions of the year and may
be classified as wetlands. The adjacent perennial stream and riparian habitat along Meadow Valley Wash
also might be classified as wetlands, although no formal delineation of wetlands has been conducted for
this proposed activity (TRW 1999b, page 3-35).
Biology: No game habitat, threatened or endangered species, or species classified as sensitive by the
Bureau of Land Management or protected by Nevada occur within the proposed station location (TRW
1999b, page 3-35).
Archaeology: Four archaeological sites have been recorded at the Caliente intermodal transfer station
site.
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Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
L.3.2.7 Apex/Dry Lake Intermodal Transfer Station
Flooding: The two proposed sites for the Apex/Dry Lake intermodal transfer station are located outside of
the 100-year and 500-year floodplain.
Wetlands: There are no springs or riparian areas within the proposed station location (TRW 1999b, page
3-36).
Biology: The only resident threatened or endangered species at this site is the desert tortoise. The
abundance of desert tortoises in Dry Lake Valley generally is low, although some areas there have a higher
abundance. One plant species, Geyer's milkvetch {Astragalus geyeri triquetrus), classified as sensitive by
the Bureau of Land Management has been found in the proposed location. Neither of these species are
restricted to floodplains or wetlands. No game habitat has been designated there (TRW 1999b, page 3-36).
Archaeology: Two archaeological sites have been recorded at the Apex/Dry Lake intermodal transfer
station site.
L.3.2.8 Sloan/Jean Intermodal Transfer Station
Flooding: The southernmost proposed site for the Jean intermodal transfer station is located in the same
general area as a 100-year flood inundation zone. The northern site proposed for the Jean intermodal
transfer station is not in an inundation zone and is outside the 500-year floodplain. The northernmost
proposed site for the Sloan intermodal transfer station is in an area with no printed Federal Emergency
Management Agency map and it is outside the 500-year floodplain.
Wetlands: There are no springs or riparian areas within the proposed station location (TRW 1999b, page
3-36).
Biology: The only resident threatened or endangered species at this site is the desert tortoise. The
abundance of desert tortoises in Ivanpah Valley generally is moderate to high, relative to other areas within
the range of this species in Nevada. One plant species, pinto beardtongue, classified as sensitive by the
Bureau of Land Management has been found in the proposed location. Neither of these species are
restricted to floodplains or wetlands. No game habitat has been designated there (TRW 1999b, pages 3-36
to 3-37).
Archaeology: Seven archaeological sites have been recorded at the Sloan/Jean intermodal transfer station
site.
L.4 Floodplain/Wetlands Effects
According to 10 CFR 1022.12(a)(2), a floodplain assessment is required to discuss the positive and
negative, direct and indirect, and long- and short-term effects of the proposed action on the floodplain
and/or wetlands. In addition, the effects on lives and property, and on natural and beneficial values of
floodplains must be evaluated. For actions taken in wetlands, the assessment should evaluate the effects of
the proposed action on the survival, quality, and natural and beneficial values of the wetlands. If DOE
finds no practicable alternative to locating activities in floodplains or wetlands, DOE will design or modify
its actions to minimize potential harm to or in the floodplains and wetlands. The floodplains that are
assessed herein are those areas of normally dry washes that are temporarily and infrequently inundated
from runoff during 100-year or 500-year floods.
L-15
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
L.4.1 FLOODPLAIN/WETLANDS EFFECTS NEAR YUCCA MOUNTAIN
DOE has not determined if rail casks will be transported in Nevada by heavy-haul trucks on existing
highways or whether to construct a branch rail line to bring the spent nuclear fuel and high-level
radioactive waste to the Yucca Mountain site. Near Yucca Mountain, however, it is possible that each of
the four washes could be affected if a rail line and a road were to access the Yucca Mountain site from
different directions. Because of this uncertainty, this assessment examines the configurations that would
cause the most disturbances to the four washes and their floodplains, as follows:
• Potential construction of a heavy-haul-capable road west of Fortymile Wash that crosses Busted Butte
Wash, Drillhole Wash, and Midway Valley Wash. Cut, fill, and drainage culverts could be used to
cross Busted Butte and Drillhole washes. A bridge could be constructed over Midway Valley Wash.
