United States
Environmental Protection
Agency
Air and Radiation
(6808J)
EPA402-R-01-004
June 2001
Public Health and
Environmental Radiation
Protection Standards for
Yucca Mountain, Nevada -
Final 40 CFR197
Evaluation of Potential Economic
Impacts of 40 CFR Part 197
(ECONOMIC IMPACT ASSESSMENT)
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TABLE OF CONTENTS
LIST OF ACRONYMS , v
EXECUTIVE SUMMARY ES-1
CHAPTER 1.0 EVOLUTION OF REGULATORY 1-1
1.1 EPA Action aad Authority 1-1
1.2 Role of this Document 1-1
1.3 40CFRPart 197 , 1-2
1.4 Legislative History 1-2
1.5 40 CFRPart 191 1-5
1.6 The National Academy of Sciences' Recommendations 1-8
1.7 Final 40 CFR Part 197 - Public Health and Environmental
Radiation Protection Standards for Yucca Mountain, Nevada ... 1-11
1.7.1 Individual-Protection Standard 1-11
1.7.2 Human-Intrusion Standard 1-12
1.7.3 Ground Water Protection Standards 1-14
1.7.4 Site-Specific Regulatory Requirements 1-16
CHAPTER 2.0 OVERVIEW OF RADIOACTIVE WASTE DISPOSAL AT YUCCA
MOUNTAIN , 2-1
2.1 Yucca Mountain as a Disposal Site 2-1
2.2 Sources and Characteristics of Radioactive Wastes to
Be Disposed , 2-1
2.3 Overview of the Repository for Disposal 2-3
2.4 DOE Estimate of the Repository Program Cost 2-4
CHAPTER 3.0 EVOLUTION OF THE YUCCA MOUNTAIN REPOSITORY
DESIGN 3-1
3.1 The 1988 Site Characterization Plan 3-2
3.1.1 Regulatory Framework forme SCP 3-3
3.1.2 Principal SCP Repository Design and Natural System
Features 3-4
3.1.3 The SCP Engineered Barrier System 3-7
3.2 Design Options in the Total System Performance Assessments of
1991,1993, and 1995 ........; 3-8
3.2.1 TSPA-1991 3-9
3.22 TSPA-1993 3-9
3.2.2.1 M&O Version of TSPA-93 3-10
3.2.2.2 SNL Version of TSPA-93 3-11
3.2.3 TSPA-1995 3-12
3.3 Design Features for the Viability Assessment - 1998 3-14
3.4 Enhanced Design Alternatives - 1999 3-16
3.4.1 Basis for the Current Design 3-17
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TABLE OF CONTENTS
(continued)
3.4.2 Selection of the Repository Design for the Site
Recommendation 3-18
3.4.3 Comparison of the EDA II and Viability Assessment
Designs 3-20
3.5 Evolution of the Comparative Contributions of Engineered and
Natural Barriers to Repository System Performance 3-21
3.6 Summary of Factors Affecting Evolution of the Repository
Design 3-25
3.7 EDA II Design and the TSPA-SR 3-26
3.7.1 New Approaches in the TSPA-SR 3-27
3.7.1.1 The Nominal Scenario 3-27
3.7.1.2 Igneous Scenarios 3-30
•3.7.1.3 Human Intrusion Scenario 3-33
3.7.2 Results of the TSPA-SR 3-35
3.8 DOE's Current Program Costs 3-37
CHAPTER 4.0 EVOLUTION OF PERFORMANCE ASSESSMENT AND BARRIER
ROLES 4-1
4.1 Performance in Comparison with the Individual-Protection
Standard 4-1
4.2 Performance in Comparison with the Ground Water Protection
Standards 4-3
4.3 Conservatism in the TSPA-VA, TSPA-DEIS, AND TSPA-SR
Evaluations 4-6
4.3.1 Assessment of Juvenile Failure 4-7
4.3.2 Local Crevice Corrosion of Alloy 22 4-8
4.3.3 Water Flow Into the Package Interior 4-9
4.3.4 Exposed Waste Form Area 4-11
4.3.5 In-Package Dilution and Transport Delays 4-13
4.4 Radiation Doses to Alternative Receptors 4-15
4.5 Alternative Means to Reduce Uncertainties and Doses 4-19
4.6 Current Repository Design and Safety Strategy 4-21
CHAPTER 5.0 EPA'S "REASONABLE EXPECTATION" APPROACH TO
REPOSITORY PERFORMANCE PROJECTIONS 5-1
5.1 Overview of Reasonable Expectation 5-1
5.2 Prior Consideration and Use of Reasonable Expectation 5-2
5.3 Comparison of Reasonable Expectation and Reasonable
Assurance 5-3
5.4 Use of Reasonable Expectation for Yucca Mountain 5-6
5.5 Impact of Implementation of Reasonable Expectation for Yucca
Mountain 5-10
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TABLE OF CONTENTS
(continued)
CHAPTER 6.0 COST IMPACTS OF THE STANDARDS IN THE RULE 6-1
6.1 The Individual-Protection Standard 6-1
6.2 Cost Impacts of the HIS Requirements 6-3
6.3 Cost Impact of the GWS Requirements 6-5
CHAPTER 7.0 SUMMARY DEMONSTRATION THAT THE EPA STANDARDS
HAVE NO COST IMPACTS ON THE YUCCA MOUNTAIN
PROGRAM AND REPOSITORY 7-1
7.1 Principal Bases for Findings of No Cost Impacts 7-1
7.1.1 Evolution of the Repository Design and Roles of
Natural and Engineered Features 7-2
7.1.2 DOE's Use of Performance Evaluations 7-2
7.1.3 Impact of the EPA Standards on Data and Analysis
Requirements 7-3
7.2 Comparative Impacts of Alternative Dose Limits for the
Individual-Protection Standard 7-4
7.3 Summary and Conclusions .... 7-5
CHAPTER8.0 REFERENCES* 8-1
* NOTE: This document has been revised to correct minor errors which were contained in the version placed in the
public docket for the Yucca Mountain Rule (docket #A-95-12, V-B-2). A listing of those changes is
located at the end of this document.
in
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LIST OF TABLES
No. Page
3-1. Repository Designs Evaluated by SNL in TSPA-1993 3-11
3-2. Spent Fuel Waste Package Inventory for TSPA-1993 3-12
3-3. Principal Results of EDA Analysis 3-19
3-4. EDA II/VA Design Comparison 3-20
3-5. Impact of EDA II Design Features on Performance Uncertainties 3-21
3r6. Implementation of Regulatory Requirements in the TSPA-SR for Regulatory
Requirements 3-34
3-7. Technical Assumptions Implemented in the Human Intrusion Scenario in
TSPA-SR 3-36
3-8. Estimates of Costs for the Yucca Mountain Program 3-39
4-1. Comparison of DEIS Ground Water Radionuclide Concentrations with MCLs ... 4-4
4-2. Change Over Time of the Roles of Natural and Engineered Barriers in
Repository System Performance 4-22
6-1, Data and Analysis Requirements for Assessing Compliance With the Human-
Intrusion Standard 6-4
6-2. Data and Analysis Requirements for Assessing Compliance With the Ground
Water Protection Standards 6-6
LIST OF FIGURES
No. Page
ES-1. Comparison of Radiation Protection Standards with Expected Values of
TSPA-SR Calculations for a Repository at Yucca Mountain for Nominal and
Igneous Scenarios ES-3
2-1. Sources of Radioactive Wastes for the Yucca Mountain Repository 2-2
3-1. Layout of the Site Characterization Plan Repository 3-5
3-2. Comparison of Radiation Protection Standards with Expected Values of
TSPA-SR Calculations for a Repository at Yucca Mountain for Nominal and
Igneous Scenarios 3-37
3-3. Expected Values of TSPA-SR Calculations for a Repository at Yucca
Mountain for the Inadvertent Human Intrusion Scenario 3-38
4-1. Summary of Groundwater Protection Performance Results of the TSPA-SR:
Combined Beta and Photon-Emitting Radionuclides 4-4
4-2. Summary of Ground-Water Protection Results for TSPA-SR for Gross Alpha
Activity 4-5
4-3. Estimates of the Consequence of an Artificial Juvenile Failure 4-7
4-4. 10,000-Year Dose-Rates for Alternative Areal Mass Loadings 4-9
4-5. Tc-99 Concentrations for Alternative Mass Loadings 4-16
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LIST OF ACRONYMS
BDCF Biosphere Dose Conversion Factor
BID Background Information Document
BWR Boiling Water Reactor
CAM Corrosion Allowance Material
CEDE Cumulative Effective Dose Equivalent
CFR Code of Federal Regulations
CRM Corrosion Resistant Material
CSNF Commercial Spent Nuclear Fuel
DEIS Draft Environmental Impact Statement
DOE Department of Energy
BBS Engineered Barrier System
EDA Enhanced Design Alternative
EDE Effective Dose Equivalent
EIA Economic Impact Analysis
EnPA Energy Policy Act
EPA Environmental Protection Agency
EPRI Electric Power Research Institute
ESF Exploratory Studies Facility
FEHM Finite Element Heat and Mechanical Model
GWS Ground Water Protection Standard
HIS Human-Intrusion Protection Standard
HLW High-Level Waste
ICRP International Commission on Radiation Protection
IPS Individual-Protection Standard
LADS License Application Design Selection
MTHM Metric Tonnes of Heavy Metal
MTU Metric Tonnes of Uranium
MWd Megawatt Days
NAS National Academy of Sciences
NRC Nuclear Regulatory Commission
NWPA Nuclear Waste Policy Act
NWPAA Nuclear Waste Policy Amendments Act
NWTRB Nuclear Waste Technical Review Board
OCRWM Office of Civilian Radioactive Waste Management (DOE)
PWR Pressurized Water Reactor
RA Reasonable Assurance
RE Reasonable Expectation
RMEI Reasonably Maximally Exposed Individual
SCC Stress Corrosion Cracking
SCP Site Characterization Plan
SDWA Safe Drinking Water Act
SR Site Recommendation
SZ Saturated Zone
TSLCC Total System Life Cycle Costs
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TSPA Total System Performance Assessment
UZ Unsaturated Zone
VA Viability Assessment
WIPP Waste Isolation Pilot Plant
VI
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EXECUTIVE SUMMARY
This Economic Impact Assessment (EIA) demonstrates that DOE's strategy for development and
design of a possible repository at Yucca Mountain has evolved so that the EPA's 40 CFR
Part 197 standards will have no impact on costs of the repository or the repository development
program. It also shows that the EPA's generic 40 CFR Part 191 standards, as well as the 40 CFR
Part 197 site-specific standards, did not influence evolution of the DOE program or the
repository design.
The EIA analysis uses three major, converging perspectives to support the conclusion that the
EPA standard for Yucca Mountain does not impose additional costs on the DOE program:
• An historical perspective in Chapter 3 traces the evolution of the repository design from
principal reliance for safety performance on natural features to principal reliance on
engineered features and the factors that influenced it This discussion concludes that the
inversion of performance roles of the natural and engineered features of the disposal
system has evolved as a result of site characterization findings, guidance from external
reviews such as those of the Nuclear Waste Technical Review Board, and evolution of
strategy for dealing with uncertainties. This discussion demonstrates that evolution
of the repository design has been independent of the EPA standards, the major
components of which have remained essentially unchanged since the 1985
promulgation of the generic 40 CFR Part 191 standards for geologic disposal.
* A performance assessment perspective in Chapter 4 traces the evolution of strategy to
achieve performance, the evolution of identification and characterization of factors that
contribute to performance, and the approach to identifying and reducing uncertainties that
are important to demonstration of compliance with standards. The discussion includes
DOE estimates of performance for the current repository design which show that,
under nominal conditions, there will be no radionuclide releases and no potential
for radiation doses for more than 10,000 years after repository closure.
The new repository design was not developed to respond to any provisions of the EPA
standard, but rather to reduce or eliminate uncertainties in the very conservative
performance assessments of the previous design. Relative to the "reasonable expectation"
approach to implementation that is included in the standard (described in more detail in
this document), the previous assessments of the older design are considered to illustrate
the impact of reasonable expectation on repository design and performance assessments.
ES-1
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* An information-needs perspective assesses the data and analyses needed to address the
IPS, GWS, and HIS components of the EPA standard, with emphasis on whether
resources beyond those needed to address the individual-protection standard, which is
fundamental to radiation protection, are needed to address the GWS and HIS standards.
This EIA demonstrates that tbe data and analysis requirements for assessing
compliance with the ground water protection and human-intrusion standards are
the same as those required for assessing compliance with the fundamental and
essential individual-protection standard. The ground water protection standard
and the human-intrusion standard, therefore, impose no incremental cost impacts,
Comparative Impacts of Alternative Dose Limits for the Individual-Protection Standard
A contentious issue in developing the individual-protection standard has been comparative
impacts of alternative dose limits, e.g., 15 millirem/year (mrem/yr) versus 25 mrem/yr. Figure
ES-1, which shows the performance projections for the newest repository design (EDA EL),
under conditions of expected performance, provides an important perspective on the dose limit
issue. Doses in the period less than 10,000 years are entirely the result of a very low probability
(the mean annual probability is 1.6xlO"8) potential igneous disruption of the disposal facility. A
very small downward shift in estimates of probability would eliminate this scenario from
consideration altogether. In addition, the consequences associated with potential releases from
igneous activity appears to be treated in an extremely conservative manner. Alternative
assumptions are possible that would eliminate releases associated with igneous activity entirely,
even in the unlikely event that such activity occurs.
The nominal scenario represents an assessment of the function of the repository when only
gradual degradation processes occur. This scenario does not lead to any releases in the first
10,000 years, despite a significant level of conservatism built into the model. The current model
of the current repository design shows lower consequences at longer times than did earlier
iterations of the TSPA. Significantly, even these earlier iterations (e.g. TSPA-VA), which
contained extremely conservative assumptions about juvenile failures of waste containers, were
able to comfortably comply with either of the alternative individual-protection standards.
As seen in Figure ES-1, the EDA II repository design demonstrates performance such that
projected doses are significantly less than either the 15 mrem/yr or the 25 mrem/yr dose limit.
Furthermore, for nominal behavior of the repository, there are no projected doses during the first
10,000 years. It is therefore evident that selection of a 15 mrem/yr dose limit rather than a
25 mrem/yr limit will not impose any additional cost impacts on the repository. This is a highly
ES-2
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£
o
CO
c
fO
105-
104
103
102
10°
10-1
10-3
INRC Proposed 25 mrem/yr IPS
"ERA'S IPS and HIS
ERA'S GWS
Combined Mean Dose Rate
.. .....
Igneous Mean
Dose Rate
/Nominal Mean Dose Rate
i i i i i i i
1000
10,000
Time (years)
100,000
Figure ES-1. Comparison of Radiation Protection Standards with
Expected Values of TSPA-SR Calculations for a Repository
at Yucca Mountain for Nominal and Igneous Scenarios.
significant finding in that the 15 mrem/yr CEDE dose limit is consistent with the
recommendations of the National Academy of Sciences and regulatory precedents.
Conclusions
The information presented in this EIA has demonstrated that the design of a repository for
disposal of radioactive wastes at Yucca Mountain has evolved without having been affected by
the EPA standards. The standards have been demonstrated to have no impact on repository
program costs, and nominal performance for the current repository design would result in no
radiation doses for more than 10,000 years. Additionally, the difference between a 25 mrem/yr
dose standard and a 15 mrem/yr standard is insignificant to program costs and performance
evaluations.
ES-3
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ES-4
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1.0 EVOLUTION OF REGULATORY REQUIREMENTS
This chapter describes the basis for this rulemaking and provides a brief history of EPA's
regulatory authority and prior rulemaking actions concerning disposal of radioactive wastes.
It demonstrates that this rulemaking is derived from provisions of the Energy Policy Act of
1992. Standards for individual protection and human intrusion are based on recommenda-
tions made by the National Academy of Sciences, and ground water protection standards are
based on the Safe Drinking Water Act and regulatory precedents.
1.1 EPA Action and Authority
The U.S. Environmental Protection Agency (EPA), pursuant to Section 801 of the Energy
Policy Act of 1992 (EnPA) has issued a rale, 40 CFR Part 197, which contains standards for the
protection of the public from releases of radioactive materials stored or disposed of in a
repository at the Yucca Mountain site in Nevada. This document was prepared to evaluate the
economic impact of this rule.
The rule contains three principal component standards: Individual-Protection Standard (IPS),
Human-Intrusion Standard (HIS), and Ground Water Protection Standards (GWS). Details of
the evolution of the rale and these standards are described in Section 3 of this document.
1.2 Role of this Document
This document describes, in detail, the basis for, and results of, the assessment of economic
impacts of the standards on the costs of storage and disposal of radioactive wastes at Yucca
Mountain.
The document traces the history of evolution of the Yucca Mountain repository design, from the
early use of a small, thin-walled canister, and repository features that were expected to dominate
safety performance reflecting ground water travel times of tens of thousands of years (circa
1988), to the current design, in which engineered features (consisting of drip shields and large,
multi-walled waste packages) dominate performance, and are expected to maintain radionuclides
in isolation for at least 10,000 years (TRWOO). The document also discusses the evolution of
performance assessments and the inversion of roles of engineered and natural barriers., the EPA's
"Reasonable Expectation" approach to performance projections and compliance decisions, and
the overall impact of the standards on Yucca Mountain costs.
This document will demonstrate that the repository design evolved not in response to the
expected provisions of the standard, but in response to improved understanding of the natural
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and engineered barrier interactions and performance expectations, as a result of 12 years of site
characterization, performance assessment and design activities performed by the DOE. The
uncertainties identified by DOE's efforts over this period could be addressed by either
developing enhanced engineering design alternatives to reduce or eliminate the uncertainties, or
by investing time and resources in more extensive characterization and testing studies. DOE has
leaned toward enhanced engineering, at least in part because inherently some uncertainties about
the characteristics and behavior of the natural system may not be amenable to unequivocal
reduction or elimination even with extensive field and laboratory testing.
1.3 40 CFR Part 197
The remainder of this chapter describes the evolution of the 40 CFR Part 197 regulation and the
rationale underlying its development. The U.S. Environmental Protection Agency (EPA) is
responsible for developing and issuing environmental standards and criteria to ensure that public
health and the environment are adequately protected from potential radiation impacts. The
regulation contains site-specific environmental standards to protect public health from releases
from radioactive materials disposed of or stored in the potential repository to be constructed at
Yucca Mountain in Nevada**. These standards provide the basic framework to control the long-
term storage and disposal of radioactive wastes at Yucca Mountain.
Other radioactive materials that could be disposed of in the Yucca Mountain repository include
highly radioactive low-level waste, known as greater-than-Class-C waste, and excess plutonium
resulting from the dismantlement of nuclear weapons.
Emphasis in this document is on the major components of the Yucca Mountain standard, namely
the Individual-Protection Standard (IPS), the Human-Intrusion Standard (HIS), and the Ground
Water Protection Standard (GWS). In reviewing the development of the current standard
attention will be devoted primarily to these components.
1.4 Legislative History
EPA has the authority to set generally applicable environmental standards for radioactive
releases under the Atomic Energy Act (AEA) of 1954, as amended (AEA54), and the President's
" No decision has been made regarding the acceptability of Yucca Mountain for storage or disposal. In this
document, the characterization of the Yucca Mountain repository as "potential" is often omitted but always
intended.
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Reorganization Plan No. 3 of 1970 (NIX70). The basic authority under the AEA, as transferred
to the EPA by Reorganization Plan No 3, includes the mandate of:
...establishing generally applicable environmental standards for the protection of
the general environment from radioactive materials. As used herein, standards
mean limits on radiation exposures or levels, or concentrations or quantities of
radioactive material, in the general environment outside the boundaries of
locations under the control of persons possessing or using radioactive materials
(AEA54).
In 1982, the Nuclear Waste Policy Act (NWPA) (Public Law 97-425) established formal
procedures regarding the evaluation and selection of sites for geologic repositories, including
procedures for the interaction of state and Federal Governments. The Act assigned the U.S.
Department of Energy (DOE) the responsibility of siting, building, and operating an
underground geologic repository for the disposal of these wastes, established provisions for the
selection of at least two independent repository sites, and limited the quantity of wastes to be
disposed of in the initial repository to 70,000 metric tons of heavy metal (MTHM)"*. The
NWPA also reiterated the existing responsibilities of the Federal agencies involved in the
national program (see AEA authority above) and provided a timetable for several key milestones
to be met by the Federal agencies. The NWPA also directed that EPA, pursuant to its authorities
under other provisions of law, was required to:
by rule, promulgate generally applicable standards for the protection of the
general environment from off-site releases from radioactive material in
repositories (NWP83).
The basic authority for EPA to establish environmental standards for the repository effort
originates from these sources.
In September 1985, EPA published 40 CFR Part 191, "Environmental Standards for the
Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic Radioactive
Wastes" (EPA85). These standards were generic and intended to apply to all sites for the deep
geologic disposal of high-level radioactive waste. In 1987, the U.S. Court of Appeals for the
First Circuit responded to a legal challenge by remanding Subpart B of the 1985 standards (the
disposal standards) to the Agency for further consideration. This regulation, which is of
considerable importance to the development of 40 CFR Part 197, will be discussed further in the
next section.
*** This is a measure of the uranium content of the spent nuclear fuel to be emplaced in the repository.
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In December 1987, Congress enacted the Nuclear Waste Policy Amendments Act (NWPAA).
The 1987 Amendments Act redirected the nation's nuclear waste program to evaluate the
suitability of only the Yucca Mountain site as the location for the first high-level waste and spent
nuclear fuel repository (NWP87), An important program change instituted by the Amendments
Act was establishment of the Nuclear Waste Technical Review Board (NWTRB). The NWTRB
was charged with providing independent technical and scientific review of the OCRWM
program. It consists of experts in various disciplines (about 10, but limited to 22) and has a
small support staff. Members of the NWTRB are appointed by the President of the United
States. The opinions and recommendations of the NWTRB have played a significant role in the
development of the repository design, as will be pointed out in other sections of this document
The NWPAA, while dramatically changing the scope and focus of the repository effort, did not
affect or alter EPA's role, i.e., to develop the environmental standards for deep geological
disposal.
In October 1992, the Waste Isolation Pilot Plant Land Withdrawal Act (WIPP LWA, Public Law
102-579) was enacted. While reinstating certain sections of the Agency's 1985 disposal
standards, the Act exempted the Yucca Mountain site from these generic disposal standards
(WIP92). In its stead, the Energy Policy Act (EnPA) of 1992 was enacted (Public Law 102-
482), which established EPA's authority to develop standards for environmental releases specific
to Yucca Mountain.
Section 801 of the EnPA directed EPA to promulgate standards to ensure protection of public
health from releases of radioactive material from a deep geologic repository to be built at Yucca
Mountain (EnP92). EPA must set standards to ensure protection of the health of the public. The
EnPA also required EPA to contract with the National Academy of Sciences (MAS) to advise the
Agency on the technical bases for the Yucca Mountain standards. These EPA standards will
apply only to the Yucca Mountain site and are to be developed based upon and consistent with
the findings and recommendations of the NAS:
...the Administrator shall, based upon and consistent with the findings and
recommendations of the National Academy of Sciences, promulgate, by rule,
public health and safety standards for protection of the public from releases from
radioactive materials stored or disposed of in the repository at the Yucca
Mountain site. Such standards shall prescribe the maximum annual effective dose
equivalent to individual members of the public from releases to the accessible
environment from radioactive materials stored or disposed of in the repository
(EnP92).
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1.5 40 CFR Part 191
The 1985 EPA standards for the management and disposal of spent nuclear fuel and high-level
and transuranic waste were divided into two main sections, Subparts A and B (EPA85).
Subpart A, which addressed the management and storage of waste, limited radiation exposure to
any member of the general public to 25 millirem (mrem) to the whole body and 75 mrem to any
critical organ for disposal facilities operated by the Department of Energy, but not regulated by
the NRC or an Agreement State. For facilities regulated by the NRC or an Agreement State, the
standards endorsed the annual dose limits given in the environmental standards for the uranium
fuel cycle (40 CFR Part 190): 25 mrem to the whole body, 75 mrem to the thyroid, and 25 mrem
to any critical organ (EPA77). The 25 mrem dose limit was based on a dosimetry system dating
from the 1977 International Commission on Radiation Protection recommendations (ICR77),
which are now outdated. The ICRP dose limit has since been revised to be consistent with
current dosimetry, so that the 15 mrem/yr CEDE dose limit in the proposed 40 CFR Part 197 rule
is essentially the same as the 25 mrem/yr limit for the 1977 dosimetry.
Subpart B imposed limits associated with the release of radioactive materials into the
environment following closure of the repository. The key provisions of Subpart B (EPA85)
were:
• Limits on cumulative releases ofradioactive materials into the environment during
the 10,000 years following disposal (§191.13)
• Assurance requirements to compensate for uncertainties in achieving the desired level
of protection (§191.14)
• Individual exposure limits based on the consumption of ground water and any
other potential exposure pathways for 1,000 years after disposal (§191.15)
• Ground water protection requirements in terms of allowable radionuclide
concentrations and associated doses for 1,000 years after disposal (§191.16)
• Consideration of inadvertent human intrusion into the repository (Appendix B)
Under §191.15 and § 191.16 of Subpart B, the annual dose to any member of the general public
was limited to 25 mrem to the whole body and 75 mrem to any critical organ (under the outdated
dosimetry system). The ground water concentration for beta or gamma emitters was limited to
the equivalent yearly whole body or organ dose of 4 mrem. The allowable water concentration
for alpha emitters (including radium-226 and radium-228, but excluding radon) was
15 picocuries/liter (pCi/L). For radium-226 and radium-228 alone, the concentration limit was
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5 pCi/L. Appendix A of the standards provided acceptable radionuclide-specifie cumulative
release limits.
In March 1986, five environmental groups led by the Natural Resources Defense Council and
four States filed petitions for a review of 40 CFR Part 191 (USC87). These suits were
consolidated and argued in the U.S. Court of Appeals for the First Circuit in Boston. The main
challenges concerned:
• Violation of the Safe Drinking Water Act (SDWA) underground injection section,
• Inadequate notice and comment opportunity on the ground water protection
requirements, and
• Arbitrary standards, not supported in the record, or not adequately explained.