Heavy-haul trucks carrying spent nuclear fuel and high-level radioactive waste could travel along this
road to the repository.
• Potential construction of a raised rail line through Fortymile Wash with appropriately-sized drainage
culverts. The rail line could join the route for heavy-haul trucks north of Drillhole Wash and cross
Midway Valley Wash on a separate rail-bridge before entering the repository. Trains carrying spent
nuclear fuel and high-level radioactive waste could travel along the rail line to the repository.
• Potential upgrading of the existing road that crosses Fortymile Wash with appropriately-sized drainage
culverts. The road could be used by legal-weight trucks to transport spent nuclear fuel and high-level
radioactive waste to the repository, as well as transporting various types of hazardous and non-
hazardous materials to and from the repository.
Construction in the washes would reduce the area through which floodwaters naturally flow. During large
floods, bodies of water could develop on the upstream side of each of the crossings and slowly drain
through culverts. Such floods, however, would not increase the risk of future flood damage, increase the
impact of floods on human health and safety, or harm the natural and beneficial values of the floodplains
because there are no human activities or facilities upstream or downstream that could be affected. A
sufficiently large flood in Fortymile Wash could create a temporary large lake up-stream of the raised rail
line and the legal-weight road. The water would slowly drain through culverts. If the flood occurred
quickly and was sufficiently large, water would flow over the rail line and roads and continue downstream.
Some damage to the rail line and the roads would be expected, but neither structure would increase the risk
of future flood damage, increase the impact of floods on human health and safety, or harm the natural and
beneficial values of the floodplains because there are no human activities or facilities downstream that
could be affected.
During and after each flood, a large amount of sediment would accumulate on the up-stream side of each
crossing. Periodically, this material would have to be removed so that future floods would have sufficient
space to accumulate, rather than overflow the structures during successively smaller floods. This material
would, when deemed necessary, be removed by truck and disposed of appropriately. Under natural
conditions this sediment would have continued downstream and been deposited as the floodwaters
receded. Compared to the total amount of sediment that is moved by the flood water along the entire
length of the washes, the amount trapped behind the crossings would be small.
During a 100-year or 500-year flood, there would be no preferred channels; all channels across the entire
width of each wash would be filled with water (Figure L-1). Therefore, the manmade crossings would not
cause preferential flow in a particular channel or alter the velocity or direction of flow on the floodplains.
L-i6
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Potential construction of a route for heavy-haul trucks or rail line would require the removal of desert
vegetation in the washes and the disturbance of soil and alluvium. These actions could adversely impact
wildlife habitat and individuals, especially the desert tortoise, which is designated as threatened by the Fish
and Wildlife Service. Prior to any construction, a biological survey would be conducted to locate and
remove tortoises that are in the path of construction and other mitigation measures would be conducted as
identified by the Fish and Wildlife Service during consultations under the Endangered Species Act for this
action.
Construction in the floodplains could also affect unidentified cultural resources that may be present. Prior
to any construction, archaeologists would survey the area following the procedure in DOE's Progranmiatic
Agreement with the Advisory Council on Historic Preservation (DOE 1988, page 5).
Potential indirect impacts on flora and fauna include increased emissions of fugitive dust, elevated noise
levels, and increased human activities. Emissions of fugitive dust would be short-term and would not be
expected to significantly affect vegetation or wildlife. Likewise, no significant long-term impacts to
wildlife are expected from the temporary increase in noise during construction. Wildlife displaced during
construction would probably return after construction was completed.
There are no perennial sources of surface water at or downstream from the Yucca Mountain site that would
be affected by the use of a route for heavy-haul trucks or the construction of a rail line. Two small well
ponds with some riparian vegetation occur in Fortymile Wash downstream of the point where Drillhole
Wash enters Fortymile Wash. During a 100- or 500-year flood, both riparian areas would likely be
damaged or destroyed by floodwaters regardless of the existence of the crossings.
Neither the quality nor the quantity of groundwater that normally recharges through Fortymile Wash would
be substantially affected due to the crossings. Water infiltration could increase somewhat after large floods
as standing water slowly enters the ground behind the crossings. The total volume of these water bodies
would be a few acre-feet at most, and much of the water would gradually drain through culverts or
evaporate before reaching the groundwater table at 274 meters (900 feet) below the surface.
The use of petroleum, oil, lubricants, and other hazardous materials during construction would be strictly
controlled and spills would be promptly cleaned up and, if needed, the soil and alluvium would be
remediated. The small amount of these materials that might enter the ground would not affect the
groundwater, which is 274 meters (900 feet) below the surface.
The nearest population center is about 19 kilometers (12 miles) to the south, along U.S. 95 at Lathrop
Wells a few miles east of Fortymile Wash. If floodwaters from a 100- or 500-year flood reached this far
downstream, there would be no measurable increase in flood velocity or sediment load attributable to the
use of a route for heavy-haul trucks or construction of a rail line compared to natural conditions. Hence,
disturbances to the floodplains of Fortymile Wash, Busted Butte Wash, Drillhole Wash, or Midway Valley
Wash would have no adverse impacts on lives and property downstream. Moreover, impacts to these
floodplains would be insignificant in both the short- and long-term compared to the erosion and deposition
that occur naturally and erratically in these desert washes and floodplains.
During operation of the repository it would be extremely unlikely that a truck carrying spent nuclear fuel
and high-level radioactive waste would fall into Busted Butte, Drillhole, or Midway Valley washes or that
a train would derail in Fortymile Wash. However, even if this occurred, the shipping casks, which are
designed to prevent the release of radioactive materials during an accident, would remain intact. The casks
would then be recovered and transported to the repository. No adverse impacts to surface water or
groundwater quality from such accidents would occur.
L-17
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
Hazardous materials needed during construction and operation of the repository would be transported
along the legal-weight access road. If these materials were released during an accident, they would be
cleaned-up quickly and the affected soil and alluvium would be remediated. No adverse impacts to
groundwater quality from such accidents would occur because cleanup could be completed before
contaminants reached the groundwater [the groundwater table is 274 meters (900 feet) below the surface].
There are no positive or beneficial impacts to the floodplains of Busted Butte, Drillhole, Midway Valley,
or Fortymile washes that have been identified from the proposed action.
L.4.2 FLOODPLAIN/WETLANDS EFFECTS ELSEWHERE IN NEVADA
L.4.2.1 Effects along Rail Corridors
Potential rail routes would cross many small, and some large, washes. In general, the impacts caused by
rail construction in any of these washes and their floodplains would be similar in magnitude to those
described for Fortymile, Busted Butte, Drillhole, and Midway Valley washes. Regardless of the route
selected, standard mitigation practices used throughout Nevada for highway construction would be used to
minimize the impacts to floodplains. Most washes and their floodplains along the five potential rail
corridors are in remote areas. Impacts to these floodplains from rail construction and operation would be
insignificant in both the short- and long-term compared to erosion and deposition that occurs naturally and
erratically in these desert washes and floodplains.
Based on current information, springs and riparian areas that may have associated wetlands occur within
three of the rail corridors (Caliente, Carlin, and Caliente-Chalk Mountain). If the rail mode of spent
nuclear fuel and high-level radioactive waste transport is selected by DOE, wetlands delineations along the
selected route would be conducted and the effects would be described in a more detailed
floodplain/wetlands assessment for public review.
L.4.2.2 Effects at Intermodal Transfer Stations
Neither the Dry Lake intermodal transfer station nor the Sloan/Jean intermodal transfer station would have
any impacts on floodplains because these station locations are not in a floodplain. The Caliente intermodal
transfer station, however, is located in Meadow Valley Wash, separated by the Union Pacific Raikoad. If
this site were selected, DOE would conduct a more detailed floodplain/wetlands assessment for public
review to address the floodplain/wetlands effects at the Caliente intermodal transfer station location. The
more detailed floodplain/wetlands assessment would also include potential upgrades to existing roads for
heavy-haul use.