In July 1987, the Court rendered its opinion and noted three findings against the Agency and two
favorable judgments. The Court's action resulted in the remand of Subpart B, the disposal
standards. The Court began by looking at the definition of "underground injection." In the view
of the Court, the method envisioned by DOE for disposal of radioactive waste in underground
repositories would "...likely constitute an underground injection under the SDWA."
Under the SDWA, the Agency is required to assure that underground sources of drinking water
will not be endangered by any underground injection. With regard to such potential
endangerment, the Court supported part, but not all, of the Agency's approach. Inside the
controlled area, the Court ruled that Congress—through the EPA—had allowed endangerment of
ground water. However, the Court accepted EPA's approach of using the geological formation
as part of the containment. This aspect of the Court's opinion is important in that it recognizes
that a portion of the area around the footprint of the geologic repository could be considered to
be an integral part of the repository system and could be dedicated to that use. This area was
designated as a controlled area in the rule and was limited to an area of 100 square kilometers
(sq. km.).
Outside the controlled area, the Court found that §191.15 would allow endangerment of drinking
water supplies. In the context of the SDWA, "endangerment" was considered when doses higher
than those allowed by the Primary Drinking Water Regulations could occur. In§191.15, an
annual dose of 25 mrem to the whole body and 75 mrem to any critical organ from all pathways
is permitted, whereas existing EPA regulations promulgated under the SDWA allowed an annual
dose of 4 mrem from drinking water. Although the Court recognized that an exposure level less
than 4 mrem could result from the ground water pathway, it rejected this possibility because the
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Agency stated that radioactivity could eventually be released into the ground water system near
the repository and that substantially higher doses could result. Therefore, the Court decided that
a large fraction of the 25 mrem limit could be received through the ground water exposure
pathway. Accordingly, the Court found that the Part 191 standards should either have been
consistent with the SDWA or the Agency should have justified the adoption of a different
standard.
The Court stated that the Agency was not necessarily incorrect in promulgating the proposed
standards. However, it noted that the Agency never acknowledged the interrelationship of the
SDWA and the Part 191 standards nor did it present a reasonable explanation for the divergence
between them. The Court also supported the petitioner's argument that the Agency had not
properly explained the selection of the 1,000-year limit for individual-protection requirements
(§191.15). The Court indicated mat the 1,000-year criterion was not inherently flawed, but
rather that the administrative record and the Agency's explanations did not adequately support
this choice. The criterion was remanded for reconsideration, and the Agency was directed to
provide a more thorough explanation for its basis.
Finally, the Court found that the Agency did not provide sufficient opportunity for notice and
comment on §191.16 (Ground Water Protection Requirements), which was added to Subpart B
after the standards were proposed. This section was remanded for a second round of notice and
comment. There were, however, no rulings about §191.16 issued on technical grounds.
In August 1987, the Department of Justice petitioned the First Circuit Court to reinstate all of
40 CFR Part 191 except for §191.15 and §191.16, which were originally found defective. The
Natural Resources Defense Council filed an opposing opinion. The Court then issued an
Amended Decree that reinstated Subpart A, but continued the remand of Subpart B.
In 1992, the WIPP LWA reinstated Subpart B of 40 CFR Part 191, except §191.15 and §191.16,
and required the Administrator to issue final disposal standards no later than six months after
enactment. On December 20,1993, EPA issued amendments to 40 CFR Part 191 which
eliminated §191.16 of the original rale; altered the individual-protection requirements; and added
Subpart C on ground water protection (EPA93).
The revised Part 191 standard finalized in 1993 retained the waste containment and assurance
requirements in the original 1985 standard. However, an important change was made for the
individual-protection requirements: the protection dose limit was recalculated according to the
newer Committed Effective Dose Equivalent (CEDE) methodology. This approach gave a dose
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limit of 15 mrem/yr. This new methodology considers the weighted relative importance of organ
doses and the accumulation of dose potential over time. The original dose limit of 25 mrem/yr in
the old methodology is equivalent to the 15 mrem/yr limit in the new system.
The revised Part 191 standard finalized in 1993 also moved the guidance on the treatment of
human intrusion into a new Appendix C dealing with implementation of the rale's numerical
standards. This guidance was subsequently supplanted by recommendations from the National
Academy of Sciences in its report on the technical bases for Yucca Mountain standards (NAS95;
see discussion below). With regard to the ground water protection standards, the revised Part
191 rule retained the requirements for specific radionnclides that were in the 1985 standards, but
the compliance period was changed from 1,000 to 10,000 years to be consistent with the
individual-protection requirement.
The WIPP LWA also exempted Yucca Mountain from the generic disposal standards set forth
under 40 CFR Part 191, Subpart B. Pursuant to specific provisions in the EnPA, EPA was
charged with setting site-specific environmental radiation protection standards for Yucca
Mountain, The 40 CFR Part 197 standard is responsive to this mandate.
1.6 The National Academy of Sciences' Recommendations
In the EnPA, the Congress directed the Academy to address three issues in particular:
• Whether a health-based standard based upon doses to individual members of
the public from releases to the accessible environment -will provide a
reasonable standard for protection of the health and safety of the general
public;
* Whether it is reasonable to assume that a system for post-closure oversight of
the repository can be developed, based upon active institutional controls, that
will prevent an unreasonable risk of breaching the repository's engineered or
geologic barriers or increasing exposure of individual members of the public
to radiation beyond allowable limits; and
" Whether it will be possible to make scientifically supportable predictions of
the probability that the repository's engineered or geologic barriers will be
breached as a result of human intrusion over a period of 10,000 years
(EnP92).
The NAS recommendations in these three areas had direct bearing on the approach used by EPA
in developing its site-specific IPS, HIS, and GWS for Yucca Mountain.
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To address these questions, the Academy assembled a committee of 15 members representing a
range of scientific expertise and perspectives. The committee conducted a series of five
technical meetings at which more than 50 nationally and internationally known scientists and
engineers were invited to participate. In addition, the committee received information from the
Nuclear Regulatory Commission (NRC), the Department of Energy (DOE), EPA, Nevada State
and county agencies, and private organizations, such as the Electric Power Research Institute.
The committee's conclusions and recommendations are contained in its final report, entitled
Technical Bases for Yucca Mountain Standards, which was issued on August 1, 1995 (NAS95).
In this report, the committee offered the Agency several general recommendations as to the
approach EPA should take in developing 40 CFR Part 197. Specifically, the NAS recommended
(NAS95, p.2):
• The use of a standard that sets a limit on the risk to individuals of adverse
health effects from releases from the repository. 40 CFR Part 191 contains an
individual-dose standard, and it continues to rely on a containment
requirement that limits the releases ofradionuclides to the accessible
environment. The stated goal of the containment requirement was to limit the
number of health effects to the global population to 1,000 incremental
fatalities over 10,000 years. We do not recommend that a release limit be
adopted.
• That compliance with the standard be measured at the time of peak risk,
whenever it occurs. (Within the limits imposed by the long-term stability of the
geologic environment, which is on the order of one million years.) The
standard in 40 CFR Part 191 applies for a period of 10,000 years. Based on
performance assessment calculations provided to us, it appears that peak risks
might occur tens or hundreds of thousands of years or even farther into the
future.
• Against a risk-based calculation of the adverse effect of human intrusion into
the repository. Under 40 CFR Part 191, an assessment must be made of the
frequency and consequences of human intrusion for purposes of demonstrating
compliance with containment requirements. In contrast, we conclude that it is
not possible to assess the frequency of intrusion far into the future. We do
recommend that the consequences of an intrusion be calculated to assess the
resilience of the repository to intrusion.
The NAS committee also recommended that policy issues be resolved through a rulemaking
process that allows opportunity for wide-ranging input from all interested parties (NAS95).
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The committee also addressed each of the specific questions posed to it by the Congress in the
EnPA. With regard to the first issue, protecting human health, the NAS committee
recommended (NAS95, pp. 4-7):
• ...the use of a standard that sets a limit on the risk to individuals of adverse
health effects from releases from the repository.
• ...the critical-group approach be used in the Yucca Mountain standards.
• ...compliance assessment be conducted for the time when the greatest risk
occurs, within the limits imposed by long-term stability of the geologic
environment.
The NAS also concluded that an individual-risk standard would protect public health, given the
particular characteristics of the site, provided that policy makers and the public are prepared to
accept that very low radiation doses pose a negligibly small risk. A necessarily important
component in the development of a standard for Yucca Mountain is the means of assessing
compliance. The NAS committee concluded as follows (NAS95, p. 9):
• ...physical and geologic processes are sufficiently quantifiable and the related
uncertainties sufficiently boundable that the performance can be assessed
over time frames during which the geologic system is relatively stable or
varies in a boundable manner. The geologic record suggests that this time
frame is on the order of 106 years. The Committee further concluded that the
probabilities and consequences of modifications by climate change, seismic
activity, and volcanic eruptions at Yucca Mountain are sufficiently boundable
that these factors can be included in performance assessments that extend
over this time frame.
• ...it is not possible to predict on the basis of scientific analyses the societal
factors required for an exposure scenario. Specifying exposure scenarios
therefore requires a policy decision that is appropriately made in a
rulemakingprocess conducted by EPA.
With respect to the second and third questions posed by the Congress in Section 801 of the
EnPA, the NAS Committee concluded (NAS95, p. 11):
• ...it is not reasonable to assume that a system for post-closure oversight of the
repository can be developed, based on active institutional controls, that will
prevent an unreasonable risk of breaching the repository's engineered
barriers or increasing the exposure to individual members of the public to
radiation beyond allowable limits.
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* ...it is not possible to make scientifically supportable predictions of the
probability that a repository's engineered or geologic barriers will be
breached as a result of human intrusion over a period of 10,000 years.
1.7 Filial 40 CFR Part 197 - Public Health and Environmental Radiation Protection
Standards for Yucca Mountain, Nevada
Three key elements of the 40 CFR Part 197 standard are the individual-protection standard
(§197.20), the human-intrusion standard (§ 197.25), and the ground water protection standards
(§197.30). These are discussed below and compared with the 40 CFR 191 generic disposal
standards and the NAS recommendations. The basis for certain site-specific aspects of the
regulation are also presented.
in
In developing a site-specific standard for the Yucca Mountain site, the generic requirements
Part 191 serve as a starting point for the process. The generic requirements in Part 191 were
examined in terms of whether their components are relevant to the Yucca Mountain geologic
setting; if they are determined to be relevant, the next issue is how they can be framed
appropriately for that setting.
In contrast to the individual, human intrusion, and ground water protection standards, Part 191
also contained a containment requirement that was not carried into the Yucca Mountain standard.
The containment requirement in Part 191 was intended to address a situation where releases from
a poorly performing geologic repository could enter into large surface water bodies, such as
rivers, lakes, or the ocean, where the contamination would be greatly diluted and the dose
distributed to a potentially large population. The containment requirement was intended to limit
such situations. For the Yucca Mountain setting, this scenario is not plausible since no large
surface water bodies exist in the arid desert environment in the site vicinity. The text below
discusses how the individual, human intrusion, and ground water protection standards were
framed for the Yucca Mountain setting.
1.7.1 Individual-Protection Standard
An individual-protection standard is a relevant and fundamental regulatory requirement for any
repository setting and therefore must be incorporated into any site-specific standard.
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The individual-protection standard in Part 197 requires DOE to demonstrate:
...using performance assessment, that there is a reasonable expectation that for
10,000 years following disposal, the reasonably maximally exposed individual
receives no more than an annual committed dose equivalent of 150 microsieverts
(15 millirems) from releases from the undisturbed Yucca Mountain disposal
system. The DOE's analysis must include all potential pathways ofradionuclide
transport and exposure (EPA01).
By way of comparison, the individual-protection standard in the 40 CFRPart 191 generic
disposal standard also specifies, at §191.15, an annual committed effective dose equivalent
(CEDE) of 15 mrem. The use of an individual-protection standard rather than a release limit is
consistent with recommendations of the NAS as discussed in 1.3 above. Further, the NAS noted
that a risk range of 10~s to 10"6 per year was a reasonable starting point for EPA's rale making
(NAS95, p. 5). Thus selection of a CEDE of 15 mrem for 40 CFR Part 197, which is equivalent
to an annual risk of 7x10"6, is also consistent with the NAS recommendations.
Total release limits in the generic Part 191 regulation were developed to protect the general
population from repository releases via all pathways. The NAS concluded that protecting public
health by establishing an individual-protection exposure limit is also an adequate means of
assuring the general population is protected. For the Yucca Mountain site, the overwhehningly
dominant exposure pathway involves releases into the ground water system beneath the
repository, followed by transport of contaminants to downgradient individual receptors. An all-
pathways standard for an individual would therefore include the most important exposure
pathways.
1.7.2 Human-Intrusion Standard
Inadvertent intrusion is an unanticipated event that could have consequences ranging from minor
to highly significant depending on the geologic setting. An HIS was included in the generic
Part 191 standard because of this potential range of consequences, and to enable the
consequences to be examined for any specific repository site. For the Yucca Mountain setting,
site characterization work has shown that potable water is the only recognized potential resource
at and near the repository location. Recognizing the relatively low probability of intrusion into
the repository for resource exploration, the NAS recommended that human intrusion be
considered only as a stylized test of repository resiliency, separate and distinct from the
evaluations of expected repository performance. The NAS did not find that consideration of
human intrusion was inappropriate for the Yucca Mountain site. It made recommendations on
framing the stylized scenario which were the bases for EPA's standard.
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As discussed in Section 1.3 above, the NAS Committee on the Technical Bases for Yucca
Mountain Standards concluded that active institutional controls would not be a reliable long-
term deterrent to human intrusion into a repository. Consistent with this finding, EPA proposed
two alternative approaches for consideration as the human-intrusion standard under 40 CFR Part
197. Under Alternative 1 for proposed §197.25, DOE would be required to demonstrate that;
...there is a reasonable expectation that for 10,000 years following disposal the
reasonably maximally exposed individual receives no more than an annual
committed effective dose equivalent of 150 microsieverts (15 mrem) as a result of
human intrusion. The DOE's analysis of human intrusion must include all
potential environmental pathways ofradionuclide transport and exposure
(EPA99).
Under this alternative NRC would determine the range of time during which intrusion occurs
based on EPA guidance provided in proposed §197.26.
Under Alternative 2 the DOE would be required to determine:
...the earliest time after disposal that the waste would degrade sufficiently that a
human intrusion... could occur without recognition by the drillers (EPA99).
In the final rule, EPA selected this second alternative in which DOE must project the time at
which waste packages have degraded sufficiently that penetration of a waste package by a
drilling intrusion could occur without being noticed by the drillers. A connection between the
repository and the underlying saturated zone below the repository is established by the intruding
borehole penetration, and doses from the single breached waste package are to be projected in
the same manner as for the individual-protection standard compliance calculations. The same
dose limit is applied, as used for the individual-protection standard, but the calculation is a
separate performance scenario independently calculated and evaluated against the 15 millirem/yr
limit. If exposures occur before the end of the regulatory period, the calculations assessments
are evaluated against the 15 millirem/yr limit. If exposures occur after the regulatory period, the
assessments are included in the repository Environmental Impact Statement.
In each case a single vertical borehole is assumed to penetrate the degraded waste package and
continue down to the saturated zone. Similar to 40 CFR Part 191, intrusion is limited to
inadvertent exploratory drilling for resources. However, the frequency of intrusion is different in
the two regulations. The Appendix C Guidance to the generic disposal standards specifies that
the drilling not exceed 30 boreholes per square kilometer per 10,000 years for repositories near
sedimentary rocks and 3 boreholes per square kilometer per 10,000 years for repositories in other
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geologic formations. This Appendix C Guidance was refined for the Waste Isolation Pilot Plant
in 40 CFRPart 194 (Criteria for the Certification and Re-Certification of the Waste Isolation
Pilot Plant's Compliance with the 40 CFRPart 191 Disposal Regulations). In §194.33 drilling
frequency is based on the frequency of drilling for resources for the past 100 years within a
particular geographic area (i.e., the Delaware Basin) surrounding the WIPP Site. This
requirement is appropriate for an area where extensive drilling for a variety of resources had
occurred. Since the Yucca Mountain area has not been subject to extensive exploration drilling,
the Agency chose the approach very similar to that recommended by the NAS, namely a
"stylized intrasion scenario consisting of one borehole of a specified diameter drilled from the
surface through a canister of waste to the underlying aquifer" (NAS95, p. 111).
1.7.3 Ground Water Protection Standards
Ground water protection standards were included in the generic Part 191 standards and in the
WIPP certification effort. Inclusion of ground water protection standards in the Yucca Mountain
standard can be considered relevant for several reasons. The repository site is located in the
unsaturated zone (UZ) directly above potable water sources; any contaminant releases into the
UZ will move downward into these aquifers, which supply water to the population downgradient
of the site. Also, protection of ground waters is well-established national policy. From a purely
technical perspective, the NAS chose not to consider the question of ground water standards,
noting that an all-pathways exposure limit would include doses from ground water use.
However, it is Agency policy, as well as national policy, and the policy of most states, to protect
ground water resources.
Throughout the NAS report the text acknowledged that EPA may elect to take approaches other
than a narrow interpretation of the committee's recommendations for reasons other than
specified in the report. In this way, the broader role of the Agency in applying policy factors as
well as technical rationale was acknowledged.
The Safe Drinking Water Act (SDWA) was enacted to assure safe drinking water supplies and to
protect against endangerment of underground sources of drinking waters (USDWs). Under the
authority of the SDWA, the EPA issued interim regulations (40 CFR Part 141, Subpart B)
covering the permissible levels of radium, gross alpha, man-made beta, and photon-emitting
contaminants in community water supply systems (EPA76). Similar to hazardous chemical
substances, limits for radionuclides in drinking water are expressed as Maximum Contaminant
Levels (MCLs). The current MCL for radium-226 and radium-228 combined is 5 pCi/1, and the
MCL for gross alpha particle activity (including radium-226, but excluding radon and uranium).
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is 15 pCi/1. For man-made beta particle- and photon-emitting radionuclides (except tritium and
strontium-90), individually or in combination, the MCL is set at an annual dose limit of 4 mrem
to the total body or any internal organ. For tritium and strontium-90, the MCLs are 20,000 pCi/I
and 8 pCi/1, respectively.
In 1991, the EPA issued a Notice of Proposed Rulemaking (NPRM) under 40 CFR Parts 141 and
142 to update the 1976 interim regulations for radionuclide water pollution control (EPA91).
The NPRM, under the SDWA, proposed the establishment of Maximum Contaminant Level
Goals (MCLGs) and Maximum Contaminant Levels (MCLs). The MCLGs and MCLs target
radium-226, radium-228, natural uranium, radon, gross alpha, gross beta, and photon emitters.
As proposed, MCLGs are not enforceable health goals. In contrast, MCLs are enforceable
standards. The EPA concluded that radionuclide MCLGs should be set at zero to avert known or
anticipated adverse health effects while providing an adequate margin of safety. In setting the
MCLs, the EPA also committed itself to evaluating the feasibility, costs, and availability of
water treatment technologies, as well as other practical considerations. The 1991 proposed
rulemaking included the following MCLs: radium-226,20 pCi/1; radium-228,20 pCi/1; radon-
222,300 pCi/1; uranium, 20 micro g/1; adjusted gross alpha, 15 pCi/1; and beta and photon
emitters, 4 mrem ede/yr.
The generic disposal standards at 40 CFR Part 191 also incorporate the 40 CFR 141 Subpart B
ground water protection requirements. EPA believes that it is prudent and appropriate to impose
requirements for waste disposal that are protective of water resources for future generations,
without imposing a burden of water treatment and cleanup on those future generations.
In the Yucca Mountain standard, DOE is required under §197.30 to provide, in its license
application to NRC:
...a reasonable expectation that, for 10,000 years of undisturbed performance
after disposal, releases of radionuclides from waste in the Yucca Mountain
disposal system into the accessible environment will not cause the level of
radioactivity in the representative volume of ground water to exceed the limits' in
Table L,,(EPA01).
Table 1 limits combined Ra-226 and Ra-228 to five picocuries per liter (pCi/1) including natural
background and gross alpha activity (including Ra-226 but excluding radon and uranium) to
15 pCi/1 including natural background. Combined beta and photon emitting radionuclides are
limited to levels where the annual dose (excluding natural background) to the whole body or any
organ will not exceed 40 microsieverts (4 mrem). These limits are the same as the maximum
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contaminant levels (MCLs) established by the Agency under the Safe Drinking Water Act
(SWDA),
1.7.4 Site-Specific Regulatory Requirements
While many elements of the 40 CFR Part 197 rule are either similar to other EPA regulations
such as 40 CFR 191 and 40 CFR Part 141 or based on recommendations of the NAS, certain
elements are based on site-specific considerations. These include the definition of the
reasonably maximally exposed individual (RMEI), the location of the point of compliance, and
the representative volume of water for measuring compliance with the ground water protection
standard. Each of these site-specific elements are discussed below.
Reasonably Maximally Exposed Individual (RMEI)
For DOE to determine whether the Yucca Mountain disposal system complies with the
individual-protection standard, they must calculate the dose to an individual or group of
individuals and compare that dose with the requirements contained in §197.20 (i.e., a maximum
annual CEDE of 15 mrem). The regulation must specify those characteristics, habits, age, life-
style, etc. which describe the individual or group of individuals. For this purpose EPA has
chosen to use, as the basis for comparison with the individual-protection standard,, the dose
received by the reasonably maximally exposed individual.
The RMEI is defined in §197.21 as a hypothetical person who:
(a) lives in the accessible environment above the highest concentration ofradionuclides
in the plume of contamination;
(b) Has a diet and living style representative of the people who now reside in the Town of
Amargosa Valley, Nevada. The DOE must use projections based upon surveys of the
people residing in the Town of Amargosa Valley, Nevada, to determine their current diets
and living styles and use the mean values of these factors in the assessments conducted
for§§ 197.20 and 197.25; and
(c) Drinks two liters of water per day from wells drilled into the ground water at the
location [where the RMEI lives]. (EPA01)
The NAS recommended that the risk to the average member of the critical group be used as the
basis of comparison with the risk limit of the standard. The NAS Committee proposed two
alternatives - a probabilistic critical group approach and a subsistence farmer critical group.
After considering these possibilities, the Agency decided to use the RMEI concept which is
consistent with other EPA programs and is believed by the Agency to provide a level of
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protection substantially equivalent to that provided by the critical group concept for small
populations. The RMEI concept involves estimating high-end doses which are in excess of the
90th percentile of the range of doses for the exposed population. The goal is to calculate doses
which are not the most extreme but are well above the average for the exposed population.
EPA considered four possible scenarios to define the RMEI including (EPA99):
• A subsistence farmer residing 30 to 40 km downgradient at a location where the
water table is near the surface, who obtains all food and water from contaminated
sources
• A commercial farmer subject to the same exposure pathways as the subsistence
farmer.
• A community located near the repository site that obtains its water for domestic use
from an underground source of drinking water.
• A rural-residential RMEI exposed to the same pathways as the subsistence farmer.
However, the rural-residential RMEI does personal gardening but does not work as a
full-time farmer.
The fourth scenario was chosen as the basis for developing the specific requirements under
§197.21. This scenario is believed to be representative of most of the current residents of the
Amargosa Valley.
Representative Volume of Ground Water
In accord with Agency policy of protecting ground water resources, the Representative Volume
(RV) concept was developed in response to consideration of the actual resource to be protected
at the site. The RV is based on current land uses involving ground water, i.e , the resource to be
protected, and the fundamental assumption is that future lifestyles and water uses will be the
same as those of the present. This assumption is necessary to avoid making judgments based on
speculation. The RV is intended to be a volume of water used annually that provides the basis
for calculating radionuclide concentrations resulting from repository releases. Resulting
concentrations would be compared to MCLs established in the SDWA.
The representative volume is the volume of water needed to supply the demands of a defined
RMEI that could exist in the future at the point of compliance for the ground water protection
standards (see discussion below for details on point of compliance). To meet such demands, the
water must contain less than 10,000 milligrams of total dissolved solids per liter (i.e., potable).
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The proposed Part 197 standards included a number of possible RVs based on current land uses
south of Yucca Mountain. One proposed alternative was 1,285 aere-feet/yr. This RV is the sum
of the water requirements for alfalfa farming and domestic use. It is based on a small farming
community of 25 people with 255 acres of alfalfa under cultivation (the average current size of
these farms in the area) which is the current economic base for the Amargosa Valley. Alfalfa
farming requires about 5 acre-feet of water per acre (255 acres x 5 acre-feet per acre = 1,275
acre-feet for irrigation). The average annual water demand for a non-farming family of four with
a garden is 10 acre-feet. TMs is also the lower bound for the amount of water used through 15
connections from public water supply serving at least 25 people (as defined in the SWDA). The
representative volume is, therefore, the sum of the water requirements for alfalfa farming and
domestic use.
Another alternative RV proposed was 120 acre-feet/yr. This value corresponds to the water use
for a small municipal community of approximately 150 individuals who use the water for
domestic and municipal purposes.
For the final rule, a representative volume of 3,000 acre-ft/yr was defined. This representative
volume, as described in the preamble to the final rale (66 FR 32074-32135, June 13, 2001),
represents a composite of the water demands for downgradient users of the ground water
resource. The composite water use estimate includes current use for alfalfa cultivation (the
largest consumer of water for agricultural purposes), and projected increases for population and
commercial/industrial uses in the Lathrop Wells area northward to the boundary of the Nevada
Test Site.
Section 197.31 describes the RV and includes specific concepts concerning how the RV could be
incorporated into the radionuclide transport modeling that will be included hi analyses to support
demonstration of compliance during the licensing process.
Point of Compliance
In the proposed rale, two mechanisms were proposed for compliance determinations, specifically
to identify where ground-water contamination and individual radiation exposures are to be
projected for comparison against the limits contained in the standard. One alternative was a
controlled area concept, similar in intent to the concept as originally used in Part 191. The
controlled area denotes a bounded geographic area within which the standards would not be
applied. The standard's limits would be applied at the boundary of the controlled area, which
serves as the beginning of the defined "accessible environment." The land within the boundary
of the controlled area is considered part of the natural barrier of the disposal system, and as such
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is dedicated to the sole purpose of isolating the radioactive wastes from the accessible
environment. The second proposed alternative was the use of a compliance point, which serves
the same purpose as the border of the controlled area - it identifies the point at which ground
water contaminant concentrations and individual exposures are calculated for comparison against
the standard's limits. The point of compliance is to be located at a specific distance from the
repository and over the point at which calculated releases from the repository are projected to be
at their highest levels in the ground water beneath Ms point.
In the proposed rule EPA included four compliance measure alternatives for consideration, two
of which incorporated a controlled area and two of which incorporated a compliance point.
These alternatives include downgradient distances of 5,18,20, and 30 km. At the present time
there is no one residing 5 km downgradient**** from the repository, since it is within the
boundaries of the Nevada Test Site (NTS); there are about 10 people residing between 18 km
(the NTS boundary) and 20 km downgradient, and hundreds of persons around 30 km
downgradient Future population increases are expected at the 20- and 30-km downgradient
locations (EPA99,01). In addition, the depth to ground water decreases from about 300 meters
near the repository location to about 50 to!5 meters within that portion of the Town of Amargosa
Valley where most of the population resides and commercial agriculture is the basis for the local
economy.
In the final rale, the Agency has incorporated a controlled area concept as a compliance
mechanism, as defined in Section 197.12 of the final rule. The controlled area concept comports
more directly with the direction of me EnPA, which explicitly mentions the "accessible
environment" and refers to its definition from Part 191 which incorporates the controlled area
concept. The controlled area concept also more clearly delineates the extent of the natural
barrier around the repository than the simpler point of compliance approach. Neither the point of
compliance, nor the controlled area approach imposes any significant cost impacts on the
repository development program, because the site characterization efforts to define the
magnitude and direction of potential releases are the same for either approach.
"** This is the same compliance point as specified in 40 CFR Part 191, the generic disposal standard.
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2.0 OVERVIEW OF RADIOACTIVE WASTE DISPOSAL AT YUCCA MOUNTAIN
This chapter briefly describes the Yucca Mountain site and the wastes that would be stored
and disposed there if the site is approved for disposal, A summary of current estimates of
repository program costs, which total approximately $57,6 billion, is included.
2.1 Yucca Mountain as a Disposal Site
The Nuclear Waste Policy Amendments Act of 1987 (which amended the Nuclear Waste Policy
Act (NWPA) of 1982) designated the Yucca Mountain site in Nevada as the only location to be
evaluated as a possible place for disposal of spent nuclear fuel and high-level radioactive wastes.
The Yucca Mountain site is located about 90 miles north of Las Vegas, Nevada, and is situated
on the boundary of the Nevada Test Site. The climate is semi-arid, and the location was
originally selected as a candidate location for disposal because it was expected that there would
be limited potential for water to enter the repository and then to transport radionuclides to distant
locations.
2.2 Sources and Characteristics of Radioactive Wastes to Be Disposed
A repository at Yucca Mountain would dispose of spent fuel from nuclear power reactors and
high-level wastes from reprocessing of spent fuel. The sources of spent fuel would be
commercial nuclear power reactors, naval reactors, and reactors used in DOE and research
programs. High-level wastes are the result of defense operations in the states of Washington,
Idaho, and South Carolina where fuel from production reactors was processed to obtain the
uranium and plutonium used in nuclear weapons. They will consist of solidified fission product
waste separated from the recovered uranium and plutonium.
The NWPA limited the amount of wastes to be disposed at Yucca Mountain to 70,000 metric
tons equivalent of uranium (MTU). The DOE has interpreted this to correspond to disposal of
63,000 metric tonnes of spent fuel and the equivalent of 7,000 MTU of high-level wastes. The
70,000 MTU limit remains in force today, but is subject to change by future Congressional
action.
The wastes would come to Yucca Mountain for disposal from commercial nuclear power sites
and DOE operations sites throughout the country, as shown in Figure 2-1. At present, the spent
fuel from commercial power reactors is primarily stored at the sites where the fuel was used in the
reactors. The amount currently in storage totals about 40,000 MTU. Such spent fuel continues to
be discharged from the commercial reactors at a total annual rate of about 2,200 MTU. If all
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• Commercial Nuclear Reactors
X DOE Sites
* Non-Power Reactor Sites (Approx. 40)
fi> Commercial High Level Radioactive Waste Storage
• Shutdown Reactors with Spent Fuel
*• Yucca Mountain Repository
Figure 2-1, Sources of Radioactive Wastes for the Yucca Mountain Repository
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reactors operate to the end of their current licenses, the total amount of spent nuclear fuel
discharged will be about 87,000 MTU.
The DOE spent fuel, which comes primarily from research and test reactors, and spent fuel from
naval nuclear reactors, is presently stored at various DOE sites. The current total amount of this
spent fuel is less than 3,000 MTU, and the amount will not increase significantly.
High-level wastes were generated by defense production operations at DOE's Savannah River,
Idaho, and Hanford, Washington sites. In the as-generated form, these wastes are liquid and have
a total amount of tens of millions of gallons. The wastes will be solidified, and the amount sent to
Yucca Mountain, in terms of number of cans of waste to be disposed, will depend on the
solidification process used. The draft Environmental Impact Statement for the proposed
repository at Yucca Mountain, issued by DOE in August 1999, estimated that the 7,000 MTU of
HLW would be contained in about 14,000 waste canisters (DOE99).
2.3 Overview of the Repository for Disposal
The basic concepts for disposal of highly radioactive wastes into geological formations were set
forth by the National Academy of Sciences in the 1950's and have been embodied in repository
design concepts and regulatory concepts ever since then. The wastes are to be emplaced in deep
geological formations which isolate them from the human environment, and a system of
engineered and natural barriers is to be used hi combination to maintain the wastes in isolation
and to prevent release of radionuclides. The Yucca Mountain site, and other sites that had been
under consideration, would use a combination of engineered and natural barriers appropriate to
the site to maintain the wastes in isolation and to demonstrate compliance with regulatory
standards for radionulides that were released from the repository.
At Yucca Mountain, the repository would be excavated hi the unsaturated zone, i.e., in a geologic
formation in which the pores and fractures hi the geologic medium are not filled with water. The
Yucca Mountain site, in comparison with other candidate sites, was unique in having capability
for this type of emplacement. It was expected that the lack of ability for water to reach the wastes
and transport them to the environment would dominate the safety performance of the repository
and enable easy demonstration of compliance with regulatory standards.
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2.4 DOE Estimate of the Repository Program Cost
DOE documented estimated repository program costs in the Viability Assessment (VA)
documents (DOE98). The principal cost elements were identified as follows:
Historical costs - . $ 5.9B
Costs to complete work to the License Application- LIB
Respository costs from licensing to closure - 18.7B
Total for the repository program - S25.7B
Estimates of costs for design options (options to the VA design) were provided in Volume 5 of
the VA document One of the options considered was use of drip shields and backfill, as is now
planned for me current design, EDA II (see Section 3.6). The estimated cost of this option was
$0.8 billion. However, this estimate did not consider the long-term total cost of these
modifications.
DOE has released an updated "Total System Life Cycle Cost" (TSLCC) estimate (DOEOla),
which gives a total cost for the repository of $57.6 billion, which includes historic costs. This
higher cost includes cost elements not included in the VA estimate, and is a more accurate
estimate of total program costs.
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3.0 EVOLUTION OF THE YUCCA MOUNTAIN REPOSITORY DESIGN
This chapter describes the evolution of design concepts for a repository at Yucca Mountain that
has occurred as a result of site characterization findings, performance assessment results,
external reviews, and strategy for dealing with uncertainties. The discussion demonstrates that
EPA's standards have not affected the design evolution.
This section describes how the design of a repository for the Yucca Mountain site has evolved
since the Site Characterization Plan (SCP; DOE88) was published in 1988. The SCP reference
design concept involved vertical emplacement of small, thin-walled canisters, with a design
lifetime on the order of 300-1,000 years, into the floor of tunnels excavated in Yucca Mountain.
The current design concept calls for horizontal emplacement of large, double-walled waste
packages, with a design lifetime of more than 100,000 years (TRWOO), into drifts excavated in
Yucca Mountain with a tunnel boring machine.
The design evolution has been driven principally by acquisition of site characterization data
which showed that the performance of the natural features of the repository system during the
regulatory period would be less effective than anticipated when the SCP was issued and data were
sparse. It was originally expected that water would flow very slowly, and in limited amounts,
through the unsaturated geohydrologic regime, that radionuelides released from the repository
and transported by water would be trapped on rock surfaces and pores along the flowpatii, and
that water would travel relatively slowly through the saturated zone. In contrast to this
expectation, site characterization data have demonstrated that water from precipitation infiltrates
into the mountain at rates much higher than originally expected; mat there are paths for rapid
transport of water from the surface to the repository horizon and possibly to greater depths; and
that flow in the saturated regime is expected to occur primarily in fractures and with limited
dilution of radionuclide concentrations. Potential for radiation doses during the regulatory period
is dominated by soluble radionuelides that are mobile and move with the water. The natural
features will constrain transport of radionuelides that are insoluble and sorbed onto rock surfaces.
The design evolution also was guided by results of a series of analyses of expected repository
performance known as Total System Performance Assessments (TSPA); by DOE/NRC technical
exchanges and NRC documents which indicate NRC expectations for licensing reviews; and by
external reviews of program documents and status by parties such as the Nuclear Waste Technical
Review Board (NWTRB), the NRC staff, and the TSPA Peer Review Panel. A series of formal
Expert Elicitations on key performance topics such as waste package degradation also played a
significant role in design evolution.
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Several stages of design evolution can be identified and associated with the SCP and a subsequent
series of TSPA reports. The SCP in 1988 was followed by a series of TSPA evaluations in 1991,
1993,1995,1998, and 2000. These evaluations were aimed at providing guidance for site
characterization activities and priorities and at exploring the effects of engineered design options
on performance. In the 1996-1997 time frame, site characterization date and results of expert
elicitations became available and provided the basis for the TSPA evaluations included in the
Yucca mountain. Viability Assessment (i.e,, the TSPA-VA), which was issued in 1998 in response
to a mandate by the U.S. Congress. The TSPA-VA was the first performance evaluation for a
potential repository design at Yucca Mountain. This assessment has been replaced by the TSPA
for Site Recommendation (TSPA-SR), which focuses on the latest repository design. This design
was developed as a consequence of findings of the TSPA-VA, as described here.
External and DOE-intemal reviews of the TSPA-VA revealed that there were highly significant
uncertainties and technical issues associated with the repository design that were the basis for the
TSPA-VA. In response to the critiques and suggestions, DOE subsequently developed and
adopted the Enhanced Design Alternative (EDA) concept, in which several improved repository
designs were evaluated. The selected alternative, known as EDA II, subsequently became the
design basis for the most recent TSPA iteration, known as the TSPA for Site Recommendation
(TSPA-SR).
Discussion of the design and associated TSPA evolution process is provided below. The current
design concept, EDA n, is described in Section 3.4. Discussion of TSPA methodology and
results is provided in Section 4. The discussion here shows how the repository design was shaped
by the evolving understanding of the site's natural features and the uncertainties involved in
projecting repository performance.
3.1 The 1988 Site Characterization Plan
The Nuclear Waste Policy Act of 1982 (NWPA) required each candidate repository site to
prepare a comprehensive site characterization plan describing how information would be obtained
to determine the site's suitability for disposal of highly radioactive wastes. After enactment of
the Nuclear Waste Policy Amendments Act of 1987, which designated Yucca Mountain as the
only candidate site to move forward with evaluation of suitability for disposal, DOE issued the
SCP for the site in 1988. The document received comprehensive, in-depth review by NRC staff,
whose comments, based on the Commission's 10 CFR Part 60 regulations for high-level waste
disposal, helped shape the path of site characterization and design.
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At the time of publication of the SCP, the site characterization database was highly limited.
Expectations of repository performance were based largely on assumptions concerning site
features and characteristics. The plans for site characterization activities were designed to obtain
data sufficient to assess compliance with existing regulatory standards in the 40 CFR Part 191 and
10 CFR Part 60 regulations. Repository development was subsequently driven by NRC
requirements.
3.1.1 Regulatory Framework for the SCP
Under provisions of the NWPA (NWP83), the EPA is to promulgate, for high-level radioactive
waste disposal, generally applicable environmental standards for protection of the environment
and human health. The NRC is to promulgate regulations to implement the EPA standards and to
review the License Application from DOE in order to evaluate compliance with the standards.
The EPA regulations were promulgated in 1985 and codified at 40 CFR Part 191; the
implementing NRC regulations were codified at 10 CFR Part 60. When the SCP was published in
1988, Part B of the EPA regulations had been remanded by a Federal District court to the Agency
for reconsideration. Part B specifies limits on cumulative, long-term radioactivity release from a
repository, and also characterizes use of performance assessment to evaluate releases. Although
Part B of the 40 CFR Part 191 regulations was being reconsidered by the Agency at the time the
SCP was issued, DOE treated the Part B requirements as an operative part of the regulatory
framework. Implementation was guided by the Issues Hierarchy (DOE86), which had at the top
of the hierarchy, as the overarching issues, the NRC's 10 CFR Part 60 subsystem performance
requirements.
The NRC's implementing 10 CFR Part 60 regulations, in addition to adopting the EPA
requirements, set performance objectives for specific parts of the repository system. These
subsystem performance requirements included:
• Containment of waste within the waste packages must be "substantially complete" for
a period of 300 to 1,000 years.
* The rate of radionuclide release (with certain exceptions) from the Engineered Barrier
System (EBS) following the containment period must not exceed one part in 100,000
per year of the inventory at 1,000 years following repository closure.
* The pre-waste-emplacement ground water travel time along "the fastest path of likely
radionuclide travel" from the disturbed zone to the accessible environment must be at
least 1,000 years. The boundary of the accessible environment was defined by the
EPA regulations to be located five km from the boundary of the repository and
covering no more than 100 km2 in area.
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These subsystem performance requirements drove the repository system design, e.g., selection of
a waste canister design with an expected lifetime of 300-1,000 years. As previously noted, the
natural features of the repository system (low and slow water flow; radionuclide holdup) were
expected to be the dominant contributors to safety performance.
3.1.2 Principal SCP Repository Design and Natural System Features
The SCP repository design was based on emplacement of 70,000 MTHM of spent fuel and high-
level waste in an array of vertical boreholes drilled into the floor of drifts in the Topopah Spring
Member of the Paintbrush Tuff Formation. (The 70,000 MTHM limit was set in the NWPA and
remains unchanged.) The areal power density for the repository was set at 57 kilowatts per acre,
, and the reference design was based on emplacement of 10-year-old spent fuel.
The SCP repository layout is shown in Figure 3-1 (DOE88a). Three main drifts traverse the
length of the repository and the emplacement panels are accessed by side drifts from the mains.
Entrance into the repository is through ramps located at the North end.
As previously noted, the site characterization database was quite sparse when the SCP was issued.
It was expected that the water that could infiltrate the mountain and cause corrosion, waste form
dissolution, and radionuclide release was "...limited to very small amounts" (DOE88). Based on
annual precipitation of 15 centimeters, only about 0.1-0.5 millimeters/year were expected to
percolate from the surface to the deep rock units where the repository would be located. Travel
times to the boundary of the accessible environment were expected to be on the order of tens of
thousands of years because flow through the unsaturated zone was expected to occur in the rock
matrix.
Characterization of Yucca Mountain for the repository project began in 1978. It involved
extensive drilling of boreholes and measurement of hydrologic properties such as hydraulic
conductivity and transmissivity. Because of the complexity of the geohydrologic regime, the
database at the time the SCP was issued was still characterized as "...scanty and incomplete." The
basic model for the unsaturated zone was one of flow dominated by the partially saturated matrix.
The saturated zone model was based on Darcian flow and a dual-porosity (fractures and matrix)
concept.
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Perimeter drift
Panel-access drifts
drifts
Waste main
Tuff main
Service main
Panel number
0 300 600 900 m
| 1 1 1
Emplacement drifts
Figure 3-1. Layout of the Site Characterization Plan Repository
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The available models and data were used to estimate hydrologic parameters important to
repository performance. The average annual precipitation was estimated to be about 150 mm/yr.
Because of the thickness and heterogeneity of the unsaturated zone above the repository horizon,
temporal and spatial variations of infiltration were not expected to be the same at depth as at the
surface.
Various estimates of the infiltration rate were made; all of them showed low rates. One estimate
found that the infiltration rate at the repository horizon would be no more than 0.2 mm/yr, and the
surface rate would be no more than 0.5 mm/yr. Another study estimated that the net infiltration
rate would range from about 0.5 to no more than 4.5 mm/yr. Yet another study estimated the
range at 0.015 to no more than 4.5 mm/yr. Modeling studies after the SCP was published
generally used infiltration rates of 1.0 mm/yr or less. As discussed below, these types of values
prevailed as a basis for unsaturated zone performance until the 1996-1997 time frame.
Because of the 10 CFR Part 60 subsystem performance requirements, estimates were made of
ground water velocities and travel times. The SCP quotes findings by Sinnock et al. that the
unsaturated zone travel time, for an infiltration rate of 0.5 mm/yr, would be a minimum of 9,345
years, a mean of 43,265 years, and a maximum of 80,095 years. If the infiltration rate was
doubled to 1 mm/yr, the minimum travel time was decreased to 3,700 years, "...still greater than
the amount of time required to satisfy the [regulations]." It was stated that "...the modeling effort
has attempted to use the best available data, and it is believed the results obtained are realistic."
As indicated by this statement, at the time the SCP was developed (and for a considerable period
of time thereafter) the travel time through the UZ was believed to be sufficient to meet the
10,000-year requirement in the EPA standard.
Estimates of travel time in the saturated zone, which were based on Darcian flow and travel paths
parallel to the hydraulic gradient and nearly horizontal, showed travel times of 30 years in the
3-km path in tuffacious beds of the Calico Hills Formation and 140 years in the 2-km path for the
Topopah Springs Member, for a total of 170 years to the 5-km boundary of the accessible
environment. It was noted that other factors such as dispersion, the existence of faults or
impermeable zones, or vertical movement of water could affect the saturated zone travel times. It
was also noted that "...at this time it is uncertain whether some or all of this mechanisms exist
along the travel path." However, page 3-220 of the SCP states that more realistic date give an SZ
travel time to the 5-km accessible environment boundary of 1,700 years (SCP88). In contrast,
recent SZ travel time estimates presented to the NWTRB (EDD01) estimated travel times to a
distance of 20 km downgradient to be between 640 years (median parameter values) to 900 years
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(mean parameter values). A "refined conceptual approach," equivalent to the SCP estimate using
more realistic data at that time, gave a travel time of 1300 years to the 20 km distance.
The SCP concluded that "...based on an upper-bound flux of 0.5 mm/yr, ground water travel time
within the unsaturated zone from the proposed repository to the water table is estimated to range
from about 9,000 to 80,000 yr," and "... the minimum ground water travel time from the edge of
the repository to the accessible environment [5 km] under present conditions is approximately
9,200 years, well in excess of the 1,000 year limits required by 10 CFR Part 60.113(a)(2)."
With these expectations of high performance for the natural features of the repository system, the
engineered barrier system could be the minimum required to meet regulatory requirements, as
discussed below.
3.1.3 The SCP Engineered Barrier System
In accord with NRC's subsystem performance requirements, the waste package for the SCP
design consisted of Type 304L stainless steel containers 4.76 m long and 0.66 m in diameter, with
a wall thickness of 0.95 cm. Most of the commercial spent fuel was expected to be consolidated,
but disposal of intact assemblies was planned for fuel assemblies with damaged rods. The HLW
containers were similar to those for spent fuel but shorter.
The containers were,to be backfilled with argon and welded shut. Fully loaded waste packages
would weigh 2.7 to 6.4 metric tons, would have a power output of about 3.3 kW at the time of
emplacement, and would have a surface gamma dose rate of about 50,000 rads per hour.
The waste packages were to be emplaced in 76-cm diameter holes bored into the floor of drifts in
the underground workings. The boreholes were to be metal-lined and had a metal support plate at
the bottom on which the waste package rested. A metal plug would be placed on the top of the
emplaced package, the upper portion of the borehole would be filled with crushed tuff, and a
metal cover would be placed on the floor of the drift. Eventually, the drifts would be backfilled
with crushed tuff.
An important concept included in the SCP design was use of heat emitted by the waste packages
to drive water in the rocks away from the emplacement cavities, thereby effectively drying out the
repository host rock. The concept was seen to make a good repository setting (the unsaturated
zone in a semi-arid environment), even better by delaying the eventual contact of water with the
waste containers. The technical difficulties in characterizing performance under high thermal .
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load conditions were recognized in the SCP and was preserved as a significant technical issue in
commentary, in 1999, on the Total System Performance Assessment for the Viability Assessment
from external parties such as the TSPA Peer Review Panel (PRP99). This uncertainty played a
significant role in DOE's decision to adopt the highly engineered EDA II repository design
(described in Section 3.4 of this document).
The engineered barrier system (EBS) design, including the waste package design, was intended to
comply with the subsystem performance requirements of 10 CFR Part 60, including ability for
retrieval after 50 years. The package was intended to provide substantially complete containment
of waste for a period of not less than 300 years, but no more than 1,000 years would be required.
Thereafter the package was to limit the rate of radionuclide release from the EBS as required by
the NRC subsystem performance objectives. With the anticipated high performance of the natural
system barriers, the relatively modest performance expectation for the engineered barrier system
was expected to be sufficient to meet the assumed (from 40 CFR Part 191) standard for
cumulative releases.
The evolution of performance assessments, and the associated changing repository design, are
described in the following sections, along with the progressively improved understanding of the
natural barrier characteristics.
3.2 Design Options in the Total System Performance Assessments of 1991,1993, and
1995
As previously noted, the TSPA evaluations reported in 1991,1993, and 1995 were intended to
guide site characterization activities and priorities, and to explore the effect of design alternatives
on repository system performance. DOE carefully noted that none of the design concepts was
intended to represent an actual repository design, and none of the results were intended to be a test
of compliance with regulatory standards. However, to have a basis for assessing study results,
outputs of the evaluations were compared to the total system performance standards in Subpart B of
EPA's 40 CFR Part 191 regulations that had been adopted by NRC's 10 CFR Part 60 regulations.
Throughout this period, results of the site characterization work and other data acquisition
programs were, as they became available, incorporated into the studies and used to improve the
performance assessment models. Because the EPA Part 191 regulations set limits on radionuclide
releases to the accessible environment boundary at 5 km, the site characterization work was
focused on and near the repository footprint The surface-based data acquisition program
included activities such as drilling numerous boreholes, geologic mapping of trenches,
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characterization of surface expression of faults, and daily acquisition of meteorological data.
Excavation of the Exploratory Studies Facility (ESF), primarily during 1995 and 1996, enabled
data acquisition activities at the repository horizon to proceed in accord with excavation progress
and in parallel with the surface-based studies.
Highlights of the 1991, 1993, and 1995 TSPA analyses are presented below with focus on design
options considered. As can be seen, the options considered ranged from the simple waste
canisters hi the SCP reference design to precursors of the VA design and the current design, EDA
II. During the time period through 1995, clear evidence of limitations on the performance of the
natural features of the repository was not yet available; the shift of emphasis to large, highly-
robust packages was driven by logistics considerations (far fewer packages to handle), the
decision to excavate the repository using a tunnel boring machine, and growing indications that
very conservative assumptions arid analyses would be expected by the licensing authority during
licensing reviews.
3.2.1 TSPA-1991
The TSPA-1991 studies were the initial attempt to demonstrate TSPA concepts and methodology.
The design concept for TSPA-1991 was that of the SCP: PWR fuel with an average burnup of
33,000 MWd/MTHM and BWR fuel with an average bumup of 27,500 MWd/MTHM would be
consolidated into vertically emplaced stainless steel waste packages. The waste package
performance evaluations were based on several assumptions not supported by detailed modeling
studies. The waste package was expected to be initially dry due to heating produced by
radioactive decay; this dry period would last from 300 to 1,300 years. After wetting, the
container was expected to have a lifetime range of 9,500 years "to reflect the great uncertainty in
container performance" (BER92). A total of 33,300 containers was included hi the repository
design.
3.2.2 TSPA-1993
Two separate but parallel performance assessments were conducted in 1993 - one by the DOE
M&O Contractor (DOE94) and one by Sandia National Laboratories (WIL94). These parallel
assessments are designated as the "M&O Approach" and the "SNL Approach" in the following
discussion. The BBS designs used in these assessments resemble the design used in the TSPA-
VA and the newer EDA II design, and represent the first attempt to examine designs that were
developed to reflect anticipated repository conditions at Yucca Mountain.
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3.2.2.1 M&O Version of TSPA-93
The M&O's TSPA-93 studies considered three areal power loadings - 28.5, 57 and 114 kilowatt
per acre. Waste packages using a thick, outer corrosion allowance material (CAM) and a thinner,
corrosion resistant material (CRM) as the inner package wall were horizontally emplaced in drifts
in the Topopah Spring Member of the Paintbrush Formation. The commercial reactor spent feel
loading was 63,000 MTHM contained in thirty-year old feel with an average bumup of 36,437
MWd/MTHM (DOE94, p. 2-3). In addition, 7,000 MTHM in HLW from the defense programs
was included. The commercial spent feel was contained in 6,468 waste packages and the defense
HLW was contained in 3,829 waste packages (DOE95, p. 8-15). (Note that this design concept
reduced the number of waste packages required for commercial spent feel by about a factor of 5
in comparison with the SCP design.)
The waste packages were comprised of an outer, mild steel corrosion allowance material and an
inner, nickel-base corrosion resistant material, Alloy 825. Three thicknesses were considered for
the outer layer: 10,20, and 45 cm. The inner layer was either 0.95 or 3.5 cm thick. The
packages were assumed to be placed horizontally on crushed tuff on the floor of the drifts.