L.5 Mitigation IVIeasures
According to 10 CFR 1022.12(a) (3), agencies must address measures to mitigate the adverse impacts of
actions in a floodplain or wetlands, including but not limited to minimum grading requirements, runoff
controls, design and construction constraints, and protection of ecologically-sensitive areas. Whenever
possible, DOE would avoid disturbing wetlands and floodplains and would minimize impacts to the extent
practicable, if avoidance was not possible. This section discusses the floodplain mitigation measures that
would be considered in the vicinity of Yucca Mountain and elsewhere in Nevada and, where necessary and
feasible, implemented during construction and maintenance in the washes.
Adverse impacts to the affected floodplains would be small. Even during 100- and 500-year floods, it is
unlikely that differences in the rate and distribution of erosion and sedimentation caused by the use of a
L-18
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
route for heavy-haul trucks or construction of a rail line near Yucca Mountain would be measurably
different compared to existing conditions. Nevertheless, DOE would follow their reclamation guidelines
(DOE 1995, pages 2-1 to 2-14) for site clearance, topsoil salvage, erosion and runoff control, recontouring,
revegetation, siting of roads, construction practices, and site maintenance. Disturbance of surface areas
and vegetation would be minimized, and natural contours would be maintained to the maximum extent
feasible. Slopes would be stabilized to minimize erosion. Unnecessary off-road vehicle travel would be
avoided. Storage of hazardous materials during construction would be outside the floodplains.
Before any potential construction could begin, DOE would require pre-construction surveys to make sure
that the work would not impact important biological or archaeological resources. In addition, the site's
reclamation potential would be determined during these surveys. In the event that construction could
threaten important biological or archaeological resources, and modification or relocation of the roads and
rail line is not reasonable, mitigation measures would be developed. Mitigation measures developed
during the pre-construction surveys would be incorporated into the design of the work. These measures
could include relocation of sensitive species, avoidance of archaeological sites, or data recovery if
avoidance is not feasible.
If hazardous materials are spilled during construction of the crossings or during transport to the repository,
the spill would be quickly cleaned-up and the soil and alluvium would be remediated. Hazardous materials
would be stored away from all floodplains to decrease the probability of an inadvertent spill in these areas.
L.6 Alternatives
According to 1022.12(a)(3), DOE must consider alternatives to the proposed action. Alternative ways to
access the Yucca Mountain site are considered in the following paragraphs, along with the no action
alternative.
L.6.1 ALTERNATIVES NEAR YUCCA MOUNTAIN
To operate a potential repository at Yucca Mountain, heavy-haul-capable and legal-weight roads and a rail
line to the facility would be considered so the spent nuclear fiiel and high-level radioactive waste could be
unloaded and emplaced underground. It is unreasonable to consider a railroad or heavy-haul-capable and
legal-weight roads that access the repository directly from the west over Yucca Mountain because of
engineering constraints, environmental damage, and cost associated with construction in such rugged
terrain. Because of these concerns, this alternative was eliminated from detailed consideration.
Access to Yucca Mountain from the east side requires that Fortymile Wash be crossed. Alternative sites
for these crossings were considered, but the impacts at any alternative site would be virtually identical to
the proposed site. Moreover, the proposed sites provide the most direct routes to the repository and would
cost less to build and/or upgrade than alternative sites that cross Fortymile Wash at wider locations.
L.6.2 ALTERNATIVE RAIL CORRIDORS AND ALTERNATIVE SITES FOR AN INTERMODAL
TRANSFER STATION
Five potential rail corridors were identified by DOE through a winnowing process that considered a host of
environmental constraints (see Chapter 2, Section 2.3.3). Other possible rail corridors in Nevada were
examined but rejected because of such things as land use, private land, and engineering constraints.
Identification of the three intermodal transfer station locations was limited to reasonable sites next to an
existing rail line in Nevada. Other sites were considered by DOE, but rejected because of ownership and
environmental concerns.