The M&O TSPA-93 assumed an ambient percolation flux with an exponential distribution and an
expected value of 0.5 mm/y. Two-thirds of the flux values were less than the expected value and
one-third were greater. These low flux values reflected SCP expectations; results of site
characterization studies had not yet had an impact.
Radionuclide sorption and decay were included in modeling of the unsaturated zone (UZ) but
diffusion was not. Six layers were used to represent stratigraphy in the UZ below the repository.
Nine vertical columns were modeled to represent UZ variability in thickness and stratigraphy over
the repository area. Temperature profiles, Darcy fluxes, and liquid saturations, were developed for
each stratigraphic layer for each thermal load as function of time. These determined dry out extent
and duration in the near field. No far-field thermal perturbation was assumed.
Climate change was incorporated by assuming that the infiltration rate would vary from 1 to
5 times the base value with an average value of 2.5, Transition to a fell glacial climate would
occur linearly over 100,000 years then return to baseline over the next 100,000 years. This cycle
was repeated over the one million year simulation time frame.
Retardation factors, developed for each nuclide for each stratigraphic unit, were similar to those
used in TSPA-1991. Sorption and decay were included in saturated zone (SZ) modeling but not
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diffusion. The SZ flux was assumed to have average value of 2 m/yr with a wide range from
4.7 x 10"6 m/yr to 390 m/yr. Only the longitudinal component of dispersion was considered in
modeling of SZ radionuclide transport. A single porosity medium was assumed for the SZ.
3.2.2.2 SNL Version of TSPA-93
The SNL TSPA-93 studies considered both vertical (in borehole) and horizontal (in-drift)
emplacement of waste packages and areal thermal loadings of 57 and 114 kilowatt per acre.
Alternative waste package designs were also considered. Details are presented in Table!3-l
(WIL94).
Table 3-1. Repository Designs Evaluated by SNL in TSPA-1993
V- '^r;8'^:fc;3'v:
- iEmplacenienJ'>i.
->:r.5i-!iMtede'v:;.::^.»
Vertical
In-borehole
Vertical
In-borehole
Horizontal
In-drift
Horizontal
In-drift
;";(tefow^|
-peE,aere)>S
57
114
57
114
'i-X:^kS^&it"&
:lf Container Description V
Thin-wall, corrosion
resistant high-Ni alloy
Thin-wall, corrosion
resistant high-Ni alloy
Mild-steel CAM over
thin-wall Mgh-Ni CRM
Mild-steel CAM over
thin-wall high-Ni CEM
SttiS'lP
*=o ••-'''.-• VTfc/fTlft/-* -- ~- -'" ''••-
*• t —'---'••: '"v^* Jt-fc"/v ; -'-«';y =s'.
'!"-?i -vCOffltalil&|":v?gS
2
2
8
8
jO"rt*l*Q|T; -^
~JQlC-:
5.6
2.8
23.2
11.6
* 2.33 km2 (577 acres) for spent fuel and 0.81 km2 (200 acres) for HLW.
The waste package for vertical, in-borehole emplacement was a thin-wall cylinder of a high-nickel
alloy such as Alloy 825. The waste package had a outside diameter of 0.71 m, a wall thickness of
0.95 cm and a length of 4.76 m. The package could handle about 2 metric tons of spent fuel (e.g.
3 PWR and 4 BWR fuel assemblies) and weighed about 5 metric tons when loaded. The waste
package for horizontal, in-drift emplacement was substantially larger with the ability to contain 21
PWR or 40 BWR fuel assemblies. The waste package was comprised of an Alloy 825 inner
barrier 0.95 cm thick surrounded by an outer barrier of mild steel 10 cm thick. The two barriers
were separated by a 0.6 cm gap. This waste package was 4.91 m long, had an outside diameter of
1.75 m and weighed more than 50 metric tons when loaded with spent fuel. This multiwall
container was too massive to permit it to be tilted and moved for vertical emplacement and
retrieval. Additional details on the two types of waste packages are summarized in Table 3-2.
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Table 3-2. Spent Fuel Waste Package Inventory for TSPA-1993
Reactor
Type
Amount of
Waste
(MTU)
Percentage
of Total
Spent liiel
Weighted
Average Age
(Years)
Weighted
Average '
Burnup
(MWd/MTD)
Hybrid
Waste
Packages
Single Type
Waste
Packages
Borehole Emplacement*
BWR
PWR
Totals
22,248
40,749
62,996
35.3
64.7
100
26,3
25,5
-
31,550
40,461
. -
28,057
1,215
2,750
32,022
In-Drift Emplacement
BWR
PWR
Totals
22,183
40,646
62,829
35.3
64.7
100
26.4
25.5
- '
31,533
40,433
-
—
—
-
3,109
4,531
7,640
* For vertical borehole emplacement, an additional 13,957 canisters would be required for vitrified HLW.
3.2.3 TSPA-1995
At the time TSPA-1995 was prepared, the regulatory framework was still in a state of flux. The
National Academy of Sciences' Committee on Technical Bases for Yucca Mountain Standards
issued its report in August 1995 (NAS95), but EPA had not promulgated the environmental
regulations specific to Yucca Mountain. Given this situation, DOE chose in TSPA-95 to evaluate
cumulative releases of radioactivity to the accessible environment based on cumulative normalized
release limits included in Table 1 of 40 CFR Part 191 and maximum doses to individuals using
ground water from a well in the tuff aquifer at the boundary of the accessible environment. In
each case, the boundary of the accessible environment was assumed to be 5 km down the saturated
zone hydraulic gradient from the edge of the repository (DOE95). Evaluations were also made
against subsystem requirements in 10 CFR Part 60.
Repository design concepts investigated in TSPA-95 were based on 63,000 MTU of spent nuclear
fuel and 7,000 MTU of defense HLW emplaced in horizontal waste packages (the same as TSPA-
93). Two areal mass loading were considered- 25 MTU/acre and 83 MTU/acre. Both backfill
and no-backfill options were analyzed as repository closure strategies. The use of backfill was
expected to act as a capillary barrier to water and as a thermal management tool. Its use would
increase waste package temperatures; evaluations of the temperature impacts of the backfill were
included in the studies.
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Commercial spent fuel was assumed to be 30 years old with a weighted average bumup of 36,666
Mwd/MTU. The same number of waste packages were assumed as in the TSPA-93 analyses
performed by the M&O contractor (DOE95, p. 8-15).
"Low"(ca. 0.02 mm/y) and "high" (ca. 1.2 mm/y) infiltration rates were considered. These rates
are in the range expected under the SCP; results of site characterization studies which showed that
infiltration rates are actually in the range of 1-10 mm/yr, and currently average about 8 mm/yr,
were not yet available for TSPA-95.
The waste package design concept for TSPA-95 was similar to that considered in TSPA-93; i.e., it
consisted of a outer mild steel corrosion-allowance material (CAM) over an inner corrosion-
resistant material (CAM) of Alloy 825. The waste container for either 21 PWR assemblies or 44
BWR assemblies was about 5.7 m long and about 1.8 m in diameter. The CAM thickness was 100
mm while the CRM thickness was 20 mm. A 21 PWR waste package would weigh about 66 tons
and produce an average of 10 kW of heat at the time of emplacement. The waste package was
assumed to rest on a gravel invert covering the bottom of a circular cross-section drift with a
diameter of 5m,
In summary, the TSPA exercises and reports of 1991,1993, and 1995 served several important
purposes in the evolution of the Yucca Mountain repository design. In brief, TSPA-91 provided
a baseline by introducing the TSPA concept and applying it to the SCP design. The subsequent
TSPA-93 and TSPA-95 efforts explored the potential ranges of contributions of engineered and
natural barriers to repository system performance. Key factors considered included the
following:
• In the 1993-1995 tune frame, DOE knew, as a result of enactment of the Energy Policy
Act of 1992, that revised dose standards and requirements for demonstration of
compliance would be forthcoming, so alternative dose standards and receptor locations
were considered. Consequently, BBS designs more reflective of changing site
characterization information were beginning to be assessed.
• As stated in TSPA-95, the SCP conceptual engineered design "... has been revised to
take into account the possibility of alternative areal mass loads, as well as the decision to
use a tunnel boring machine for the excavation of the emplacement drifts." In addition,
the large multi-purpose canister design was adopted. These design considerations led to
investigation of the performance characteristics of large, horizontally emplaced waste
packages with alternative design details, such as the type and thickness of wall
materials.
* Site characterization data were being incorporated into the TSPA-95 models and
information base as they became available, but it was becoming increasingly apparent
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that there was a high degree of inherent variability in natural system parameters, that
performance of the natural barriers might not meet expectations expressed in. the SCP,
and that performance of the natural barriers might be difficult to demonstrate with
confidence in licensing reviews.
» As a result of a limited database (limited in part by the fact that the high variability of
natural features would require an extensive database for reliable characterization),
potential bounds of the performance of the natural features were explored, using models
not well founded. For example, TSPA-95 recognized that the principal contribution of
the saturated zone to performance would be dilution, and the TSPA-95 developed and
used models which predicted overall SZ dilution factors, for an infiltration rate of 1.25
mm/yr, of 4,500 at 5 km and 31,000 at 30 km. Subsequent expert elicitations confined
the expected SZ dilution factor range to 1 - 100.
Collectively, these exploratory studies and their results laid the foundation for the Viability
Assessment reference design and the TSPA-VA performance evaluations discussed below.
33 Design Features for the Viability Assessment - 1998
The Energy and Water Development Appropriations Act of 1997 specified that DOE prepare a
viability assessment of the Yucca Mountain repository, thereby providing a status report on the
project and identifying critical issues that must be addressed before the Secretary of Energy can
make a recommendation concerning suitability of the Yucca Mountain site for disposal. The
Viability Assessment report, which included a Total System Performance Assessment - Viability
Assessment (TSPA-VA), was published in December 1998 (DOE98). Although the EPA
standards had not been developed, DOE based its analyses on annual radiation doses to the
individual members of the general public. DOE assumed a radiation dose limit of 25 mrem/yr.
Releases from the ground water to the biosphere were evaluated at a point 20 km downgradient
from the repository. Multiple exposure pathways were included in calculating doses to humans.
Time histories to one million years were considered.
As previously noted, DOE considers that the TSPA-VA evaluations are the first that address a
potential repository at the site. The major features of the repository design were similar to those in
TSPA-95. However, hi response to recommendations from the expert elicitation on waste package
degradation, the waste package inner wall was Alloy 22 to provide enhanced corrosion resistance.
The drifts were assumed to be concrete lined. Backfill was not included in the reference design
but was examined as a design option. Use of ceramic coatings and drip shields were also briefly
investigated as options.
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The areal mass loading in the reference design was 85 MTU per acre with an initial heat output of
about 100 kilowatt per acre. This is based only on 63,000 MTU of commercial spent fuel which
will be emplaced in about 7,650 waste packages (DOE98, p. 3-30). According to the Draft
Environmental Impact Statement (DEIS) (DOE99), the 7,000 MTU of DOE spent fuel and HLW
waste also to be emplaced in the repository will require a total of about 22,000 waste packages.
UZ flow modeling for the TSPA-VA included climate, infiltration, mountain-scale flow and
seepage into emplacement drifts. Climates modeled included the present day dry climate with an
average annual rainfall of 170 mm/y, a long-term average climate with a rainfall of 300 mm/y and
a superpluvial climate with an average rainfall of 450 mm/y. About 90 percent of the one million-
year modeling period is spent under long-term average climate conditions.
The net infiltration rate in the TSPA-VA was assumed to be about 8 mm/yr (DOE98, p. 3-10) for
the current dry climate. This value is substantially higher than the value of about 1 mm/yr used in
TSPA-93 and TSPA-95, and it reflects the results of site characterization studies. The increased
flow includes rapid travel through fast-path fractures which was not apparent from the earlier
equivalent continuum models where fracture and matrix flows were closely coupled. The TSPA-
VA used a dual permeability model to represent the full range of possible fracture-matrix coupling
possibilities. Specifically, UZ transport was modeled using a three-dimensional, dual permeability
finite element code (FEHM).
As noted above, Alloy 825 in the TSPA-95 was replaced with Alloy 22 (a highly corrosion-
resistant nickel alloy) for the CRM in the VA waste packages. The drifts were lined with concrete.
The waste packages were placed on carbon steel supports which in turn rest on a concrete invert to
create level floors in the drifts. A typical 21 PWR waste package was 4.89 m long (without lifting
extensions) and 1.65 m in diameter. The inner barrier of Alloy 22 was 2 cm thick while the outer
barrier of A516 carbon steel was 10 cm thick (DOE98a).
The TSPA -VA was the first performance assessment in which the importance of fuel element
cladding as a long-term barrier to radionuclide release was considered.
The TSPA-VA base case assumed that one waste package would fail by some unspecified juvenile
failure mechanism at 1,000 years after repository closure (DOE98a). The probabilistic base case
assumed 0 to 10 waste package failures at 1,000 years based on a log-uniform distribution.
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The base-case expected-value TSPA-VA evaluations projected dose rates to the average individual
withdrawing water from a well 20 km downgradient from the repository (based on conservative
scenarios and modeling) as follows (DOE98, Figure 4-12):
» 0.04 mrem/yr at 10,000 years
« 5 mrem/yr at 100,000 years
• 50 mrem/yr at one million years
Results of more elaborate probability-weighted dose assessments (DOE98, Figure 4-26) show
mean and median values for the peak dose at 10,000 years of 0.1 and 0.002 mrem/yr, respectively.
Hence, all applicable dose values were found to be well below the proposed 15 mrem/yr individual
protection limit. As discussed in Section 4, these results were developed using highly
conservative, and in some cases unrealistically conservative, assumptions concerning performance
factors and models for training the performance scenarios analyzed.
The analyses found that the most important radionuclides contributing to individual dose for the
first 10,000 years are Tc-99 and 1-129; for the first 100,000 years they are Tc-99 and Np-237, and
for one million years they are Np-237 and Pu-242.
The most important factors contributing to uncertainty in the peak dose rate over the first 10,000
years (in decreasing order of importance) were determined to be the fraction of waste packages
contacted by seepage water, the mean corrosion rate of the waste package Alloy 22 inner barrier (a
contributing uncertainty is the effect on corrosion rates of carbonate dominated ground waters
resulting from contact with the drift lining), the number of juvenile waste package failures, and the
saturated zone dilution factor (DOE98, Figure 4-34). These uncertainties were to be addressed by
the design alternatives examined and selected for the new repository design (EDA II) as described
below.
The TSPA-VA assessment results showed that calculated doses within 10,000 years were
dominated by very conservative release assumptions. These assumptions, in turn, were associated
with arbitrary and non-mechanistic assumed juvenile failures of the waste packages. As a
consequence, subsequent attention focused on improved approaches for evaluating such juvenile
failures.
3.4 Enhanced Design Alternatives -1999
As stated in the VA documentation, the design concept used for the VA and the TSPA-VA
evaluations was intended to be a step in design evolution to the design that will eventually be used
3-16
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for the license application. Even though the site characterization data indicating infiltration rates
that were much higher than previously expected were available for the VA, other data (e.g.,
concerning corrosion of waste package materials) were still limited, and the VA made extensive
use of the results of seven expert elicitations that had been conducted during 1996 and 1997.
Subsequent to publication of the VA, DOE began to develop an improved repository design. The
basis for the design development effort was a group of Enhanced Design Alternatives (EDA). Six
EDA designs were evaluated and the EDA II design (described below) was recommended by the
M&O contractor to DOE as the preferred approach. This recommendation was accepted by DOE
management in September 1999. Design features for the EDA II design are discussed in
Section 3.4.2.
In parallel with DOE's EDA design development effort, substantive action to revise the regulatory
framework was occurring for the first time since the original NRC and EPA regulations for Yucca
Mountain were promulgated in the 1980's. On February 22,1999, the NRC published their
proposed 10 CFR Part 63 regulations which set a dose limit of 25 mrem/yr and eliminated the
subsystem performance objectives included in 10 CFR Part 60. In August 1999, EPA issued for
comment the proposed 40 CFR Part 197 environmental radiation protection standards for Yucca
Mountain (EPA99). These standards would require DOE to demonstrate a reasonable expectation
for 10,000 years after disposal that the annual committed effective dose equivalent to the
reasonably maximally exposed individual is no more than 15 mrem (CEDE). The draft standard
also imposed ground water protection requirements. The EPA's proposed rule had not been
published at the time the ED As were being evaluated, but the individual dose standard is the same
as that incorporated in the generic standard (40 CFR Part 191) and used in the WIPP certification
process.
3.4.1 Basis for the Current Design
Reviews of the repository design concept and performance assessment results for the Viability
Assessment by parties such as the Nuclear Waste Technical Review Board, the NRC, and the
Performance Assessment Peer Review Panel determined that some of the engineered features of
the VA repository contributed significantly to uncertainty in the Total System Performance
Assessment (TSPA) results. Major design factors contributing to performance uncertainty
included:
• The high areal mass (thermal) loading, 85 MTU/acre, and resulting high temperatures
in the rocks surrounding the repository caused significant uncertainties concerning
thermal, hydrological, chemical, and mechanical coupling effects. It also caused
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uncertainties concerning the behavior of rock structure and ground water surrounding
the drifts during repository temperature variations with time.
• The use of concrete lining in the drifts caused concerns about the effect of materials in
the concrete on the chemical constituents in ground water that contacts waste packages
and the effect of those constituents on the corrosiveness of the water.
• The use of carbon steel as the Corrosion Allowance Material and the outer wall of the
waste packages, and use of Alloy 22 as the Corrosion Resistant Material and the inner
wall of the waste packages, caused concern that the carbon steel could create potential
for crevice corrosion of the Alloy 22, thereby increasing the rate of penetration of the
Alloy 22 by about a factor of 25 and consequently greatly reducing the waste package
lifetime.
» The waste packages were not protected from the potential that ground water at the
repository horizon could, at times relatively soon after emplacement, drip onto the
packages and thereby produce aqueous corrosion, enter the package interior, contact the
waste form, leach out radionuclides, and transport the radionuclides to the environment
The DOE's development and selection of an improved repository design was directed at being
responsive to these concerns.
3.4.2 Selection of the Repository Design for the Site Recommendation
DOE used the License Application Design Selection (LADS) process to select the engineered
design for the Site Recommendation. Six Enhanced Design Alternatives (EDA) were defined and
comparatively evaluated. They were identified as EDA options I, II, Ilia, Illb, IV, and V. Options
Ilia and Illb differed in the choice of waste package materials but were otherwise the same.
In defining the EDA options, specific design features were used to address the important
performance uncertainties. All EDA options use a drip shield of corrosion-resistant material to
divert water from the waste packages and to control the waste package environment; all EDA
options also use a corrosion-resistant material as the outer wall of the waste package and limit the
use of cementitious material in the repository. The options differ in their use of high or low
thermal loading, emplacement configurations and waste package energy densities, and backfill.
Use of evaluation criteria and a comparison methodology produced the results of analyses of the
EDA options shown in Table 3-3. These results produced a recommendation by the DOE's
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Table 3-3. Principal Results of EDA Analysis
(Source: K.J, Coppersmith, TRB99a)
-.••', "«'?i".3 ' ^Performance Categories , '
Performance Factors
Licensing
Probability/Safety
Factors
Construction,
Operations, and
Maintenance Factors
Flexibility Factors
Cost
Margin
Time to 25mrem
Peak Annual Dose
Rock Temperatures
Waste Package
Corrosion
Number of Waste
Packages
Length of Emplacement
Drifts
Key Construction,
Operations, and
Maintenance Issues
Emplacement area for
70,000 MTHM
Ability to Change to
Lower Temperature
Ability to Change to
Higher Temperature
Repository Life Cycle
Cost, _
Net Present Value
• EDAI | EDAII
2,500
290,000 years
85 mrem
Always below 96°C
Does not enter
aggressive corrosion
range
15,903
132km
Operational impacts
of more packages and
longer drifts;
blending
1,400 acres
N/A
Requires
development of larger
packages and coupled
models for PA
$25,1 billion
$13,4 billion
3,550
3 10,000 years
85 mrem
>96°C several m's into
drift for hundreds of yrs.
Does not enter
aggressive corrosion
range
10,039
54km
Blending; emplacement
of backfill
i, 050 acres
Requires longer
ventilation
Requires development
of coupled models for
PA
$20.6 billion
$11.0 billion
EDAsfflii/IHb
1,500
290,000/310,000 years
2 15/1 00 mrem
>96°C across most of
repository
Some WPs in
aggressive corrosion
range for 1,000s of
years
10,213
55km
Fabrication of dual .
corrosion-resistant
material package in
nib
740 acres
Requires changes in
drift spacing
N/A
$20.1 billion/
$2 1.3 billion
$10,7 billion/
$11, 4 billion
EtiAIV
180,000
100,000 years
1,200 mrem
>96°C across most of
repository
Humid air corrosion
of WPs begins as
early as 100 years
10,213
60km
Fabrication, welding,
and handling thick
WPsjempI, of
backfill
740 acres
High temp, integral to
WP performance
N/A
$2 1,7 billion
$11. 3 billion
EDAV/
1,250
300,000 years
200 mrem
>96°C across
essentially all of
repository
Some WPs in
aggressive corrosion
range > 10,000 years
10,039
54km
Blending
420 acres
Requires changes in
drift spacing
N/A
$20.0 billion
$10.8 billion
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Management & Operations contractor that the EDA II option be selected for the Site
Recommendation (SR). DOE endorsed the contractor's recommendation in September 1999, and
this design is now being used as the basis for development of the SR.
3,4.3 Comparison of the EDA II and Viability Assessment Designs
The principal EDA II and VA engineered design features are compared in Table 3-4. DOE
estimated that the net present value for development, construction, operation, and closure of the
VA repository would be about $10.1 billion; the estimated net present value for the EDA II
repository is about $11.0 billion (Table 3-3). The cost difference for the two designs is minimized
by the assumption that the drip shields and backfill for the EDA n design would be installed at the
time of repository closure, ie., 50 years or more after the end of emplacement operations.
Table 3-4. EDA D/VA Design Comparison (Source: M.C. Tynan, TRB99a)
Design Characteristics '• .
Areal Mass Loading
Drift Spacing
Drift Diameter
Total Length of Emplacement Drifts
Ground Support
Invert
Number of Waste Packages
Waste Package Material
Maximum Waste Package Capacity
Peak Waste Package Power (bleEding)
Drip Shield
Backfill
Preclosure Period
Preclosure Ventilation Rate
.• .', • . • -.•^ED&nv:-:,:v:;;';.;- • •",
60 MTU/acre
81m
5.5m
54km '
Steel
Steel with sand or gravel ballast '
10,039
2 cm Alloy 22 over
5 cm stainless steel 316L
21 PWR assemblies
20 percent above average PWR
waste package power
2 cm Ti-7
Yes
50 years
2 to 10 cubic m/s
1 ViabiHtVrAssessnient DesiHtf "
85 MTU/acre
28m
5.5m
107km
Concrete lining
Concrete
10,500
10 cm carbon steel over
2 cm Alloy 22
21 PWR assemblies
95 percent above average PWR
waste package power
none
none
50 years
0. 1 cubic m/s
The EDA II and VA designs are compared qualitatively with respect to the performance
uncertainties discussed in Section 3.4.1 in Table 3-5. As shown in this table, the EDA II design, in
comparison with the VA design, has a significantly reduced areal mass loading, no concrete liner,,
a waste package design which has the corrosion resistant material on the outside rather than on the
inside, and use of drip shields and backfill to help reduce and defer contact of water with the waste
packages. Each of these design features is responsive to concerns for performance
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Table 3-5. Impact of EDA II Design Featured on Performance Uncertainties
Design Feature.
Areal Mass Loading
Drift Spacing
Drift Liner and
Invert Material
Waste Package
Materials
Peak Waste Package
Power
Drip Shield
Backfill
¥A ReDOsitbrv
85 MTU/acre
28 meters
Concrete
10 cm carbon steel
over 2 cm Alloy 22
95 percent above
average
None
None
EDA Illleijdsitorv
60 MTU/acre
81 meters
Steel
2 cm Alloy 22 over 5
cm 3 16L stainless
20 percent above
average by blending
assemblies
2 cm Titanium 7
Yes
.:-•-•'•••. • IDA II Impact '^.s^Vf "•••-;
Reduce thermal coupling issues
No temperature rise above boiling point
in rock between drifts; reduces overall
performance uncertainty
Eliminate effect of concrete constituents
on water chemistry; reduce corrosion rates
and radionuclide release rates; increases
package lifetime
Eliminate crevice corrosion potential;
reduce Alloy 22 corrosion rate by factor of
25 or more; increases package life
Reduce thermal gradients; less driving
force for water movement and degradation
processes
Protect waste packages; defer contact by
water and eliminate juvenile failure
potential
Divert water from waste packages; protect
against rockfall
uncertainties in the VA design; each helps to mitigate performance uncertainties and to improve
expected repository system performance with respect to timing and quantities of radionuclide
release. Improvement is obtained either by delaying penetration of the waste package walls or by
changing the expected physical/chemical conditions to reduce the amount of radionuclides that
could be transported out of the BBS by migrating ground water that moves through the
repository.
3.5 Evolution of the Comparative Contributions of Engineered and Natural Barriers to
Repository System Performance
As previously noted, the evolution of repository design and performance has been characterized
by greatly augmented contribution of engineered barriers to performance and greatly diminished
contributions of the natural barriers. The natural barriers of principal significance are the rate of
infiltration of water into the mountain; the water percolation flux at the repository horizon; the
rate of seepage of water into the drifts and onto the waste packages; travel tunes in the
unsaturated and saturated zones; radionuclide holdup on rock formations as a result of sorption;
and dilution of radionuclide concentrations as a result of dispersion and mixing of contaminated
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and tmcontaminated water. Acquisition of data to characterize these performance factors has been
underway since inception of the Yucca Mountain project, is continuing today, and will continue
through the post-emplacement performance confirmation period if a repository is built at the site.
The diminished role of natural barriers in repository performance expectations occurred relatively
abruptly in the 1996-1997 time frame, and was first made evident in the TSPA-VA evaluations
(which, as previously noted, were the first TSPA evaluations for a potential "actual" repository at
the site). In comparison with the prior TSPA studies, the TSPA-VA evaluations used greatly
increased infiltration values and greatly reduced dilution factors for the saturated zone. For
example, the SCP and all TSPA studies prior to the TSPA-VA assumed infiltration rates on the
order of one mm/yr or less; in contrast, the TSPA-VA used a current-climate average infiltration
rate of 7.7 mm/yr and a long-term climate average infiltration rate of 42 mm/yr. Models and
analyses in TSPA-95 projected overall dilution factors for the saturated zone on the order of 1,000
to 100,000; TSPA-VA used a dilution factor range of 1-100 with a median value of 10.
i
These changes were brought about principally by the following:
• In 1996, Flint et al. (FLI96) reported analysis of accumulated site characterization data
which demonstrated that the infiltration rate is on the order of 1-10 mm/yr and is highly
variable over the area of the repository footprint.
» In 1997, D'Agnese et al. reported a regional scale model of the Death Valley
hydrologic regime in Nevada and California (DAG97).
• In 1997, an Expert Elicitation on unsaturated zone flow was conducted; based on
available data, the experts estimated the mean infiltration rates to range from 3.9 mm/yr
to 12.7 mm/yr (DOE97).
• Data showing that Cl-36 from nuclear weapon tests had traveled to the repository
horizon in 50 years or less were interpreted to show that there are fast paths for flow
through the •unsaturated zone, the infiltration rate had to be at least about 2 mm/yr, and
the fast flow apparently took place in the fracture zones (Fab98).
• An improved model for flow and transport in the unsaturated zone, based on integration
of hydrologic, mineralogic, structural, hydrochemical and geochemical site
characterization data, was reported and made available in 1997 for the TSPA-VA
(BOD97).
• An Expert Elicitation on flow and radionuclide transport in the saturated zone was
conducted (GEO98). The experts rejected the models used in TSPA-95 which showed
very large dilution factors, and they emphasized the limitations of processes that would
cause dilution of contaminant concentrations. The experts also took note of the
extreme lack of data to characterize the geohydrologic regime in the saturated zone
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beyond the 5-km boundary of the accessible environment (the result of prior focus on
the requirements of the EPA's 40 CFR Part 191 regulations). The experts expressed
their belief that radionuclide transport would be by movement in vertically thin plumes
through flow tubes beneath the repository; they also recommended that the overall
dilution factor be constrained to the range of 1 to 100, with a median value of 10.
The results of these activities and findings were incorporated into the basis for the models and
performance parameter values used in the TSPA-VA. For example, the Expert Elicitation
recommendations concerning dilution in the saturated zone were adopted directly, and a new one-
dimensional stream tube model for radionuclide transport in the saturated zone was developed in
response to the experts' opinions concerning flow in the saturated zone.
Overall, the models and assumptions adopted for the TSPA-VA analyses resulted in essentially no
contribution to performance from transit and holdup in the unsaturated zone, and dilution of
radionuclide concentrations during transit of the saturated zone to a location 20 km from the
repository occurred by only a factor of 10 in the base case. Dilution during pumping by the dose
receptor was assumed not to occur.
Despite minimization of the role of natural barriers in the TSPA-VA analyses, the TSPA Peer
Review Panel (PRP99) stated, "The current treatment of saturated zone (SZ) flow and transport at
Yucca Mountain is far from satisfactory." The Panel noted three main areas of weakness in the
TSPA-VA treatment:
• The lack of data for some important parameters,
• The incomplete nature of site characterization, and
• Continuing questions regarding the adequacy of the numerical models.
The basic remedy for these weaknesses, which could permit increased and justified reliance on
performance of the natural barriers, is to significantly expand the database of site characteristics
and, by so doing, increase understanding of the functioning of the natural barrier. To do so would,
however, be costly and time-consuming, and may not be necessary given the extreme reliance on
engineered barriers that has been developed to reduce the importance of uncertainties in natural
barrier performance (see the description of the current repository design in Section 3.4.2). Indeed,
in 1996 the Nuclear Waste Technical Review Board noted that "...there are no data to support a
realistic estimate of dilution...[and it is not clear] whether further characterization can provide the
data for reducing the uncertainty...further studies of the saturated zone beyond those now planned
or under way...may not be cost-effective" (TRB96). These considerations indicate that the DOE's
move to a more highly-engineered repository design was directed by a realization of the
limitations of further characterization efforts on the complex flow system in and around the site,
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and the recommendations of external parties to move in the direction of enhanced design to lower
the uncertainties.
At present, Nye County, in cooperation with DOE, is conducting a drilling and testing program
using boreholes drilled approximately along a radius 20 km from the proposed repository location.
These data will expand knowledge of the characteristics of the saturated zone in the valley-fill
alluvium. Data available to date indicate that the geologic formations are highly
complex, and that flow may occur principally in channels within the alluvium (NYEOO). The
results of these and other tests planned by DOE may serve only to confirm that significant
contributions to performance from features of the saturated zone are not to be expected.
In contrast to the situation for the saturated zone, ongoing experiments in the unsaturated zone at
the repository horizon may provide a basis for increased reliance on, or confidence in, perform-
ance of natural features in the unsaturated zone in future TSPA evaluations. Experiments
concerning seepage into drifts (which has been consistently shown by TSPA evaluations to be one
of the most important performance parameters) are showing that seepage is highly limited, and no
natural seepage into drifts excavated to date has yet occurred. A world-wide investigation of
natural analogs has also shown that seepage dripping into underground openings like those that
would be characteristic of the repository is highly limited or non-existent because of capillary
forces (TRBOOa). The most recent report on the seepage work (TRBOOb) indicated that the current
seepage model matches the limited available data reasonably well, and that the model predicts a
seepage threshold of 200 mm/yr for the rock formations at the repository horizon.
Seepage was incorporated into TSPA modeling for the first time in the TSPA-VA; The TSPA
Peer Review Panel found the modeling approach to be "...both novel and informative" (PRP99).
The modeling approach assumed steady-state flow in a fracture continuum, in which seepage starts
where conditions exist for the drift surface to become fully saturated. The percolation flux
threshold was estimated to be in the range 2-3 mm/yr, i.e., approximately the same as the current
infiltration rate.
As noted above, experiments to date are indicating that the seepage threshold is actually on the
order of 200 mm/yr. (This value corresponds to the high end of the values used in the TSPA-VA
for the superpluvial glacial period in the VA climate model.) Available data are, however, limited,
and the threshold will be highly sensitive to geometric and wetting conditions on the drift wall. In
addition, seepage patterns and rates may change as a result of thermomechanical and
thermochemical effects, and rock fall as a result of seismic events. The Peer Review Panel
recommended further testing, which is currently underway (TRBOOb).
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DOE has recently adopted a technique termed'"neutralization''analysis" to characterize the
contribution of individual performance factors to overall repository system performance (TRWOO).
The technique is being applied to the EDA II design; its use, and the relative roles of the
engineered and natural barriers for the EDA II design, are discussed in Section 4.6. In general, the
natural barriers play even less of a role in the current EDA II repository design than in the VA
design because of further augmentation of engineered barriers in the EDA II design.
3.6 Summary of Factors Affecting Evolution of the Repository Design
As described above, the evolution of the design of the Yucca Mountain repository and its
engineered barrier system has been an iterative process occurring, to date, over an eleven-year
period from 1988, when the SCP was issued, until 1999, when the EDA II design was selected to
be the basis for the Site Recommendation scheduled to be made in 2001. The evolutionary process
has been driven principally by the following factors:
« Findings, from site characterization data, that performance of the natural barrier system
will be significantly less than was expected when the SCP was issued. Specifically,
infiltration rates are much higher than had been expected, water travel times in the UZ
are faster than had been expected, and dilution of radionuclide concentrations will be
much less than had been modeled as recently as 1995.
• Findings, from TSPA evaluations of design options and natural barrier performance
models, that the SCP engineered barrier design concepts resulted in a high degree of
uncertainty of ability to achieve eonapEanee with EPA's 40 CFR Part 191 total system
release standards and NRC's 10 CFR Part 60 subsystem performance requirements.
• As a result of DOE/NRC Technical Exchanges, development of NRC's Issue
Resolution Status Reports, and external reviews, development of understanding of the
rigor, depth, and limits on uncertainty that must be addressed in order to prepare a
safety case adequate for licensing reviews.
« Results of external reviews such as those by the NWTRB, the TSPA Review Panel, and
NRC staff, and understanding of the sources and magnitudes of uncertainties and
technical issues in data, performance models, and performance assumptions that are
significant to the adequacy and defensibility of the safety case.
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In summary, the engineered design of the repository has evolved as a result of progress along a
learning curve involving understanding of what the engineered and natural barriers can and cannot
do in the Yucca Mountain setting, understanding of the essential elements of a safety case that is
adequate for licensing reviews, and understanding of the needs for design approaches and data to
bring uncertainties to acceptable levels. Identification of "acceptable levels" of uncertainties is
related to EPA's concept of "reasonable expectation" and NRC's concept of "reasonable
assurance", discussed in Section 5. The EPA standards have included, since promulgation of
40 CFR Part 191 in 1985, and through revised Part 191 in 1993, Part 194, and proposed Part 197,
individual-protection standards of 15 mrem/yr CEDE (or equivalent), human-intrusion standards
of 15 mrem/yr CEDE (or equivalent), and pound water protection standards derived from the Safe
Drinking Water Act.
It is noteworthy that the design evolution has not been driven by EPA's 40 CFR Part 191 standards
concerning radionuclide releases or by anticipated EPA dose standards. Examination of the DOE
performance evaluations to date show that there are many alternative means to reduce uncertainties
in performance projections, even with limited contributions of natural barriers to repository system
performance. What is necessary is to build a solid foundation, through use of data, reasonable
performance models, and reasonable assumptions, to demonstrate that the safety case is a
reasonable and appropriate representation of expected repository performance.
3.7 EDA II Design and the TSPA-SR
As discussed in Section 3.6, DOE has evolved the repository design over a number of years from
one emphasizing the natural barriers of the site to one with much greater reliance on engineered
barriers. Among the reasons for this shift in emphasis was an increasing realization that collecting
data to resolve residual uncertainties in the behavior of the natural system would be more costly
than to develop and use engineered barriers that would eliminate the concern over those
uncertainties. Following the Enhanced Design Alternatives program in 1999 (Section 3.4), the
program focused on the EDA II design as the basis for the next iteration of the TSPA, known at the
TSPA for Site Recommendation (TSPA-SR).
The TSPA-SR is intended as an update and improvement of the TSPA for Viability Assessment
(TSPA-VA) (DOE98a), and as technical support for the Site Recommendation. Changes made to
the TSPA models were intended to address criticisms of the TSPA-VA modeling approaches, to
evaluate the system with more elaborate and soundly based modeling approaches. In addition,
greater emphasis was placed on quantification of uncertainties that were not addressed in the
TSPA-VA. In particular, in TSPA-SR greater emphasis was placed on the potential for igneous .
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disruption of the repository, on waste package* degradation1 mechanisms potentially leading to early
failures, and on potential human intrusion events. Considerably more attention was focused on
evaluating the robustness of model assumptions and the influence of various engineered barriers
than had been done previously.
The TSPA-SR supports the mandated site recommendation process in Sections 112 and 114 of the
Nuclear Waste Policy Act (NWP83, NWP87). The site recommendation is an advanced stage of
development of a recommendation by the Secretary of Energy to the President regarding the
suitability of the proposed site for development. Since it is an integral part of the legal process for
determination of the suitability of the repository to proceed toward a key decision step, the intent
is for the TSPA-SR to be a strongly defensible analysis, and to form the foundation for the TSPA
to be used in a license application.
3.7,1 New Approaches in the TSPA-SR
The primary scenarios evaluated in TSPA-SR are: (1) a nominal scenario, (2) an igneous scenario,
and (3) a human intrusion scenario. In addition, assessments were conducted that evaluate the
robustness of the analysis to extreme assumptions regarding system behavior, such as very early
failure of engineered barriers. These assessments were conducted as part of a series of analyses
intended to investigate "barrier neutralization," "uncertainty importance," sensitivity, and
robustness of the TSPA. As such, they are regarded as parallel and supporting lines of argument in
the Repository Safety Strategy, but are not central to TSPA-SR conclusions regarding regulatory
compliance.
3.7.1.1 The Nominal Scenario
The "nominal scenario" is intended to represent the "sequence of anticipated conditions"
(TRWOOa). This is contrasted with "discrete, unanticipated events that disrupt the nominal case
system" (TRWOOa). That is, the sequence of external events and processes influencing the system
in the nominal scenario represent only gradual degradation processes, with discrete, rapid
degradation processes characterized as "disruptive events." The intent of the TSPA is both to
show "how the system is thought to behave, but also to provide information on how much
uncertainty is associated with each total system performance assessment component..." (TRWOOa).
To that end, the analyses in the nominal scenario are intentionally biased toward
conservatism in assumptions and choices of parameters. Consequently, despite using scenarios
that represent "anticipated conditions," the expected values of the consequences of the nominal
scenario should not be interpreted as the expected consequences of the repository. Instead, the
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"expected values" are a mathematical expression of a conservative representation of reality. This
approach is generally acknowledged to be an appropriate approach to developing defensible TSPA
analyses for repositories. Nevertheless, while a conservative approach to defining performance
scenarios is typically used in TSPAs, proper interpretation of the results and subsequent decision
making must be done with an understanding of the nature and extent of the conservatism
embedded in the TSPA results. These points are key to understanding the TSPA-SR results in the
context of reasonable expectation (described in Section 5) of compliance.
There appears to be consensus among DOE and EPRI commentators that the assumptions in the
nominal case of the TSPA-SR are defensible and conservative, and in some cases very
conservative. EPRI (EPROO) provided a long list of "departures from reality5' in assumptions in
the TSPA-SR. Essentially all potentially non-conservative assumptions listed were offset by an
associated conservative assumption. However, there were numerous conservative assumptions
that were not offset by any balancing approach. Among the most important conservative
assumptions in the TSPA-SR are (EPROO):
• The model for hydrogen absorption on the titanium drip shield can be considered very
conservative since it assumes that all the hydrogen absorbed during general corrosion
will remain in the residual wall thickness and is available to induce hydrogen-induced
cracking (HIC). This constitutes a very conservative assumption for the materials in
the EDA n design. Without hydrogen absorption, dripshield lifetimes would be
extended to greater than 30,000 years (EPROO). The primary effect of modifying this
assumption would be to displace the dose curve out further in time, lowering doses
calculated in the first 100,000 years by perhaps two orders of magnitude.
• The model for crevice propagation, if it were to initiate, is conservative. The crevice
propagation is assumed to progress in a conservative non-mechanistic manner that may
allow moisture ingress into the waste package. However, EPRI (EPROO), in comparing
the potential effects of crevice corrosion on the failure time of the waste packages,
found that it had only moderate effects (about 1,000-2,000 years) on the failure time.
* The initiation of stress corrosion cracking in the annealed final closure weld is a
conservative assumption. EPRI argued that the material properties and the stress-state
the waste package will experience imply that the probability of initiation of stress
corrosion cracking is negligible, approaching zero. Eliminating this mechanism from
the model may delay the onset of releases for several ten of thousands of years (EPROO,
Figure 5-17).
* The cladding is assumed to be in an extremely aggressive environment, representing
severe conditions for corrosion (DOE01). It is assumed that fluoride enters the waste
package and comes in contact with only the cladding. The model does not account for
buffering the fluoride by the basket internals. Accounting for this buffering would tend
to provide a competitive mechanism for reaction of the fluoride, in turn providing a
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much less aggressive environment for the cladding; In addition, for fluoride to enter
the waste package, significant water would need to flow through the crack, diluting the
concentration of the fluoride and lessening the impact. It is unclear whether these
concentrations might be decreased enough to eliminate fluoride corrosion initiation
entirely. If fluoride effects are eliminated, one would expect the onset of releases to be
significantly delayed, since the reaction of cladding with fluoride is the primary
initiation reaction in the DOE model. The TSPA-SR also assumes that the fluoride
contacts the cladding in a limited area, which is argued by EPRI (EPROO) to be
extremely conservative. In presenting an alternative model for cladding corrosion, in
which corrosion was treated as general in nature (not specifically driven by contact
with fluoride), EPRI calculated the median time to cladding failure as between 25,000
and 70,000 years, for dripping and dry conditions, respectively. This result contrasts
with the barrier sensitivity analysis presented by DOE (DOE01, Figure 4-214), which
shows little difference between the base case analysis and one in which virtually no
credit is given for cladding corrosion.
In addition, it is noted that the flow model at the repository level includes an assumption that
seepage initiates when a percolation threshold of 10 mm/yr is reached. Research on this effect
suggests that a threshold value of 200 mm/yr is needed to overcome capillary effects (TRBOOb).
Notably, the only extant measurements associated with the threshold value indicate 200 mm/yr in
the middle nonlithophysal unit of the Topopah Spring Tuff(DOE01, p. 4-92). This value is treated
as an extreme end of a probability distribution in the TSPA-SR. Consequently, this assumption
represents a significant level of conservatism, and particularly overestimates the effects of wet-
climate states. Applying a higher threshold value would imply that the emplacement drifts would
experience dry conditions for a considerably longer time.
A key change to the TSPA-SR compared with the earlier TSPA-VA was the treatment of
manufacturing defects in the waste package. In the TSPA-VA, DOE assumed that some defects
would lead to almost instantaneous releases from the repository. These early failures dominated
the dose consequences hi the period less than 10,000 years. However, these assumed early failures
were somewhat arbitrary and not based on any known mechanism. For the TSPA-SR, the
initiation of early failures was evaluated based on established engineering approaches for
evaluating the likelihood of manufacturing defects, which are subsequently not identified during
inspections. This approach, which is far more reasonable than the TSPA-VA approach, is
nonetheless coupled with conservative models and parameters for corrosion initiation and
propagation. The resulting approach, while still conservative, has shown the early failures used in
the TSPA-VA to be non-mechanistic and implausible (DOE01).
Despite the apparent level of conservatism of the nominal scenario, there are no significant doses
to the RMEI in the time period over which the performance objectives apply. The conservatism of
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the nominal scenario leads to releases and subsequent doses to the RMEI during the period 10,000
to 100,000 years. Less conservative assumptions could well delay the releases until after 100,000
years.
3.7.1.2 Igneous Scenarios
The igneous scenario is subdivided into two scenarios: eruption and intrusion. The eruption
scenario refers to penetration of the repository, leading to total disruption of waste packages and
drip shields encountered by the magma, bringing waste to the surface. Doses result from ash
eruption, with downwind transport, redistribution of ash at the surface, and subsequent human
exposures. The intrusion scenario refers to penetration of the repository by magma, leading to
total disruption of waste packages and drip shields encountered by the magma, but without further
movement of radionuclides. However, since the engineered barriers are assumed to be totally
destroyed, this scenario functions as equivalent to assessing juvenile failures of waste packages.
Releases for the magma intrusion scenarios are via releases to ground water from the disrupted
waste packages.
DOE01 has described the process by which the probability of occurrence of the igneous scenarios
was derived. A panel often experts representing a wide range of expertise was assembled to
interpret the volcanic hazard. The panel evaluated existing data, tested alternative models and
hypotheses, and produced an integrated assessment of the volcanic hazard. The use of this
procedure may have elicited slightly overstated probability of occurrence. The panel was
concerned that some past basaltic activity in the area may have been eroded or buried by younger
sediments. Consequently, the panel formally recognized this possibility by including these
undetected volcanos into their estimates of the number that have occurred. DOEOOa stated that
most common multiplier for hidden events was 1.1 to 1.2 of the known volcanic events, despite the
fact that there is no known episode of magmatic intrusion in the Yucca Mountain region that has
not been accompanied by a surface expression.
The mean estimated annual frequency of intersection of the repository by a dike is 1.6xlO"8. The
5th and 95* percentiles of the annual probability are 7.6xlO"10 and S.OxlO"8, respectively. Shifting
even selected probability values by 10-20 percent is unlikely to reduce the mean annual probability
below the scenario cutoff value of 10"8. Furthermore, DOEOOa cites a series of estimates for the
probability of intersection of the proposed repository at Yucca Mountain published during 1982-
1999. These values cluster between l-3xlO"s, with a few values as high as 10"7 for very conserva-
tive assumptions, and other values as low as 10"10 for less conservative assumptions. Regardless, a
series of investigators have suggested that a probability slightly above 10"8 is credible. Hence,
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while 'the probability may be slightly overstated by the TSPA-SR analysis, it is unlikely that the
igneous scenario can be eliminated solely by arguments related to the probability of occurrence.
By contrast, the consequence analysis conducted for the TSPA-SR appears to be very strongly
biased toward conservatism. All eruptions are assumed to be violent strombolian for their entire
duration. The justification for this assumption is that this is a conservative approach, and that it is
consistent with the capabilities of an existing NRC computer code, ASHPLUME. EPRI (EPROO)
strongly criticized this assumption, and concluded that strombolian eruptions are both rare in
extensional environments like Yucca Mountain, and are not consistent with existing basaltic
deposits associated with past events in the region. EPRI (EPROO) suggested that the Pu'u O'o
eruption of Kilauea Volcano, Hawaii would be a better model for the type of eruption that may
occur in the Yucca Mountain region. This type of eruption would have much less severe
consequences than would a violent strombolian eruption. NRC (NRC99a) notes that such "...low-
energy, low-dispersivity eruptions have limited potential to disperse HLW to critical group
locations."
In the TSPA-SR it is assumed that the magma destroys all waste packages and drips shields that it
contacts, making the mil inventory of those packages available for transport. The justification for
this assumption is that it is conservative, and that other assumptions would be difficult to support
(TRWGOa). The TSPA-SR is based on a very high temperature (1200 C) in the dike. It has been
noted (EPROO) that literature information is available that would indicate that dikes of similar size
to the drifts would solidify in 10 to 20 days, and that the expected contact temperature between the
magma and the containers would be substantially (as much as 40%) lower than the value used by
DOE. Taking these effects into account would drastically reduce release rates associated with this
scenario, since the containers would likely survive intact at lower temperatures. EPRI (EPROO)
also notes the existence of natural analogues for this effect, in which cars, telephone poles, and
other objects in the magma path are embedded in the magma rather than consumed by it. In the
supporting documentation for the TSPA-SR, DOE (DOEOO) acknowledges these temperature
effects, conduct modeling of the thermal interactions of waste packages and magma, and presents a
conceptual model in which the waste packages are primarily intact after interactions with magma.
This conceptual model was not used in the TSPA-SR.
These two assumptions (waste package destruction and type of eruption), if modified, have
the potential by themselves to lead to minimal or zero releases from the waste packages in
the case of igneous activity.
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A number of additional assumptions in the TSPA-SR igneous models are also conservative
(EPR00), but would tend to have less profound impacts on the results:
• Effects associated with magma viscosity and velocity are conservative. It is assumed
that sufficient magma enters the emplacement drift to contact between 6 to 18 waste
packages and move them around, contributing to waste package failure. Assumptions
of less violent behavior would tend to decrease releases directly in proportion to the
number of damaged containers.
• The assumed waste form particle size after disruption is conservative. When the waste
form is exposed to the erupting magma it is assumed that the spent fuel is pulverized
into very fine particles. The shearing forces involved in magma eruption are unlikely
to be able to cause enough grinding of the ceramic fuel to pulverize the majority of the
fuel into a fine powder. This is conservative for the eruption scenario because a fine
powder is more easily dispersed over long distances. This assumption is inconsistent
with the conceptual model of dike-waste package interactions presented by DOEOO. In
that report, waste packages were described as being substantially intact following
interactions with a dike. If the waste is not pulverized during the eruption, the eruption
scenario, which relies entirely on an airborne pathway, would likely be inconsequential.
• The fuel particles are assumed to be on or near the top of all of the magma and eruptive
material as it falls back to earth. This assumption is conservative since the majority of
the dose from the eruptive scenario is via the inhalation pathway. Waste buried deeper
within the fallen ash is less likely to be resuspended by the wind. The particle size
assumption discussed above would make this assumption even more conservative.
• The wind is conservatively assumed to always blow toward Amargosa Valley, thereby
ensuring the ash faE lands on the greatest local population. The SCP Chapter 5
(DOBSSa) shows that no more than about 15 percent of the surface winds are from the
north, and at higher elevations winds are generally from the east or southeast.
Consequently, this assumption likely represents a conservatism of on the order of a
factor of 2-3 in the probability-weighted dose.
• A magma conduit is always assumed to be centered on a drift. This will tend to be
conservative since a conduit not centered on a drift should intersect less waste
containers. Based on the ratio of the area of the drifts to the area of the repository, this
assumption is likely to be conservative by less than an order of magnitude.
* The major faults on either side of the repository have the potential to divert any magma
around the repository. This has been conservatively ignored. The effect of accounting
for such diversion around the repository would be to lower the probability of its
occurrence. Given that the mean probability of occurrence of the scenario is, only
marginally above the value that should be considered in the TSPA5 altering this
assumption may eliminate the igneous scenario from further consideration.
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3.7.1.3 Human Intrusion Scenario
The human-intrusion scenario is a hypothetical analysis of the potential effects of a drilling event
at the site. In this analysis, a stylized drill hole is assumed to penetrate a waste package and
continue to the saturated zone. The scenario therefore serves both to disrupt a waste package
prematurely, and to provide a reasonably enhanced pathway to the saturated zone. DOE
developed the human intrusion scenario for the TSPA-SR to be consistent with existing guidance
in the proposed 40 CFR 197 (EPA99), the proposed version of 10 CFR 63 (NRC99), and the
proposed version of 10 CFR 963 (DOE99a). The implementation of the regulatory requirements
was conducted in the TSPA-SR as shown in Table 3-6 (TRWOOa). The central feature! for
treatment of these requirements was to be consistent with the more conservative of the proposed
requirements from the draft regulations. Most notably, the intrusion is assumed to occur at 100
years, consistent with the proposed NRC requirement (NRC99). Intrusion at later times, when
(consistent with EPA99) a waste package might reasonably be more degraded to allow an
unrecognized drilling penetration, was treated as a sensitivity case study. :
As illustrated in Table 3-6, similarities between the proposed 40 CFR Part 197 and the proposed
10 CFR 63 consist of:
• the intrusion event is a single borehole that penetrates a waste container and continues
to the saturated zone,
• doses to the driller are not considered,
• doses are evaluated only for gradual processes occurring at the repository, and
• borehole properties are consistent with current technical practices.