L-19
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
L.6.3 NO-ACTION ALTERNATIVE
Selection of the No- Action Alternative would avoid impacts to floodplains and wetlands. If Yucca
Mountain was selected as a site to construct a repository, transport of spent nuclear fuel and high-level
radioactive waste to the Yucca Mountain site would be required. In that case there would be no other
practicable alternative to taking action in floodplains and wetlands because there would be no way to
transport spent nuclear fuel and high-level radioactive waste to the Yucca Mountain site during repository
operation without passing through some wetlands areas and floodplains.
L.7 Conclusions
DOE prepared this assessment in compliance with 10 CFR Part 1022. The assessment evaluates the
effects to the floodplains near Yucca Mountain (Fortymile Wash, Busted Butte Wash, Drillhole Wash, and
Midway Valley Wash) and generically to floodplains and wetlands elsewhere in Nevada from construction
of a rail line or an intermodal transfer station and associated upgrades to existing highways for heavy-haul
trucks.
Near Yucca Mountain, the closest man-made structure within Fortymile Wash is U.S. 95 more than 19
kilometers (12 miles) south of the confluence of Drillhole and Fortymile washes. Lathrop Wells, the
nearest population center to Yucca Mountain, is also about 19 kilometers to the south along U.S. 95 and
two miles east of Fortymile Wash. Construction- and operations-related impacts to the 100-year and
500-year floodplains of Fortymile Wash, Busted Butte Wash, Drillhole Wash, and Midway Valley Wash
would be small. None of these impacts would increase the risk of future flood damage, or increase the
impact of floods on human health and safety, or harm the natural and beneficial values of the floodplains.
There are no positive or beneficial impacts to the floodplains of Busted Butte, Drillhole, Midway Valley,
or Fortymile washes from the proposed actions that have been identified.
Elsewhere in Nevada, effects to floodplains and wetlands would probably be small, although a detailed
floodplain/wetlands assessment would be conducted by DOE when more information is available upon
selection of a rail corridor or route for heavy-haul trucks.
REFERENCES
Blanton 1992 Blanton, J. O., m, 1992, Nevada Test Site Flood Inundation Study: Part
of a Geological Survey Flood Potential and Debris Hazard Study, Yucca
Mountain Site for the U.S. Department of Energy (Office of Civilian
Radioactive Waste Management), Bureau of Reclamation, U.S.
Department of the Interior, Denver, Colorado. [230563]
DOE 1988 DOE (U.S. Department of Energy), 1988, Programmatic Agreement
Between the United States Department of Energy and the Advisory
Council on Historic Preservation for the Nuclear Waste Deep Geologic
Repository Program, Yucca Mountain, Nevada, Yucca Mountain Site
Characterization Office, Nevada Operations Office, North Las Vegas,
Nevada. [HQX. 19890426.0057]
DOE 1991 DOE (U.S. Department of Energy), 1991, Floodplain Assessment of
Surface-Based Investigations at the Yucca Mountain Site, Nye County,
Nevada, YMP/91-11, Yucca Mountain Site Characterization Office, Las
Vegas, Nevada. [MOL. 19990607.0238]
L-20
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
DOE 1992
DOE 1995
DOE 1998a
DOE 1998b
FWS 1995
Hansen et al. 1997
Rautenstrauch, Brown, and
Goodwin 1994
Squires and Young 1984
Steen et al. 1997
TRW 1997
TRW 1998a
DOE (U.S. Department of Energy), 1992, Floodplain Assessment of Site
Characterization Activities at the Yucca Mountain Site, Nye County,
Nevada, YMP/92-30, Yucca Mountain Site Characterization Project
Office, Las Vegas, Nevada. [NNA. 19921028.0084]
DOE (U.S. Department of Energy), 1995, Reclamation Implementation
Plan, YMP/91-14, Revision 1, Las Vegas, Nevada.
[MOL. 19960222.02 18]
DOE (U.S. Department of Energy), 1998a, "Potential Rail Alignments,"
map, YMP-98- 104.0, Office of Civilian Radioactive Waste
Management, Yucca Mountain Project Office, Las Vegas, Nevada.