The primary differences between the two proposed regulations are: .
• different dose criteria (15 vs. 25 mrem/yr), and ;
• the time of intrusion (100 years vs. a credible time for unrecognized penetration).
The DOE approach presented in Table 3-6 was to be consistent with the proposed regulations
where they are consistent, and to consider both proposed regulations where they differ. The
human intrusion standard in EPA's final regulations is unchanged in the aspects described in
Table 3-6 (EPA01).
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Table 3-6. Implementation of Regulatory Requirements in the TSPA-SR for Regulatory
Requirements (Table adapted from TRWOOa). Key differences between the NRC
and EPA assumptions are indicated as underlined text.
NRC Base Assumptions (from
Proposed 10 CFR Part 63)
Assumed intrusion is a drilling
event.
Drilling result is a single, nearly
vertical borehole that penetrates a
waste package and extends down to
theSZ.
Intrusion occurs 100 years after
closure
Borehole properties (diameter,
drilling fluids) are based on current
practices for resource exploration.
Borehole is not adequately sealed to
prevent infiltrating water.
Hazards to the drillers or to the
public from material brought to the
surface by the assumed intrusion
should not be considered.
A separate consequence analysis is
required, identical to the
performance assessment, except for
the occurrence of the specified
human intrusion scenario.
Peak dose is not to exceed 25
mrem/yr. in the first 10,000 years.
EPA Additional and/or Conflicting
Assumptions (from Proposed 40
CERPartlST)
Assumed intrusion is an acute and
inadvertent drilling event.
Borehole penetrates a degraded
waste package, and extends to the
SZ.
Intrusion time should take into
account the earliest time after
disposal that a waste package could
degrade sufficiently that current
drilling techniques could lead to
waste package penetration without
recognition by the drillers.
Borehole results from exploratory
drilling for ground water. Borehole
properties are consistent with current
practices.
Natural degradation processes
gradually modify the borehole, the
result is no more severe than the
creation of a ground water flow path
from the crest of Yucca Mountain
through the potential repository and
to the water table.
Only consider releases through the
borehole to the SZ; consider releases
occur gradually through air and
water pathways, not suddenly as
with direct removal.
Unlikely natural processes and
events are not included, but analysis
could include disturbances by other
processes or events that are likely to
occur. Separate consequence-only
analysis.
Peak dose is not to exceed J_S
mrem/vr., in the first 10,000 years.
• ;
Conceptualization for-TSPA-SK ;
Inadvertent drilling event.
Single vertical borehole from surface
through a single waste package to
the SZ.
Intrusion occurs at 100 years (a
10,000 year intrusion time is
examined in a sensitivity,
simulation).
Borehole diameter consistent with an
exploration ground water well.
Infiltration and transport through the
borehole assumes a degraded,
uncased borehole, with properties
similar to a fault pathway.
Ground water is only pathway
considered.
Intrusion borehole is applied to
nominal case; effects of volcanism
are not included.
Does not affect simulations.
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The approaches used in TSPA-SR for evaluating these conditions are shown in Table 3-7. The
analyses are based on a representation of an exploratory drilling intrusion, which leads to
disruption of a waste package and an enhanced pathway through the unsaturated zone. The
saturated zone and biosphere analysis are the same as in the nominal scenario.
3.7.2 Results of the TSPA-SR
The results of the TSPA-SR show the following characteristics. The results are composed of the
combination of the nominal scenario and two igneous scenarios. The dose curves from these
scenarios are weighted by their probabilities so they can be combined, as shown in Figure 3-2.
These curves are then intended to be compared with proposed dose criteria, which are also shown
in Figure 3-2. Human intrusion is treated as a separate scenario, which is not combined with the
results from the nominal and igneous scenarios.
The nominal scenario produces nil dose values during the compliance period (<10,000 years). The
only significant doses associated with the nominal scenario occur in the post-compliance period
(>10,000 years). This is the result of complete containment of the waste by the design-basis
engineered barrier system during the first 10,000 years.
TRW (TRWOOa) states that doses in the first 2,000 years after closure are dominated by the
eruption scenario. From 2,000 years until after 10,000 years, the doses are dominated by igneous
intrusion followed by releases to ground water from the magma-disrupted waste packages. After
10,000 years, the dose curves are a more complicated function of the probability weighted doses
from each of the three scenarios (nominal, eruption, and intrusion).
In all cases the mean dose rate from the combined scenarios is substantially less than the
regulatory standards over 10,000 years. In addition, analyses presented in the TSPA-SR
(TRWOOa) show that none of the TSPA realizations exceeded any of the proposed regulatory
criteria during the 10,000-year compliance period. As discussed in Section 3.7.1 above, the results
within 10,000 years are likely to be extremely conservative because of the conservative treatment
of igneous activity. Modified assumptions for repository behavior during interaction with magma
have the potential to eliminate all calculated doses in the first 10,000 years.
It is interesting to contrast these results with earlier TSPA results presented in the TSPA-VA
(DOE98). In the TSPA-VA, doses in the period less than 10,000 years were dominated by
artificially introduced juvenile failures of the waste containers from manufacturing defects. These
early doses have been eliminated in the TSPA-SR through a combination of an improved waste
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Table 3-7. Technical Assumptions Implemented in the Human Intrusion Scenario in TSPA-iSR
(Table excerpted from TRWOOa).
Issue
Borehole diameter
Infiltration into borehole
Seepage into penetrated
waste package
Type of waste package
penetrated
Thermal and geochemical
conditions in waste package
Waste form degradation
Solubilization of
radionuclides in water
Borehole flow and transport
properties
Borehole location
Borehole length
SZ
Biosphere processes
Key Component
Affected
Infiltration
Borehole Transport
Infiltration
Infiltration
Waste Mobilization
Waste Mobilization
Waste Mobilization
Waste Mobilization
Waste Mobilization
Infiltration
Borehole Transport
Infiltration
SZ Transport
Borehole Transport
SZ Transport
Biosphere
.:' '"V:A -.";/'"•" 5^:.^ • '•'/": •.•'" ' -.• ^--v- ''• /'•'
i TSPArSR Implementation x
Typical water well borehole has a diameter of 20.3 cm
(Sin.)
Assumed infiltration rate distribution is based on modeled
infiltration in the Yucca Mountain region for the glacial
transition climate. Values at the high end of the
distribution inherently include the possibility of surface
water collection basin focusing.
Volumetric flux is equivalent to infiltration rate times
borehole area. Volume of drilling fluid is ignored.
Sampled from CSNF and co-disposed waste packages,
Co-disposed packages contain both DSNF and HLW glass.
Assume temperature and in-package chemistry as
calculated in nominal scenario. This assumes Well J-I3
water and ignores any chemical effects of the drilling fluid.
Waste in penetrated package is assumed to have perforated
cladding from drilling disturbance.
Infiltrating water can mix with waste in entire waste
package. Solubility is based on temperature and in-
package chemistry as in nominal scenario.
Volumetric flux consistent with seepage into the waste
package. Transport properties consistent with a UZ fault
pathway.
Random over the footprint of the potential repository.
Uncertainty in location is captured hi infiltration rate and
location that radionuclides enter the SZ.
Borehole length from the potential repository to SZ
conservatively assumes water level consistent with glacial
transition climate.
Assume SZ flow and transport properties identical to
nominal scenario.
Assume exposure pathways and receptor characteristics
identical to nominal scenario.
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£,
E
o>
"as
105
104
103
102
JNRC Proposed 25 mrem/yr IPS
HIS
/Nominal Mean Dose Rate
j i i i i i i i
1000
10,000
Time (years)
100,000
Figure 3-2. Comparison of Radiation Protection Standards with Expected
Values of TSPA-SR Calculations for a Repository at Yucca
Mountain for Nominal and Igneous Scenarios (Figure adapted
from TRWOOa) ;
package design, and improved, more realistic modeling of juvenile failures associated with the
new design. However, in the assessments of doses within 10,000 years these juvenile failures from
manufacturing defects have been replaced in the TSPA-SR by juvenile failures associated with the
igneous scenarios, with their associated assumptions about early complete destruction of the waste
containers, and very conservative assumptions for eruption characteristics.
Mean dose-rate results from the human-intrusion scenario are presented in Figure 3-3. As
discussed in Section 3.7.1, the base case represents a conservative assumption of intrusion at 100
years, in keeping with NRC guidance (NRC99). Mean dose-rate results from a sensitivity case are
also shown on the figure, in which the intrusion occurs at 10,000 years in keeping with EPA
guidance (EPA99). The mean dose-rate is not significantly higher at 100 years than at 10,000
years. The mean dose-rate is well below the relevant regulatory standards at all times.;
3.8 DOE's Current Program Costs
The cost figures in Table 3-8 reflect DOE's most recent estimates (DOEOla) for both historical
costs for the repository program to the year 2000, and projected costs through the closure and
decommissioning phases. These cost estimates are adjusted to a common basis of constant dollars
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1 10-V
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f
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i i i MIH
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: i i i i M n
100
Figure 3-3.
100© 10,000
Time (years)
100,000
Expected Values of TSPA-SR Calculations for a Repository at
Yucca Mountain for the Inadvertent Human Intrusion Scenario
(Figure adapted from TRWOOa)
at year 2000. Table 3-8 retains DOE's cost estimates that were presented in the Viability i
Assessment documents (DOE 98) for site characterization work, since comparable detail for these
expenditures were not given in the newest cost estimates.
Cost figures indicate that the combined cost of the EDA n design waste package and drip,shield
fabrication is estimated at $13.2 Billion, Emplacement costs for the waste package and drip
shields is estimated at an additional $8.2 Billion (DOEOla, p. 3-10), giving a total cost of
implementing this component of the EDA n design of $21.4 Billion. This sum is considerably
higher than the cost of planned additional site characterization investigations and reflects DOE's
choice to use enhanced engineering to reduce or eliminate uncertainties in the behavior of the
natural barrier.
As discussed in Chapters 4 and 5 of this document, overly conservative assumptions included in
performance assessment scenarios produce dose projections that will be considerably higher, by
orders of magnitude, than what should be expected for more realistic assessments. Typically,
performance assessment analyses are deliberately framed with conservative assumptions. This is
done to provide a measure of confidence that the assessments represent a conservative, and
3-38
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Table 3-8. Estimates of Costs for the Yucca Mountain Program*
1. Historical Total, Mined Repository FY 1983-2000 (DOEOla TSLCC, p. 1-2)
2. Complete Work to License Application (DOEOla TSLCC, p. 1-3):
3, Details of Completion Work, FY 1999-2002 (DOE98, VoL4, Table 6-2)
Site Investigations (total)
Nye County
SZ data analysis
SZ modeling
Repository Design
Performance Assessment
Final analyses
EIS
Site Recommendation
Licensing
Field Operations
Other Support
Financial Assistance
$15.6 million
3,4
2.2
8.3
4. Repository (2003-21 19) (DOEOla, p. 3-8)
Licensing (2003-2006)
Pre-Emplacement Construction (2006-2010)
Emplacement Operations (2010-2041)
Monitoring (2041-2110)
Closure and Decommissioning (2110-2119)
5. Design Options to the VA Repository
$ 189.2 million
296.5
63.6
64.1
2.9
76.6
106.1
277.3
61.8
$1138.1 million :
$ 1.3 billion
4.4
19.7
6.0
4.0
$ 35.4 billion
Drip Shields and Backfill Fabrication (DOEOla, p. 13-2)
6. Total Repository Cost (1982-21 19) (DOEOla, p. 3-8)
7. Total Program Cost (DOE01, o. 1-2") $49,3 B + Historical Costs S8.2B
S8.2B
S0.8B
$35.4 B
$13.2 B
$36.3 B
$57.6 B
* Costs from the Total System Life Cycle Cost Estimate (DOEOla) are in constant year 2000 dollars.
perhaps "worst case" analysis so that the acceptability of the disposal system's projected
performance can be evaluated with a greater public acceptance and a fundamentally conservative
performance case for the licensing process. Counterbalancing this conservative assessment bias
must be a recognition that excessive conservatism in framing performance scenarios can lead to
design choices which may be significantly more "robust" than necessary to provide a reasonable
3-39
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expectation of satisfactory performance. Greatly increased costs can result if the conservative bias
in framing performance scenarios is taken to excess. Chapters 4 and 5 of this document discuss
the evolution of DOE's performance assessment approaches for the Yucca Mountain repository,
and the conservatism incorporated in them, as well as the contrast between these performance
scenario assumptions and the "reasonable expectation" approach inherent in the Agency's
standard. '••
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4.0 EVOLUTION OF PERFORMANCE ASSESSMENT AND BARRIER ROLES
This chapter summarizes and evaluates the repository system performance assessments that
have been conducted by the Yucca Mountain program. Results of recent performance
assessments demonstrate that the current repository design is able to meet, by a large margin, a
15 mrem CEDE individual-protection standard and the ground water protection and human-
intrusion standards,
This section presents and discusses performance assessment results that have been conducted by
DOE for Yucca Mountain. It also discusses conservatism in the models and assumptions that led
to the assessment results, and alternative results that might be obtained through selection of
alternative dose receptors or repository designs. The sub-sections of this chapter examine DOE's
performance assessments to date and the use of conservatism in the definition of the performance
scenarios and their analysis.
There will always be uncertainties inherent in modeling the interaction of the natural and
engineered components of the repository system over the long time frames involved hi projecting
the repository's performance, and the performance projections are always subject to these
uncertainties. Uncertainties should not always be assumed to mean the repository performance
will be worse than quantitative estimates indicate, but it is always desirable to reduce uncertainties
to the extent possible and practical. To reduce uncertainties, the DOE repository effort could elect
to enhance the repository engineered components to reduce or eliminate the potential effects of the
uncertainties, or expend additional effort to characterize and model the interaction between the
natural and engineered systems more realistically to remove overly conservative assumptions used
in prior assessments. The results of the assessments described here indicate that the repository
design evolution was not driven by the components of the EPA standard, but rather by the
uncertainties in the interaction of the natural and engineered systems at the repository site, as well
as the very conservative approach taken in framing the performance scenarios in the DOE
performance assessments.
4.1 Performance In Comparison with the Individual-Protection Standard
The TSPA-SR included a comprehensive TSPA effort, and was intended to be a complete
demonstration of the ability of the system to meet proposed technical requirements. The TSPA-SR
performance evaluations used a complex system of linked computer codes to model the
performance factors; and used a suite of highly conservative assumptions concerning performance
of the engineered features of the repository as the basis for the performance models. Most notably,
the TSPA-SR assumed violent disruption of the repository by strombolian igneous intrusion,
4-1
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leading to complete destruction of waste packages contacted by the magma. As discussed in
Section 3.7, both the existence of strombolian activity at Yucca Mountain and the subsequent
behavior of the magma in contact with waste packages are questionable, and are likely to be
extremely conservative. Modification of any of the key assumptions associated with the igneous
scenarios would likely lead to negligible releases from the repository in the 10,000-year
performance period.
The TSPA-SR represents the latest step in an evolution of the TSPA of Yucca Mountain.
The earlier TSPA-VA methodology and assumptions were used to produce the performance
assessment results presented in the DEIS for a repository at Yucca Mountain (DOE99). A key
point is that the TSPA-VA analyses of the anticipated conditions (nominal scenario) were
generally more conservative than those in the TSPA-SR. Despite this additional conservatism, the
TSPA-VA was able to meet all applicable standards for Yucca Mountain. Consequently, TSPA-
VA results for the nominal scenario continue to be relevant as a conservatively biased
representation of Yucca Mountain performance relative to current understanding and the current
EDA H design. Furthermore, this means that conclusions made in the DEIS regarding the ability
of Yucca Mountain to meet performance objectives are still correct and appropriate.
Minor modifications to the TSPA-VA models were made for the DEIS evaluations in order to
accommodate the DEIS options that were considered (e.g., alternative areal mass loadings and
alternative waste quantities disposed), but the intent for the DEIS performance evaluations was to
use the same basis used for the TSPA-VA evaluations. The DEIS included estimates of
radionuclide concentrations in ground water mat can be compared with EPA's ground water
protection standards, discussed in Section 4.2.
The uncertainties in performance of the EDA II repository are also significantly less than those for
the VA repository; as previously discussed, and as illustrated in Table 3-5, the EDA II design
features were selected specifically to reduce performance uncertainties as well as to improve the
margin between, expected performance and the regulatory standard.
In summary, it is evident that the expected performance in TSPA-SR is significantly better than
that of the VA repository; this is the result of design features specifically selected to improve
expected performance and to reduce uncertainties in expected performance. Furthermore,
improved model rigor and supporting data have eliminated consideration of juvenile failure
mechanisms that led to early releases in TSPA-VA. Currently, the only credible mechanisms for
release from the repository in the performance period are associated with igneous activity. As
4-2
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discussed earlier, this scenario is treated with extreme conservatism. A more reasonable treatment
of igneous activity would likely lead to negligible releases from this scenario.
The EPA individual-protection standard of 15 mrem/yr at 10,000 years and 18 km therefore is not
controlling or forcing, DOE's approach to repository design. As discussed in Section 3, the
evolution of the repository design, performance assessment methodology, and performance
assumptions has been driven by factors other than the EPA IPS standard,
4.2 Performance in Comparison with the Ground Water Protection Standards
In the DEIS for Yucca Mountain, DOE calculated and reported ground water concentrations of
radionuclides released from a repository at Yucca Mountain. The evaluations used the VA design
and modeling methods and were, therefore, as previously noted, highly conservative, i.e., they
overstate the expected concentration by several orders of magnitude. Furthermore, they overstate
expected concentrations with respect to the current EDA II design and the TSPA-SR results.
The results of the DEIS concentration evaluations for the radionuclides released during periods up
to 10,000 years and transported to locations at 5, 20, and 30 km downstream from the repository
are summarized and compared to the current (1976) Maximum Concentration Limits (MCLs) in
Table 4-2. The DEIS concentration values are strongly influenced by the assumed juvenile waste
package failure at 1,000 years and by assumptions of limited dilution during transport. As a result
of the assumptions that maximize the amount of release from the repository and minimize dilution
during transport, the radionuclide concentrations shown in Table 4-2 are much higher than would
reasonably be expected with more realistic assumptions for the performance scenarios.:
As can be seen in Table 4-2, the concentrations reported in the DEIS for the TSPA-VA repository
are well below the current MCL values despite the conservative assumptions and design that are
the basis for the performance calculations.
As shown above in Section 3.7, no radionuclide releases from the EDA II repository would be
expected during 10,000 years unless it is violently disrupted by volcanic activity. The results for
the EDA II design from the TSPA-SR for comparison with the ground water protection MCLs are
shown in Figures 4-1 and 4-2. The ground-water protection analyses assumed a representative
water volume of 1285 acre-feet/yr centered on the highest concentration in the plume in the
saturated zone. It was recognized in the TSPA-SR (TRWOOa) that the regulatory time period for
ground-water protection is 10,000 years. However, the analyses were carried out 't
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Table 4-1. Comparison of DEIS Ground Water Radionuclide Concentrations with MCLs
Radionuclide
Contributors
to lOK-Year
Dose
Tc-99
1-129
C-14
Current (197.6)
MCL, in pCS/1
900
1
2,000
Mean Cone.
for 85
MTU/acre*,
5km
20km
30km
45
30
10
0.13
0.07
0.04
2.1
1.1
0.64
95th PercentOe
Cone, for 85
MTU/acre,
5km
20km
30km
390
84
130**
0.57
0.12
0.20
8.2
1.8
3.1
Mean Cone,
for 25 -
- MTU/acre,
5km
20km
30km •
17
7.3
4.5
0.10
0.50
0.02
1.6
0.79
0.40
95* Percentile
Cone, for 25
MTU/acre, "
5km
20km
30km
1.9
14
6.3
0.40
0.15
0.0
5.6
5-?
0.21
* The 85 MTU/acre thermal loading is the VA design value. The DEIS Proposed Action corresponds to the VA
design, but the DEIS also considered options of 60 and 25 MTU/acre. '
** The apparent inversions of concentrations with distance are a consequence of the modeling methods used for
the DEIS performance evaluations.
1000
10,000
Time (years)
100,000
Figure 4-1. Summary of Groundwater Protection Performance Results of the
TSPA-SR: Combined Beta and Photon-Emitting Radionuclides
(Figure adapted from TRWOOa)
4-4
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103
102
1
10°
10"1
I 10-2
I 10*
03
10*
Gross Alptra Activity (excluding Rn and U)
Total Radium Activity (Ra-226 and Ra-228)
1000
10,000
Time (years)
100,000
Figure 4-2. Summary of Ground-Water Protection. Results for TSPA-SR
for Gross Alpha Activity (Figure adapted from TRWOOa)
to 100,000 years to ensure that no significant degradation of the performance occurs after 10,000
years.
The performance of the repository in the TSPA-SR is shown to be significantly improved
compared to the performance presented in Table 4-2 for the TSPA-VA over 10,000 years. This
dramatic improvement in calculated performance is the result of improved design and more
credible treatment of the failure of waste packages.
Sequential analyses on several designs and using several TSPAs have been analyzed for
comparison with current ground water MCLs. These have included comparisons in the DEIS,
TSPA-VA, and TSPA-SR. In the TSPA-VA the MCLs were met by a substantial margin, despite
significant levels of conservatism built into model assumptions, which would increase the
compliance margin to orders of magnitude if more realistic scenario and model assumptions were
used. In the DEIS, the MCLs were met despite even more conservatism applied to the analysis. In
the TSPA-SR, ground-water concentrations are projected to be zero for the first 10,000 years. The
current ground water protection MCLs therefore are not expected to affect the repository design or
costs.
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4.3 Conservatism in the TSPA-VA, TSPA-DEIS, AND TSPA-SR Evaluations
As previously noted, DOE exercised considerable conservatism in the modeling methods and
E
assumptions for the TSPA-SR dose projections. These assumptions were more realistic and less
conservative than the earlier TSPA-VA approaches for the nominal scenario, but still retain a
significant conservative bias. Both TSPA-SR and TSPA-VA reports, and their supporting
technical basis documents, provide a comprehensive description of the modeling methods and
assumptions. The DEIS states that the TSPA-VA methods and assumptions were used to produce
the TSPA-DEIS results, except for minor modifications to accommodate the waste inventory and
thermal loading options considered in the DEIS but not considered in the VA. Since the TSPA-
VA has been shown to be more conservative than the TSPA-SR for the nominal scenario, the
results and conclusions of the DEIS remain appropriate.
The strategic approach used by DOE for TSPA-VA, TSPA-DEIS, and TSPA-SR modeling and
assumptions can be summarized as follows:
« Values and distributions for natural system performance parameters such as water
infiltration rates were as realistic as possible on the basis of data available at(the time of
the analyses. Uncertainties in these performance factors were so high that it would be
difficult to identify and characterize conservatism for them; values for many of these
parameters, such as dilution during transit of the saturated zone, were based, as
necessary, on the results of expert elicitations.
* Biosphere dose-conversion factors were as realistic as possible on the basis of standard
pathway parameters and local data on current human locations and activities'such as
farming.
* Some performance factors that could contribute significantly to deferral of radionuelide
release from the repository, to reduction of release quantities, and to reduction of
radionuclide concentrations in the biosphere were simply omitted from the performance
model system if parameter values and distributions of values could not readily be
established and defended. Such factors include dilution of radionuclide concentrations
in water within a failed package, delays in radionuclide release from a failed; package as
a result of low water entry rates, and dilution of ground water concentrations at the
dose receptor location as a result of pumping. Each of these factors will tend to either
delay doses to later times, or to lower the peak dose.
• Conservatism was exercised for engineered barrier system performance parameters, for
which a data and/or experience base exists and enables a characterization of;
conservatism. This implementation of conservatism is discussed below for specific
performance factors: juvenile failures, crevice corrosion, water flow into the package
interior, exposed waste form area, and in-package dilution and transport delays.
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4.3.1 Assessment of Juvenile Failure
In the TSPA-VA, doses prior to 10,000 years were dominated by juvenile failures, specifically by
the potential for weld failures associated with defects at emplacement despite rigorous inspection
procedures. The potential for juvenile failures is inevitable, owing to the possibility for human
errors in manufacturing, inspection, and emplacement. In the absence of such effects, the design
basis lifetime for the waste package in the TSPA-VA was very long, and precluded releases during
the first 10,000 years, with early corrosion failures limited to less than 20 waste packages out of a
population of about 10,000 within 10,000 years (DOE98a, Volume 3, Figure 4-13). In the TSPA-
VA, therefore, the potential for these problems was treated using a conservative screening
approach. The subsequent results therefore constituted a real-world worst case scenario.
104
nr io3
•^ 102
I 10'
^ 10°
c 10-3
io-4
j. 1 i 1 1 1 • • I I : : : . ....
• Maan Annual Dosa for Single CSNF Juvenile Failure
j -——Mean Annual Dose for Single Co-Disposal Juvenile Failure •
:
:
!
!
•
: ^**~
~ — »T*,.. , .... ., , ,.i. ,„, . .,.-,..,:
1000 10000
Time Since Closure (years)
100000
Figure 4-3. Estimates of the Consequence of an Artificial Juvenile
Failure
Penetration of a single waste package was assumed in the TSPA-VA base case to occur at 1,000
years as the result of a phenomenon such as failure of a bad weld. The TSPA-VA assumed entry
and exit holes form at the same tune. Seepage was assumed to enter the package, since the entire
waste package was assumed to be wetted. These assumptions provided essentially an instantan-
eous high release rate, which is an unrealistic and very conservative treatment of weldTfailure
effects. The penetration was assumed to result in immediate release of radioactivity from
1.25 percent of the cladding. This single package failure assumption contributed about 50 percent
of the base-case 10,000-year dose rate of 0.04 mrem/yr. For this juvenile failure to occur, water
would have to drip directly onto a bad weld. Absent this juvenile failure assumption, penetration
4-7
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of a waste package wall was not be expected to occur sooner than about 4,000 years, and
penetration at that time would occur only if crevice corrosion occurs.
This mode of failure was determined to lack credibility for the design used in TSPA-SR. Instead,
juvenile failures were evaluated using a more elaborate model of the corrosion of the EDA II
design system accounting for the likelihood of technical, administrative, and inspection failures
i
and their distribution at the waste package surface. As discussed in Section 3.7, the resultant
treatment of corrosion remains quite conservative in its treatment of the details of the corrosion
mechanisms (e.g. hydrogen absorption, stress corrosion cracking, crevice propagation).;
In addition, sensitivity analyses were conducted to assess introduction of an artificial juvenile
failure (TRWOO) at 100 years for the EDA II design. This assessment is not based on any known
mechanism, and is not considered to be a credible occurrence. It was evaluated solely for the
purpose of evaluating extreme behavior in the system and investigating the role of the wjaste
package in system performance. In addition, the release mechanisms associated with this juvenile
failure were, as discussed above, treated in a very conservative manner leading to rapid releases
from the waste package. Hence, this analysis represents a comparable approach to the manner in
which waste package failure was treated in the TSPA-VA. Results of this juvenile failure are
shown in Figure 4-4. Even in these extreme conditions of unrealistic failure behavior at very early
times, the resulting doses are not large.
4.3.2 Local Crevice Corrosion of Alloy 22 ;
Early penetration of the corrosion-resistant Alloy 22 waste package was assumed hi the TSPA-SR
to occur as a result of crevice corrosion, which produces a local pit-type penetation. The Alloy 22
is assumed to be potentially vulnerable to crevice corrosion as a result of water dripping directly
on it from a point in a failed drip shield. The electrochemical conditions for crevice corrosion are
not expected to occur in the repository (TRWOOa). This is a significant modification from the
TSPA-VA design and analysis, in which crevice corrosion initiated as a result of its being under a
carbon steel outer wall. As a result of this design modification, crevice corrosion of this type no
longer plays a significant role in early waste package failures. Consequently, in the TSPA-SR the
waste packages fail either as a result of manufacturing defects or by general corrosion. The net
effect of this change in mechanism is a significantly longer expected lifetime for the containers,
with juvenile failure becoming far less important than in the TSPA-VA.
4-8
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0.35
u. 0.3 -
>
£
i 0.25 H
E
8 0.2 H
o
"O
o
0.15 -
0.1 -
0.05 -
0
85 MTU/acre
10
15
20
25
30
Km
Figure 4-4. 10,000-Year Dose-Rates for Alternative Areal Mass Loadings (compiled from
DOE 99)
4.3.3 Water Flow Into the Package Interior ;
The amount of water that enters the interior of a penetrated package and can contact the exposed
waste form depends on the precipitation rate onto the top of the mountain, the fraction of the
precipitation that infiltrates into the mountain, the fraction of the infiltration flow that arrives at the
repository horizon as percolation flux, the fraction of the percolation flux that seeps into the drifts,
the extent to which the surface of a waste package contacted by seepage flow is wetted, and the
fraction of the waste package surface area that is open, as a result of corrosion, to permit seepage
water to enter the interior.
Key elements of the TSPA modeling of these performance factors included the following:
« Precipitation and infiltration as a function of location in the repository footprint were
characterized, for current climate conditions, using available site characterization data.
* After 600 years, the climate is assumed to change to what was termed long-term
average conditions, under which the precipitation and infiltration rates are '
approximately five times greater than for current climate conditions. This is a
modification from the TSPA-VA, in which the change was assumed to occur 5,000
4.9
-------
years in the future. The estimate of a 600-year initiation of this wetter climate state is
argued in the TSPA-SR to be representative of past climatological cycles. Hpwever,
EPRI (EPROO) has suggested that this assumption does not adequately account for
greenhouse effects on climate over the next few hundred years. They argue that
greenhouse effects may well lead to a drier climate over a significant length of time.
• The total percolation flow at the repository horizon is assumed to be the same as the
infiltration flow, i.e., there was no holdup or lateral diversion during flow through the
unsaturated zone above the repository horizon. Current data do not appear adequate to
justify alternatives to this assumption.
• The portion of the percolation flow that was in fractures is assumed to be available to
seep into drifts.
• The surfaces of waste packages contacted by seepage flow into the drifts are assumed
to be totally wetted. This assumption may well be overly conservative at low flow
rates.
• Seepage water that contacts and wets a waste package was assumed to enter the
package interior in proportion to the fraction of the waste package surface area that is
open as a result of corrosion.
• A seepage flow model was developed hi which, under current climate conditions, about
five percent of the waste package inventory would be contacted by seeps into the drifts,
and the seepage flow contacting each package would be on the order of 10-20 liters per
year. Under long-term average climate conditions, about 25 percent of the waste
packages would be contacted, and the seepage flow onto each waste package would be
on the order of 100-200 liters per year. It was stated in the TSPA-VA that there "...is a
great deal of uncertainty about seepage, particularly hi the fraction of waste packages
contacted by seepage." In addition, as discussed previously, there is recent evidence
that the threshold for seepage may be much higher (200 mm/yr) than the threshold used
in the TSPA-SR. Indeed, while the value of 200 mm/yr is treated as an extreme
minimum value in the TSPA-SR analysis, this value was obtained in field data for the
middle nonlithophysal unit of the Topopah Spring Tuff (DOE01, p. 4-92). Applying a
higher threshold value would significantly increase the amount of time before the
packages are wetted.
Within mis modeling framework, the assumptions concerning infiltration, percolation, and seepage
rates constitute conservative conditions based on currently available information. These
assumptions are likely to more strongly influence the timing of the release than the potential peak.
However, by delaying the release sufficiently, doses in the first 100,000 years may be dramatically
decreased using alternative assumptions.
The assumption that the entire surface of a waste package that is dripped on by seepage water is
wetted and therefore susceptible to aqueous corrosion is highly conservative. It is reasonable to
4-10 !
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expect that only water that drips onto a narrow band of the top of the package (e.g., at most a
20-degree arc of the 180-degree arc of the top half of the package) has real potential to initiate
aqueous corrosion. To totally wet the package surface, such drips (which could occur, according
to the seepage model, at a maximum rate of 10-20 liters per year under current climate conditions),
would have to spread uniformly over the package surface, which has a total area of about 40
square meters. This situation would produce a water film only about 0.2 millimeters thick, which
is an unrealistic condition to produce and sustain the Alloy 22 corrosion that is presumed to be the
mechanism for waste package failure.
4.3.4 Exposed Waste Form Area '
For commercial spent nuclear foel (CSNF) waste packages, which are the dominant (by two orders
of magnitude) source of radionuclide releases in the TSPA-VA analyses, the exposed waste form
area that can be a source of released radionuclides is related directly to the status and performance
of the CSNF cladding as a barrier. The TSPA-SR analyses assumed that eight percent of the
cladding will be failed at the time of emplacement owing to creep failure and stress corrosion
cracking. The TSPA-SR noted that "this mean percentage is very conservative and likely above
the amount of creep and SCC that the NRC will tolerate of operators of dry storage facilities." The
stainless-steel-clad rods were assumed to be distributed among the waste packages, and the entire
CSNF area in any failed rod was assumed to be exposed for contact with water. Zircaloy cladding
degradation by general corrosion and other means such as crushing by rockfall was assumed to be
a long-term phenomenon of no significance to 10,000-year dose estimates.
The assumptions concerning CSNF exposed area are highly conservative. Specifically:
• Stainless-steel-clad fuel rods will not be distributed throughout the waste packages
except as a result of deliberate effort. Less than one percent of the CSNF assemblies
have fuel rods with stainless-steel cladding, and they probably would actually be
disposed together in less than one percent of the total waste package inventory, in order
to reduce personnel exposures and operating costs.
• The estimate that eight percent of the Zircaloy-clad fuel rods are failed at the time of
emplacement is very conservative in comparison with available data. The observed
historical incidence of failure is less than 0.05 percent, is perhaps as low as 0.01
percent, and is confined to fuel manufactured in the early days of nuclear power or
subjected to external failure factors such as debris in the reactor coolant. Fuel yet to be
discharged from operating reactors (about 50 percent of the ultimate repository
inventory) can be expected to have no failures, so the incidence of at-emplacement rod
failure in the final repository inventory will be significantly less than the historical
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incidence to date and significantly less than the incidence assumed for the TSPA
analyses.
• In over 90 percent of the cases, "failure" of Zircaloy cladding has been found, in post-
service examinations, to consist of pinhole penetrations or hairline cracks. Therefore,
only a very small fraction of the fuel contained in a failed rod will be exposed as a
source of released radionuclides if contacted by water. In contrast, the TSPA
evaluations assumed that the entire spent fuel area in a fuel rod would be exposed for
contact with water and release of radionuclides. This assumption overstates the
exposed area, based on available data, by about three orders of magnitude.
• Many potential modes of Zircaloy cladding degradation, such as hydride formation and
creep failure, have been identified and characterized because cladding integrity is so
important in its reactor service conditions. EPA has performed and dpcumented a
comprehensive review and analysis of available information and has concluded that
degradation of cladding by any of the failure mechanisms is not expected to occur
under repository conditions after emplacement for disposal. The exposed waste form
area will therefore be that which exists at emplacement for disposal until very long-
term failures, such as package crashing by a rockfall, occur (SCA99). ;
Collectively, the TSPA-SR assumptions concerning exposed waste form area overstate the area
available for nuclide release by about four orders of magnitude (i.e., three orders of magnitude on
the exposed area per failed rod, and a factor often on the number of failed rods). They also
overstate the potential for long-term degradation of the cladding. If realistic assumptions
concerning performance of the EDA II repository are used, water would not contact the cladding
for more than 100,000 years, and cladding performance would be irrelevant to dose potential
before that time. Cladding performance will, however, be important to estimation of long-term
peak doses. In comparison with me preliminary estimate of peak dose of 85 mrem/yr at; 650,000
years for the EDA II repository (Table 4-1), a realistic estimate of cladding performance and
exposed waste form area would decrease the peak dose estimate by several orders of magnitude.
It is important to note that assumptions concerning cladding performance as a barrier and the,
amount of waste form area exposed for radionuclide release are essentially independent ;of
assumptions concerning performance of engineered features of the EDA II design. The link
between the EDA II design features and cladding performance is the design temperature limit for
the cladding. This limit is the same, 350°C, for both the VA and EDA II designs, and the expected
actual maximum cladding temperature in both designs is about 250°C. The 8% failure rate used in
the TSPA-SR was acknowledged to be very conservative. It represents a mean value for failure
rates at low (177-227 C) temperature. However, DOE also reports, the mode of this distribution as
about 2 percent. Hence, the mean appears to be skewed to a high value by a few outlier high
values (the maximum value is 19.4 percent). With "blending" of subassembly allocations to the
4-12 [
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waste packages in order to reduce thermal gradients, confidence in assumptions concerning
cladding performance that are less conservative than those used for the TSPA evaluations would
be increased.
In summary, the TSPA evaluations are highly conservative regarding cladding performance in
comparison with reasonable interpretations of the available information base. Assumptions of
exposed waste form area exposed for nuclide release for each failed fuel rod exceed actual exposed
areas by about three orders of magnitude; assumptions concerning the number of Zircaloy-clad
failed rods exceed the actual number by about a factor of 10. A realistic approach to these
assumptions based on principles of Reasonable Expectation is described in Section 5.
Nonetheless, despite these highly conservative assumptions, releases from the waste packages do
not occur within 10,000 years according to the TSPA-SR. Modification of these assumptions may,
however, improve long-term dose estimates for times greater than 10,000 years.
4.3.5 In-Package Dilution and Transport Delays
If water enters a penetrated waste package at the seepage rate or some fraction thereof, significant
delay could occur before the water contacts exposed CSNF and initiates radionuclide release. For
example, if the package interior fills slowly from the bottom up (as a result of trickle-down from a
small hole in the top and needs first to corrode through basket materials), and if the exposed CSNF
area(s) are in a subassembly near the top, thousands of years could pass before contact between the
water and the exposed waste form occurs.
Subsecpent to water/waste contact, released radionuclides that are mobile must be transported to
the point of exit from the package interior by advective and/or diffusional processes. By the time
release and transport occur, temperature gradients will be too low to drive significant advective
transport processes, and temperature levels will be too low for inside-to outside wall corrosion to
occur and to create an exit path at the bottom of the package. Consequently, radionuclide transport
rates will be low, the package interior will have to fill with water in order to enable radionuclides
to exit through the same penetration that provides water ingress, and the volume of water to fill the
package interior will be available to dilute the radionuclide concentrations.
The void volume of the interior of a 21-assembly PWR waste package is about 3,000 liters. If
water enters and exits the packages at rates in the range 6 to 400 liters per year, which corresponds
to the seepage rate range for the TSPA-SR long-term-average climate conditions, at steady state
and with complete in-package mixing, the in-package dilution factor would be in the range
3,000/400 ~ 7 to 3,000/6 = 500. Additional dilution would men occur during transit of the near-
4-13
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field, the unsaturated zone, and the SZ by the contaminated water that exits the package.; Such
dilution mechanisms may be particularly important for radionuclides not limited by their; elemental
solubilities, such as 1-129. '
These in-package delay and dilution possibilities were considered and analyzed in the studies
described in the TSPA-VA Technical Basis Document (DOE98a). They were not, however,
included in the base-case performance assessment models for TSPA-VA and TSPA-SR because of
uncertainties, and an inability to justify the assumptions. If included in the models, they could
have reduced the predicted TSPA-VA 10,000-year dose by one to two orders of magnitude,
depending on how probabilities for the relevant performance factors are taken into account,
i
To incorporate these performance factors into the TSPA models, it would have been necessary to
develop probability distributions for factors such as time elapsed between water entry to;the
package interior and time of water contact with the exposed waste form. DOE chose to develop
probability distributions many of the performance factors external to the packages, but chose to
omit the in-package performance factors and associated probability distributions, from the TSPA
models. It is worth noting that the in-package performance factors are potentially as important to
the TSPA results as climate change and seepage rate. It is reasonable, for example, to expect, at a
minimum, some degree of dilution of contaminant concentrations in water exiting a waste package
as a result of mixing with nearby water in the near field and the UZ.
As for the role of exposed waste form area in performance of the EDA II repository, the in-
package performance factors will not be important to evaluation of 10.,000-year doses if realistic
assumptions concerning performance of the EDA II drip shields and waste packages are1 used, such
as has been done in TSPA-SR. The in-package performance factors could, however, help to show
that long-term peak doses will be low in the period after 10,000 years. Specifically, penetration of
the EDA II waste packages will occur so far into the future that there will be virtually no thermal
driving force for radionueh'de release and transport in the waste package interior. Mixing and
homogenization of concentrations within the waste package would have to be driven by diffusional
processes, ;
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4,4 Radiation Doses to Alternative Acceptors
To date, eleven alternative dose receptors have been identified by DOE, the NAS, NRC, EPRI, and
EPA as the potential basis for evaluating compliance with the individual-protection standards for
Yucca Mountain. The options include alternatively-characterized individuals and critical groups.
Each has, to some degree, taken cognizance of site-specific conditions and each has, to some
degree, utilized ICRP principles for designating a group or individual receptor, e.g., to base
assumption of the future receptor's habits on present-day habits. Each also seeks to identify and
characterize the receptor with the highest dose potential, without being extreme, in order to assure
protection of other individuals.
DOE's TSPA-VA, DEIS, and TSPA- SR used the so-called "average resident" as the dose
receptor. This individual was located 20 km from the repository, and had habits corresponding to
those of current residents, as determined in a survey performed by DOE. The TSPA-VA states
that this person consumes part of Ms food from local sources and consumes 1.8 liters per day of
drinking water contaminated with radionuclides released from the repository, at the maximum
contaminant plume concentration. The DEIS, which was stated to use the same TSPA evaluation
methodology as the VA, states that the average resident receptor consumes 2.0 liters of
contaminated water per day. The TSPA-SR states that the average-resident receptor hi Amargosa
Valley consumes slightly more than 2.0 liters of water per day (753 liters per year), and this value
is used in me assessment. With this water consumption rate, the DOE's average-resident is
essentially equivalent to EPA's "rural residential" RMEI dose receptor.
Results of the DOE's average-resident dose evaluations at 10,000 years, based on the assumptions
and methods described above in Section 3.7, can be summarized as follows:
• TSPA-SR mean all-pathways dose (using probabilistic evaluations) is 0.10 mrem/yr.
This value is the same as the mean value at 10,000 years for the TSPA-VA: However,
this agreement is fortuitous, as entirely different scenarios are associated with the dose.
For the TSPA-VA, the dose was associated with juvenile failures of waste packages
that were unrealistic and gave high releases. The TSPA-SR treats these juvenile
failures more realistically, and these failures do not affect pre-10,000 year doses in the
TSPA-SR. By contrast, in the TSPA-SR the dose at 10,000 years is associated with
igneous intrusion, with subsequent releases to ground water. This scenario appears to
be highly conservative, giving unrealistically high releases as well.
• VA base-case dose using an evaluation with all parameters set at their expected values:
0.04 mrem/yr. A similar result is not reported for TSPA-SR. Instead, the range of
doses at 10,000 years is about lO^-lO0 mrem/yr (5* to 95th percentiles), with the mean
about 10"' mrem/yr and the median about 10"2 mrem/yr.
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• DEIS mean all-pathways dose for the Proposed Action, which corresponds to the VA
repository: 0.22 mrem/yr (this result presumably differs from the VA result of
0.10 mrem/yr because of modeling adjustments made for the DEIS evaluations in order
to be able to address the DEIS options concerning waste inventories and thermal
loadings). :
Since the DOE's average resident corresponds to EPA's rural residential RMEI, these results are
representative of the results that would be obtained using the EPA's rural-residential RMEI at 18
km as the receptor and the TSPA-SR methodology and assumptions. As previously noted, and
discussed in Section 3.7, these results overstate the dose to be expected as a result of the
conservative assumptions used in the evaluations. ;
The DEIS also evaluated doses to the average resident at alternative locations and for the
alternative areal mass loadings considered. Results are shown graphically in Figure 4-4;
corresponding Tc-99 concentrations in ground water, and assumed saturated-zone dilution factors
at each distance, are shown in Figure 4-5. Variations of 1-129 concentration with location and
areal mass loading are similar to those for Tc-99, but 1-129 concentration levels are about two
orders of magnitude less than those of Tc-99. ;
50
O
n.
5
40-1
30-
0)
g 20
o
O
«
a>
6 10
(5.15)
10
15
20
25
!30
Km
* The MCL for Tc-99 is 900 pCi/1. Numbers in parentheses are the dilution factors used at each distance.
Figure 4-5. Tc-99 Concentrations for Alternative Mass Loadings (compiled from DOE99)
4-16
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The reason for dose variation with areal mass loading is n6t evident from available documentation.
For example, repository temperatures, which would affect corrosion rates, are virtually identical
for the 85 and 60 MTU/acre loadings during the period from 1,000 to 10,000 years, while the
repository temperatures for the 25 MTU/acre loading are about 40°C less (DOE99). Similarity of
results for the 85 and 60 MTU/acre cases might therefore be expected; Figures 4-3 and 4-4 show,
however, that the results for 60 and 25 MTU/acre are most similar. Also, Figure 4-4 does not
show any correlation between SZ dilution factors and distance from the repository.
The variations of dose with areal mass loading may be the result of differences in repository areas
and attendant differences in transport and dilution in the unsaturated zone. The 85, 60, and 25
MTU/acre repositories, for the reference inventory of 70,000 MTU of wastes, occupy 740, 1050,
and 2,520 acres, respectively. The 60 MTU/acre repository occupies two emplacement blocks and
the 25 MTU/acre repository is spread over several emplacement blocks. Transport and dilution in
the UZ may therefore have been modeled differently for the three loading options.
While the DEIS evaluated doses for the same receptor at alternative locations and for alternative
repositories, the VA characterized doses for alternative receptors. The VA reported dose
evaluation results for only the average resident as receptor, but it also characterized doses for a
subsistence farmer receptor and a so-called residential farmer receptor. All food and water
ingested by the subsistence farmer was assumed to be contaminated; only part of the fobd
consumed by the residential farmer was assumed to be contaminated. DOE's surveys found no
current residents who correspond to either of these receptors. The characterizations determined,
however, that the Np-237 biosphere dose conversion factor (BDCF) for the residential farmer
would be three times greater than that for the average resident, and the BDCF for the subsistence
farmer would be about six times greater. The 1-129 BDCF for the subsistence farmer was stated to
be about 10 times greater than that for the average resident. The VA also stated that the most
important factor for doses due to 1-129 and Tc-99, which are the only radionuclides of significance
released in the 10,000-year time frame, is leafy vegetable consumption, and that direct;
consumption of contaminated ground water contributes about 50 percent of the dose.
The NRC defines a critical group as the dose receptor in the proposed 10 CFR Part 63 regulations
for Yucca Mountain. The critical group is described as residing within a fanning community
located approximately 20 km south of Yucca Mountain. Members of the group would have
characteristics that are consistent with current conditions and that result in the highest expected
annual doses. The group would be a farming community of up to 100 individuals residing on 15 to
25 farms. The behaviors and characteristics of the average member of the critical group would be
based on the mean value of the group's variability range.
4-17
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The average member of the NRC's critical group would be predicted to incur less dose than either
the DOE average resident or the EPA's rural residential RMEI, and this choice of dose receptor
would therefore be less protective of the general population. Less dose would be predicted for two
principal reasons: the dose conversion factors for the NRC critical group would be based on mean
values of the dose factors, whereas the EPA's RMEI uses maximum values for one or more of the
dose factors (e.g., drinking water rate); and the members of the fanning critical group would be so
spread out that only a fraction of the group would use contaminated water at the maximum
concentration. The current average size of an alfalfa farm, which is the dominant farming activity,
is about 255 acres; in the most compact configuration, a square, 25 farms of current average size
would occupy an area more than three miles long on each side. The VA shows (p. 3-137 of
Volume 3) that the contaminant plume width is only about 1 mile at 20 km distance from the
repository. Many of the members of the NRC's critical group would therefore, in reality, receive
no dose or significantly less man the maximum dose, so that the average would be unrealistically
low. Ground water flow systems dominated by fractured rock hydrology would be expected to
produce narrow contamination plumes (see the BID for descriptions of the fracture-flow
dominated hydrologicai system at the Yucca Mountain site.)
If 25 average-size alfalfa farms are located 20 km from the repository (e.g., at Lathrop Wells), the
number of farms that intercept the plume at that distance will depend on how the farms are located
relative to each other. If the farms are in an east-west line, only one farm would intercept the
plume. If the farms are adjacent to each other in a square, at most five farms would intercept the
plume. If the farms are in a north-south line, some of the farms would extend beyond 30 km from
the repository, i.e., beyond the current Amargosa Farms area (SCAOO). .
In summary, the dose receptors considered in DOE's TSPA-VA and TSPA-SR are similar to the
EPA's RMEI as described in the rule, and may actually be somewhat more conservative. For
instance, the TSPA-SR assumes slightly more than the 2 liters/day drinking water consumption
specified in the rule. In addition to these earlier treatments, the critical group receptors 'evaluated
in the TSPA-SR are subject to exposures to contaminated ash in the eruption scenario. Dose
estimates in both of these TSPAs are well below the 15 mrem/yr individual protection limit,
despite the use of very conservative assessment scenarios and models. Based on these :
considerations, the EPA's choice of an RMEI rather than a critical group for the dose receptor does
not have any impact on repository development costs or progress. As described above, the
proposed farming community critical group potentially makes assessment defensibility more
difficult and subject to challenge, owing to the requirement for arbitrary assumptions on the size
and location of farms. These may not necessarily be consistent with current and projected land use
in the area, so as to ensure that all members of the critical group receive some level of exposure.
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4.5 Alternative Means to Reduce Uncertainties and boses
As noted in Section 3.4.1, principal objectives in selecting the EDA II design as the basis for the
site recommendation were to improve the real performance potential of the repository and to
reduce uncertainties in projections of performance. The benefits of the EDA II design are
illustrated in the differences between dose projections in the TSPA-VA and the TSPA-SR, which
shows that projected doses for the EDA II repository up to 10,000 years are substantially less than
those for the VA repository. Indeed, the EDA II design only produces doses in the first 10,000
years as a result of potential igneous activity. In terms of performance for the nominal behavior of
the system, the improvement in performance over 10,000 years is extremely dramatic. As shown
in Figure 3-2, the nominal case for the EDA II analysis exhibits no releases over 10,000 years.
Overall, it can be said that the objective of the EDA II design is to defer and reduce the potential
for, and uncertainties in, thermally driven degradation processes such as corrosion and advective
radionuclide transport. Alternatives to the EDA II design that address this objective are illustrated
by the EDA options considered, from which the EDA II option was selected for the TSPA-SR
(Table 3-3). Comparison of these options shows that they reflect widely different strategies for
meeting the objective. For example, the EDA I option takes a direct approach by reducing the area
mass loading and repository temperatures. The EDA V design takes the opposite approach: it
drives 'the temperatures to high levels in order to greatly defer the time at which water can enter
the repository and initiate high-rate degradation processes.
Other advanced repository designs which incrementally improve the VA design might have been
identified and evaluated. For example, the waste package design with the Alloy 22 on the outside
could have been adopted with all other EDA II parameters except use of drip shields and backfill.
This choice would have considerably increased waste package performance by eliminating the
crevice corrosion process that greatly accelerated package failures in the VA design (17 packages
failed by this mechanism within 10,000 years), thereby extending expected waste package
lifetimes beyond 10,000 years. Another incremental design feature that could be added would be
to tilt the packages along the axis at emplacement in order to have drips run off the surface, or to
use weld shields rather than drip shields that cover the entire package. These are simple,
inexpensive design features that could reduce the potential for juvenile failure and subsequent
releases.