[MOL. 19990526.0034]
DOE (U.S. Department of Energy), 1998b, "Nevada Routes for Heavy-
Haul Truck Shipments of SNF and HLW to Yucca Mountain," map,
YMP 97-263.9, Office of Civilian Radioactive Waste Management,
Yucca Mountain Project Office, Las Vegas, Nevada.
[MOL. 19990526.0035]
FWS (Fish and Wildlife Service), 1995, Death Valley Nevada
1 :250,000-scale Wetland Map of National Wetlands Inventory, U.S.
Department of the Interior, St. Petersburg, Florida. [244053]
Hansen, D. J., P. D. Greger, C. A. Wills, and W. K. Ostler, 1997,
Nevada Test Site Wetlands Assessment, DOE/NV/1 1718-124, Ecological
Services, Bechtel Nevada Corporation, Las Vegas, Nevada. [242338]
Rautenstrauch, K. R., G. A. Brown, and R. G. Goodwin, 1994, The
Northern Boundary of the Desert Tortoise Range on the Nevada Test
Site, Report 1 1265-1 103, EG&G Energy Measurements, Inc., Las
Vegas, Nevada. [240498]
Squires, R. R., and R. L. Young, 1984, Flood Potential ofFortymile
Wash and Its Principal Southwestern Tributaries, Nevada Test Site,
Southern, Nevada, WRI-834001, U.S. Geological Survey, U.S.
Department of the Interior, Carson City, Nevada. [203214]
Steen, D. C, D. B. Hall, P. D. Greger, and C. A. Wills, 1997,
Distribution of the Chuckwalla, Western Burrowing Owl, and Six Bat
Species on the Nevada Test Site, DOE/NV/1 1718-149, Bechtel Nevada
Corporation, Las Vegas, Nevada. [242253]
TRW (TRW Environmental Safety Systems Inc.), 1997, The
Distribution and Relative Abundance of Desert Tortoises at Yucca
Mountain, BOOOOOOOO-0 17 17-5705-00033, Las Vegas, Nevada.
[MOL. 19980 123. 0643]
TRW (TRW Environmental Safety Systems Inc.), 1998a, Classification
and Map of Vegetation at Yucca and Little Skull Mountains,
BOOOOOOOO-017 17-5705-00083, Revision OOB, Las Vegas, Nevada.
[MOL. 199902 11.0519]
L-21
Floodplain/Wetlands Assessment for the Proposed Yucca Mountain Geologic Repository
TRW 1998b
TRW 1998c
TRW 1999a
TRW 1999b
TRW 1999c
USGS 1983
TRW (TRW Environmental Safety Systems Inc.), 1998b, Bats of Yucca
Mountain, Nevada, BOOOOOOOO-017 17-5705-00050, Revision 02, Las
Vegas, Nevada. [MOL.19981014.0308]
TRW (TRW Environmental Safety Systems Inc.), 1998c, Species
Composition and Abundance of Reptile Populations in Selected Habitats
at Yucca Mountain, Nevada, with Annotated Checklist, BOOOOOOOO-
017 17-5705-00038, Revision 00, Las Vegas, Nevada.
[MOL.199812014.0305]
TRW (TRW Environmental Safety Systems Inc.), 1999a, Environmental
Baseline File for Soils, BOOOOOOOO-017 17-5700-00007, Revision 00,
Las Vegas, Nevada. [MOL. 19990302.0 180]
TRW (TRW Environmental Safety Systems Inc.), 1999b, Environmental
Baseline File for Biological Resources, BOOOOOOOO-01717-5700-00009,
Revision 00, Las Vegas, Nevada. [MOL.19990302.0181;
MOL. 19990330.0560, map attachments)
TRW (TRW Environmental Safety Systems Inc.), 1999c, Nevada
Transportation Engineering File, Las Vegas, Nevada.
[MOL. 19990324.0257]
USGS (U.S. Geological Survey), 1983, Busted Butte Quadrangle,
Nevada-Nye County, 7.5-Minute Series Topographic Map, U.S.
Department of the Interior, Denver, Colorado. [101711]
L-22
CONVERSIONS
METRIC TO ENGLISH
ENGLISH TO METRIC
Multiply
by
To get
Multiply
by
To get
Area
Square meters
10.764
Square feet
Square feet
0.092903
Square meters
Square kilometers
247.1
Acres
Acres
0.0040469
Square kilometers
Square iciiometers
0.3861
Square miles
Square miles
2.59
Square kilometers
Concentration
Kilograms/sq. meter
0.16667
Tons/acre
Tons/acre
0.5999
Kilograms/sq. meter
Milligrams/liter"
1
Parts/million
Parts/ million''
1
Milligrams/liter
Microgram s/liter°
1
Parts/billion
Parts/billion"
1
Micrograms/liter
Micrograms/cu. meter"
1
Parts/trillion
Parts/trillion"
1
Micrograms/cu. meter
Density
Grams/cu. cm
62.428
Pounds/cu. ft.
Pounds/cu. ft.
0.016018
Grams/cu. cm
Grams/cu. meter
0.0000624
Pounds/cu. ft.
Pounds/cu. ft.
16,025.6
Grams/cu. meter
Length
Centimeters
0.3937
Inches
Inches
2.54
Centimeters
Meters
3.2808
Feet
Feet
0.3048
Meters
Kilometers
0.62137
Miles
Miles
1.6093
Kilometers
Temperature
Absoluie
Degrees C+ 17.78
1.8
Degrees F
Degrees F - 32
0.55556
Degrees C
Relative
Degrees C
1.8
Degrees F
Degrees F
0.55556
Degrees C
Velocity/Rate
Cu. meters/second
2118.9
Cu. feetyminute
Cu. feet/minute
0.00047195
Cu. meters/second
Grams/second
7.9366
Pounds/hour
Pounds/hour
0.126
Grams/second
Meters/second
2.237
Miles/hour
Miles/hour
0.44704
Meters/second
Volume
Liters
0.26418
Gallons
Gallons
. 3.78533
Liters
Liters
0.035316
Cubic feet
Cubic feet
28.316
Liters
Liters
0.001308
Cubic yards
Cubic yards
764.54
Liters
Cubic meters
264.17
Gallons
Gallons
0.0037854
Cubic meters
Cubic meters
35.314
Cubic feet
Cubic feet
0.02831-
Cubic meters
Cubic meters
1.3079
Cubic yards
Cubic yards
0.76456
Cubic meters
Cubic meters
0.0008107
Acre-feet
Acre- feet
1233.49
Cubic meters
Weight/Mass
Grams
0.035274
Ounces
Ounces
28.35
Grams
Kilograms
2.2046
Pounds
Pounds
0.45359
Kilograms
Kilograms
0.0011023
Tons (short)
Tons (short)
907.18
Kilograms
Metric tons
1.1023
Tons (short)
Tons (short)
0.90718
Metric tons
ENGLISH TO ENGLISH
Acre-feet
325,850.7
Gallons
Gallons
0.000003046
Acre-feet
Acres
43,560
Square feet
Square feet
0.000022957
Acres
Square miles
640
Acres
Acres
0.0015625
Square miles
a. These widely used conversions are only valid under specific
temperature and
pressure conditions.
METRIC
PREFIXES
Prefix
Symbol
Multiplication factor
exa-
E
1,000,000,000,000,000,000 =
I0'«
peta-
P
1,000,000,000,000,000 =
10'-'
tera-
T
1,000,000,000,000 =
lO''
giga-
G
1,000,000,000 =
lO'
mega
M
1,000,000 =
lO'
kilo-
k
1,000 =
lO'
deca-
D
10 =
lO'
deci-
d
0.1 =
10"
centi-
c
0.01 =
10"
1
milli-
m
0.001 =
10"
3
micrc
-
H
0.000 001 =
10"
6
nano-
n
0.000 000 001 =
10"
9
pico-
P
0.000 000 000 001 =
10"
12
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