With respect to the impact of juvenile waste package failures, their treatment in the TSPA-VA was
extremely conservative and consequently releases from such failures dominated doses within the
10,000 year period. In the TSPA-VA, an exit hole in a waste package was assumed to exist as
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soon as an entry hole was created. Under this assumption a juvenile failure from a manufacturing
defect (weld failure) resulted in immediate releases into the waste package surroundings. In the
TSPA-SR, a more realistic treatment of juvenile failures was incorporated, eliminating the extreme
conservatism of the TSPA-VA treatment. These modeling changes, along with the move to an
improved waste package design that eliminated the potential for accelerated failure associated with
crevice corrosion, were sufficient to greatly improve the projected performance. Improving
performance projections and reducing uncertainties could be done in a variety of ways, with cost
impacts that vary according to the extent and nature of changes in the repository and waste
package design, and according to increases in data needs for the assessment of performance.
Analysis results for the EDA options that are presented in Table 3-3 show that the optiqns meet the
objective to varying degrees and with different costs. In exarnining the performance factor results
in Table 3-3, it is important to remember that these results were produced using the same
performance models and conservative assumptions that were used to produce the TSPA-VA
results. More realistic evaluations, using reasonable parameter values, models, and assumptions,
would produce peak annual doses at least several orders of magnitude less than those shown in the
Table. Realistic evaluations and assumptions that would lead to lower doses are discussed in
Section 5, which addresses Reasonable Expectation.
To paint a realistic picture of repository performance potential, it is important to acknowledge the
benefits of the design features in the models and assumptions used to make performance
predictions. For example, the backfill/drip shield/waste package design features of the EDA II
repository completely eliminate the real potential for juvenile waste package failures or corrosion-
related radionuclide releases for 10,000 years and more. Similarly, assessments of long-term peak
dose potential that use reasonably-expected parameter values and assumptions will show dose
levels that are orders of magnitude less than those that have been reported to date, even; without
including performance factors such as in-package dilution that have been omitted from the model
hierarchy to date.
In summary, the EDA II repository design, which is the basis for the TSPA-SR, is a highly
conservative design with extensive redundancies that assure no radionuclide releases in the
nominal scenario during 10,000 years. The design has enabled modifications to the TSPA models
and assumptions that reflect the benefits of the design to repository system performance. The use
of this robust design has allowed DOE to use a number of very conservative assumptions in its
assessment. Modification of these assumptions to more reasonable, yet still credible, approaches
would result in very significant delays in releases from the repository; there is the potential that
modified assumptions would produce no significant calculated doses in the first 100,000 years.
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4.6 Current Repository Design and Safety Strategy"
As part of its program evolution to the TSPA-SR, DOE has recently revised its Repository Safety
Strategy previously described in 1998 (DOE98d). The most recent description of the revised
strategy and plans for its implementation is provided in TRWOO.
The TRWOO Rev 3 strategy updates the previous version of the strategy (DOE98) which was the
basis for the VA. It reflects the EDA II design (Section 3 of this document), which is the current
stage of evolution of the repository design. The revised strategy also reflects recent additions to
the program database; response to the regulatory framework; and internal and external comments
on the VA design and TSPA-VA methodology, and the eventual implementation as TSPA-SR.
Under this strategy, the postclosure safety case is based on developments and evaluations in five
principal areas: performance assessment; safety margin and defense-in-depth; consideration of
potentially disruptive processes and events; insights from natural analogs; and long-term
performance confirmation.
The design evolution (from the VA to EDA II) and the safety strategy evolution are intended to be
responsive to the concerns about uncertainties and technical issues associated with the TSPA
methodology and assumptions as it evolved from TSPA-VA to TSPA-SR. The approach will
reduce potential difficulties during licensing reviews by reducing or eliminating the TSPA
uncertainties and issues that would create difficulties in licensing reviews.
The EDA II design and the Rev 4 Safety Strategy are the latest step in evolution of the repository
concept. Over time, as shown in Table 4-3, the relative contributions of the engineered and natural
system features of the repository to overall performance have inverted: site characterization has
shown that the natural features will not contribute nearly as well to performance as was expected
in the SCP, and the performance of the engineered barriers has been increased dramatically to
compensate for the lesser natural barrier performance expectations, and to respond to licensing
requirements for defense-in-depth and minimization of uncertainties and technical issues.
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Table 4-2. Change Over Time of the Roles of Natural and Engineered Earners
in Repository System Performance
Project Era
SCP (1988)
Early TSPAs
(1991-1995)
Viability
Assessment
(1998)
TSPA-SR(2001)
Infiltration Rate
1 mm/yrmax
1 mm/yrmax
8-40 mm/yr now;
200 in supeipluvial
0.4-12 mm/yr now
4.7-20 in monsoon
climate
GWTT* .:•%¥:
9K-80Kyrs
9K-80Kyrs
As short as 50 yrs for
fast paths in the UZ
Mean delay in UZ is
1,000 years and mean
delay in SZ is 1,300
years
; WP** Lifetime
300 -1,000 yrs
Various designs, 300
-10,000 yrs
Less than 20
packages fail within
10,000 years; 20,000
years for general
corrosion
expect no radio-
nuclide release for
10,000 + years
:•• ffiSS Features >"•
Thin-walled can,
vertical in floor
Horizontal, robust
WP considered
Horizontal WP used
steel over Alloy 22
Alloy 22! on outside
of WP; add drip
shields
* GWTT « Ground Water Travel Time to the Accessible Environment
** WP = Waste Package
Another facet of the Safety Strategy has been an extensive evaluation of parameter uncertainty and
sensitivity. The TSPA-SR (TRWOOa) reported three kinds of evaluations of parameter uncertainty
and sensitivity: Uncertainty Importance Analysis, Sensitivity Analysis, and Robustness Analysis.
Uncertainty Importance Analysis refers to the use of regression analyses to determine the most
important parameter contributors to the spread of output results, and classification-tree analyses to
determine the parameters leading to extreme outcomes in the distributions. Sensitivity Analysis
refers to single-parameter sensitivity analyses, in which one parameter is varied while the others
are held at particular values. Robustness Analysis (also referred to as Degraded Barrier Analysis in
the TSPA-SR) refers to a focused approach to examining parameters associated with extreme
degradation behavior of individual barriers, keeping intact the remaining analysis of the1 system.
Uncertainty importance analyses were performed beginning with stepwise linear rank regression
analysis. The results of these analyses were evaluated using classification and regression tree
analysis to determine decision rules that determine whether a particular realization would produce
doses at me upper or lower end of the output distribution. These approaches were used to evaluate
the spread in doses at a particular time and the spread of times needed to produce a particular dose.
Particular attention was also focused on the extreme high end of the output distribution, to
determine which parameters lead to the extremes of the output.
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The uncertainty importance analyses showed that the waite package and saturated zone processes
are the most important factors in the nominal scenario, whereas the probability of the occurrence
of igneous disruption of the repository is the most important factor for igneous scenarios. As
discussed in the TSPA-SR, the assessment that these are the "most important" in this uncertainty
importance analysis reflects two factors: the change in variance of dose rate with variance of the
parameter, and the change of the dose rate itself with changes in the parameter. If either of these
two derivatives is small, the techniques used in the TSPA-SR will tend to show the parameter to be
unimportant.
Sensitivity analyses, as used in the TSPA-SR, refer to a single parameter variation method. This is
considered to be a complementary technique to the uncertainty importance analysis. In this
approach, a single parameter was ranged between its 5th and 95th percentiles, and other •
parameters were fixed at particular values.
The robustness analyses were conducted by setting a suite of parameters associated with a
particular barrier at their 5th or 95th pereentile, whichever tends to maximize the dose rate over
the time period of interest. For the sake of completeness, the results are also shown compared to
results from the same suite of parameters set at the opposite end of the behavior (i.e., values which
tend to minimize dose consequences). The intent of these robustness analyses is to present the
behavior of the system as a whole if any part of the system degrades quickly, and functions
according to its extreme behavior. Robustness analyses were conducted on nine facets of system
behavior (TRWOOa):
» UZ. This barrier represents the function of the UZ above the potential repository in
limiting the amount of water that reaches the potential repository. This barrier includes
the climatic conditions at Yucca Mountain, the processes at and near the surface that
lead to infiltration, and flow through the UZ above the potential repository. Parameters
treated in the robustness analysis were the seepage-uncertainty factor and the flow-
focusing factor. Degraded conditions for these parameters resulted in a small increase
in dose rate over the base case.
• Seepage into emplacement drifts. This barrier represents the function of the drifts
themselves as a capillary barrier that limits the amount of water that enters the drifts.
Both infiltration and seepage parameters were set to their degraded behavior for this
analysis. Degraded conditions for these parameters resulted only in about a factor of 5
increase in dose rate over the base case.
* Drip shield. The first of the engineered barriers, the drip shield limits the amount of
water that reaches the waste package. In the robustness analysis, the general corrosion
rate parameters were set to their extreme values. While the drip-shield lifetime is
significantly degraded in this analysis, there is almost no change in the dose rate. This
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results reflects the fact that the waste package degradation model is independent of the
drip shield function. This appears to be an example where the high degree of
conservatism in one model masks the importance of a different function, as discussed in
TRWOOa. [
Waste package. The primary engineered barrier, the waste package limits the amount
of water that reaches the waste form and limits radionuclide transport out of the BBS.
Degradation parameters considered in the robustness analysis were: residual hoop-
stress state and stress intensity factor at the closure-lid welds; number of manufacturing
defects at the closure-lid welds per waste package; Alloy-22 general corrosion rate;
microbially-induced corrosion enhancement factor for general corrosion; and
enhancement factor for Alloy-22 general corrosion from aging and phase stability. The
enhanced case (optimistic parameters) led to no releases from the waste package for the
first 100,000 years. The degraded parameters show a somewhat earlier failure profile,
with first failure occurring at 7,000 years compared to 12,000 years for the base case.
For the degraded case there is 50 percent probability that 1 percent of waste packages
fail at about 10,000 years and 10 percent of waste packages fail at about 12,000 years.
For the base case it is about 25,000 years for the 1 percent failure and about 50,000
years for the 10 percent failure. Accordingly, the predicted mean dose starts | earlier
(about 8,200 years versus about 15,000 for the base case), and the predicted mean dose
rates are much higher. ;
CSNF cladding. The Zircaloy cladding is an engineered barrier that is part of the waste
form. It limits the amount of water that reaches the CSNF portion of the waste and
limits radionuclide transport out of the CSNF waste form. (CSNF is planned to be
approximately 90 percent of the mass of waste in the potential repository.) Four of the
five parameters in the cladding degradation model were evaluated in the robustness
analysis: the number of rods initially perforated in a CSNF waste package, the
uncertainty hi localized corrosion rate, the uncertainty of the CSNF degradation rate,
and the uncertainty in the unzipping velocity of the cladding. It was concluded that
these parameters are unimportant for performance in the first 100,000 years, :but that
they contribute to the spread of doses during the period 100,000-1,000,000 years. The
effect of these parameters on dose rate in the robustness analysis is not reporjted by
TRWOOa.
Concentration limits. This barrier represents the function of environmental conditions
and radionuclide solubility limits in limiting radionuclide transport out of the BBS.
The primary dose contributor in the first 30,000 years is technetium-99. The solubility
of Tc-99 is assumed to be large (1 M), and is not treated as uncertain. The primary
radioelements for the period after 30,000 years are neptunium, americium, and
uranium. The solubilities of each of these is controlled by pH in the TSPA-SR model.
The pH, in turn, is assumed to not vary widely in the invert. This limits the variability
of the dose rate as a function of any other factors in the near-field model, ^particular,
TRWOOa notes that most of the releases are by a diffusive mechanism, hence controlled
by diffusion-related parameters. This too appears to be an area in which a strong
structural conservatism of the model (in this case the assumed diffusional releases) tend
to mask the importance of other effects.
4-24 !
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* BBS transport. This barrier represents the function 'of environmental conditions and
diffusion in the drift invert in limiting radionuclide transport out of the BBS. In this
case of the robustness analysis, the combined effects of degraded concentration limits
and high diffusion cases. The results are reported as a decrease in the time to early-
arrival doses (defined as time to 10"3 mrem/yr) of several thousand years, and an
increase in the peak dose rate of about a factor of five.
» UZ transport. This barrier represents the function of the UZ below the potential
repository in delaying radionuclide transport to the biosphere. An extensive set of
robustness analyses were presented for this function. The degraded cases showed
between a factor of 5-10 higher dose rates than the base case, whereas the enhanced
cases showed significantly improved behavior (many orders of magnitude) over the
base case. That is, since the base case is little different than the degraded case but very
different than the enhanced case, this means that the base case is strongly biased toward
the conservative end of the spectrum of behaviors.
* SZ. This barrier represents the function of the SZ in delaying radionuclide transport to
the biosphere. The robustness analysis was used to investigate parameters associated
with travel time in the saturated zone: sorption, and flow rate. The difference between
degraded and enhanced performance in these analyses is between one to two orders of
magnitude, with the base case very close to the upper end of this variability. Again,
this indicates a strong bias toward conservatism in the base case.
The TSPA-SR explicitly acknowledges that the results of these analyses are dependent upon the
scenarios and conceptual models implemented in the TSPA-SR. They note that the conservatism
of parameter values and assumptions may tend to mask the importance of some of these to the
results, or may mask the importance of others. Two of these situations were noted above in the
discussion of robustness analysis: the conservatism of the drip shield treatment masks the
importance of the waste package behavior, and the assumption that diffusion dominates releases
together with an assumption of high solubilities tends to mask the importance of other phenomena
in the waste package. These assumptions therefore compound the conservatism of the analysis,
since they are, by themselves, conservative, and they also minimize the functional importance of
other barriers. ;
The strong reliance on evaluations of parameter sensitivity and uncertainty analyses skews the
evaluation of the TSPA-SR results. Instead, the model is in some cases so structurally biased
toward conservatism that appropriate conclusions cannot be drawn. For instance, one conclusion
of the TSPA-SR is that the primary factor influencing the consequences of the igneous scenarios is
its probability of occurrence. All other parameters investigated in the sensitivity analysis were
found to have relatively minor influence on the dose from igneous disruption. However, as
discussed previously, such a conclusion ignores the heavy conservative bias of the consequence
modeling assumptions. Given the extremely conservative basic assumptions of the consequence
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model, one would not expect parameter variations to significantly affect the results. By; contrast,
changes in the basic assumptions about interaction of magma and waste containers could decrease
releases and their associated doses by orders of magnitude, or eliminate them altogether.
Similarly, the use of a model for release from the waste package in the nominal scenario that
assumes diffusion in the absence of significant amounts of water near the package, results in a
significant conservative bias. This assumption and associated model masks the importance and
utility of the presence of the drip shield. The lack of significance of the drip shield in the TSPA-
SR nominal case is therefore seen to be an artifact of the conservative bias of the waste package
release model, rather than a fundamental property of the repository.
The reliance on evaluations of parameter uncertainly illustrates (potentially deceptively) small
uncertainties in relatively high consequences associated with the repository. Uncertainties in
parameters, as shown by the robustness analyses, lead to at most about an order of magnitude
increase in dose rate under unfavorable conditions. Application of favorable sets of parameters
were shown to potentially decrease the dose rate by several orders of magnitude in some cases, and
to push the doses out to much longer times, in some eases past 100,000 years. By contrast,
uncertainties in assumptions (conceptual model uncertainty) have the potential to lead to dramatic
improvements in consequence analyses. Alternative conceptual models for the igneous scenarios
have the potential to lead to minimal or zero releases from these effects, thus eliminating the
consequences associated with igneous activity. Alternative conceptual models for the waste
package in the nominal scenario would likely show early releases at much later times, perhaps
with minimal release in the first 100,000 years. In addition, me use of less extreme assumptions
may lead to a better understanding of the effects of design features such as the drip shields.
Consideration of these less conservative, yet defensible and physically realistic, models is
consistent with the principles of Reasonable Expectation (see Chapter 5), as well as with the
concept and intent of Importance Analysis (KOZ97).
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5.0 EPA'S "REASONABLE EXPECTATION" APPROACH TO REPOSITORY
PERFORMANCE PROJECTIONS
This chapter discusses reasonable expectation and reasonable assurance as concepts to be used
in implementing the standards. We believe the reasonable expectation approach is more
appropriate for repository compliance determinations and provides a more realistic link between
design and anticipated performance in the iterative process of developing a repository design for
licensing.
5.1 Overview of Reasonable Expectation
The impact of the EPA standards on repository design and data collection is complicated by the
fact that NRC will adopt and implement the standards, as mandated by the NWPA. The NRC is
therefore the agency that determines what is needed to comply with the EPA standards. The
method of implementation of the standards then becomes a deciding factor in evaluation of
compliance. This chapter discusses the issue of compliance methodology, i,e., reasonable
expectation versus reasonable assurance.
The EPA standards call for use of "reasonable expectation", rather than "reasonable assurance," as
a basis for assuring compliance with the EPA standards. Reasonable expectation and reasonable
assurance are both compliance assessment approaches and can be distinguished as discussed
below. In brief, the intent of reasonable expectation is to recognize the inherent uncertainties
involved in repository safety performance evaluations, and to encourage realistic treatment of the
uncertainties in performance assessments and evaluations of compliance with the disposal
standards. Reasonable expectation takes what might be termed a realistic or best-value approach
to dealing with uncertainty in performance projections when compliance issues are complicated by
uncertainties imposed by extrapolations of data and projections of performance over long time
periods. Reasonable assurance is a concept that has been used in the licensing of facilities which
involve only short term extrapolations of performance.
In developing a repository design, there is an iterative process between design and performance
assessment that evolves over time to a final design and compliance calculations that are presented
for licensing. A process that recognizes and deals realistically with inherent uncertainties would
offer an efficient approach to optimizing design and performance.
The 40 CFR Part 197 standards require that DOE demonstrate compliance with the individual-
protection, human-intrusion, and ground water protection standards under principles of
"reasonable expectation." The standard states, at §197.14, that reasonable expectation requires .
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less than absolute proof, because absolute proof is impossible to attain for disposal due to the
inherent uncertainly in projections of long-term performance. The rule also states that Reasonable
Expectation (RE) focuses performance assessments and analyses upon the full range of defensible
and reasonable parameter distributions rather than only upon extreme physical situations and
parameter values.
The Preamble to the proposed 40 CFR Part 197 (EPA99) described RE and its use as follows:
In carrying out performance assessments under a "reasonable expectation "
approach, all parameters that significantly affect performance would be identified
and included in the assessments. The distribution of values for these parameters
would be made to the limits of confidence possible for the expected conditions in
the natural and engineered barriers and the inherent uncertainties involved in
estimating those values. Selecting parameter values for quantitative performance
assessments would focus upon the full range of defensible and reasonable
parameter distributions rather than focusing only upon the tails of the distributions
as is more commonly done under the "reasonable assurance " approach. The
"reasonable expectation" approach also would not exclude important parameters
from the assessments because they are difficult to quantify to a high degree of
confidence.
5.2 Prior Consideration and Use of Reasonable Expectation
Reasonable expectation is the basis for evaluation of compliance with the Subpart B and C
standards in EPA's 40 CFR Part 191 (amended at 58 FR 66414, Dec. 20,1993), and is
implemented in 40 CFR Part 194, the criteria for certification of WIPP (61 FR 5224, February 9,
1996). Use of the concept was upheld by the U.S. Court of Appeals, First Circuit, in its decision
concerning the suits brought against the EPA for the 40 CFR Part 191 standards issued in 1985.
The Court stated, in its decision:
Given that absolute proof of compliance is impossible to predict because of the
inherent uncertainties, we find that the Agency's decision to require "reasonable
expectation " of compliance is a rational one. It would be irrational for the Agency
to require proof which is scientifically impossible to obtain. Any such purported
absolute proof would be of questionable veracity, and thus of little value to the
implementing agencies. Nor can we say that this provision is arbitrary and
capricious because it will afford the implementing agencies a degree of discretion,
since such imprecision is unavoidable given the current state of scientific
knowledge. Thus we are again faced with a decision that is within the Agency's
area of expertise and on the frontiers of science, and, as such, we refuse to
substitute our judgment for that of the Agency. (824 F.2d 1258 (1st Cir. 1987, at
page 1293)).
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Hie use of reasonable expectation is the same"'in 40 CFR Part 191 and 197. Part 191 states, at
§191.15, Individual-Protection Requirements:
Disposal systems for waste and any associated radioactive material shall be
designed to provide a reasonable expectation that, for 10,000 years after disposal,
undisturbed performance of the disposal system shall not cause the annual
committed effective dose, through all potential pathways from the disposal system
to any member of the public in the accessible environment to exceed 15 millirems
(150 microsieverts).
The uidividual-protection standard for Yucca Mountain is stated, in §197.20, as:
The DOE must demonstrate, using performance assessment, that there is a
reasonable expectation that, for 10,000 years following disposal, the reasonably
maximally exposed individual receives no more than an annual committed effective
dose equivalent of 150 microsieverts (15 millirem) from releases from the
undisturbed Yucca Mountain disposal system. The DOE's analysis must include all
potential pathways ofradionuclide transport and exposure.
5.3 Comparison of Reasonable Expectation and Reasonable Assurance
.Reasonable expectation can be compared to reasonable assurance, used by the NRC in licensing of
nuclear power reactors and other engineered fuel cycle facilities. In engineered facilities licensed
by the NRC, parameter values usually lie within a narrow range around an expected value which is
well known as a result of testing and experience, and the range itself will be based on actual testing
and experience. For example, testing multiple samples of an alloy to measure the brittle fracture
strength will result in a mean value with a small range of variability.
For reactors, the projected performance of engineered components of the facilities can be verified
during their in-service lifetimes, which are only a few decades long. Consequently, the
extrapolation of laboratory testing results over the relatively short reactor operating lifetime allows
confirmation of the projections. This "real time" verification has been a part of the licensing
experience for power reactors. Extrapolation of important natural processes in reactor licensing is
limited to predictions of seismic hazards which in practice is done only for short periods of
decades.
In contrast, repository performance projections involve the extrapolation of natural processes and
events, and laboratory performance testing of engineered materials over time periods of
10,000 years and beyond. Such extrapolations have to date been applied only to WIPP in EPA's
certification of that disposal facility.
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All engineered elements of a reactor are subject to performance verification, integrity of welds can
be confirmed, quality of construction can be verified, and training of personnel can be confirmed.
The NRC can, therefore, establish a measured pedigree for every factor important to system
performance and can expect and require, to a very high degree of assurance, that the facility will
operate as intended and expected. Principles and methods of reasonable assurance were developed
to serve these circumstances. Transferring reactor licensing experience and expectations unaltered
into regulatory decision making for deep geological disposal is not an appropriate adoption of
reasonable assurance used for licensing of reactors and other fuel cycle facilities. In adapting the
reactor-based reasonable assurance to the geologic repository application, NRC has adopted a
weighted probabiEstic approach to evaluate performance projections. This approach moves
significantly toward a recognition of the inherent differences between reactor licensing and deep
geological disposal (e.g., the difficulty in verifying long time frame performance projections).
However, a probabilistic approach does not, by itself, unequivocally guarantee that repository
performance projections will appropriately incorporate the inherent uncertainties in these
projections in a way that is not excessively conservative.
In contrast to reasonable assurance, reasonable expectation takes into account, for long-term, deep
geologic disposal, the fact that many relevant performance parameters cannot be clearly:
characterized as can those for an engineered facility with a forty-year lifetime. Specifically, many
natural features important to repository performance cannot be extensively characterized, and
many exhibit a high degree of inherent variability. In addition, performance characteristics of
engineered features of the repository must be extrapolated well beyond the time period for which
measurements can be made.
For example, ground water flow in the volcanic rocks in the vicinity of Yucca Mountain will occur
primarily in fractures which have highly variable physical characteristics such as width, length,
and connections to other fractures. Tests can establish characteristics of fractures for locations
where the testing was done, but testing at various locations will produce different results, which
can vary widely. (Reflecting these variations, yield from water supply wells located in fractured
rocks can vary widely over short distances.) In addition, the hydrological behavior of a.fractured
rock system can change over time, as tectonic processes and seismic activity readjust the stress
state in the area. Fracture networks could be enlarged, and their connectivity and flow behavior
could be gradually altered either favorably or unfavorably over long time periods. In aggregate,
thorough test results will produce a picture of what is a reasonable interpretation of the range of
results, and this would be the basis for implementation of the concept of reasonable expectation. It
1
would not be reasonable to base performance assessment models and parameter values only OE
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results which show limited fractures and HmitSd flow, or dfa results which show extensive
fracturing and high flow rates.
A specific example for Yucca Mountain is the case of so-called bomb-pulse Cl-36 detected at the
repository horizon in the Exploratory Studies Facility (ESF) (FAB98). These data indicate that
there are pathways in the rock formations above the proposed repository horizon that can rapidly
transmit infiltration water to the repository horizon in about 50 years; the pathways may extend to
greater depths. However, the data showed that the bomb-pulse Cl-36 was present in only a small
fraction of samples taken at the repository horizon, and these results could be correlated with well-
known fractures (FAB98).
These results demonstrate that it would be reasonable to expect that some relatively small fraction
of the entire UZ flow will occur via fast paths, and that modeling of UZ flow should take this into
account. It would not be reasonable, however, to base the evaluation of UZ performance on fast
paths alone. The reasonable expectation is that most of the UZ flow and radionuclide transport
will occur in accord with the bulk characteristics of the UZ geohydrologic regime.
In comparison with the reasonable assurance concept, reasonable expectation accommodates the
necessity for performance assessment results for a geologic repository to recognize the inherent
uncertainties and limitations of characterizing the natural system. Performance models can be
defined with as much mathematical sophistication as they are for reactors, and the analyses can be
as analytically complex as they are for reactors, but some of the models and parameters used in
repository performance analyses will inherently be less well defined than those used for reactors.
This can lead to particularly difficult problems if parameters are expected to be measured to too
high a degree of confidence, accuracy, or precision; in such a case, excessive conservatism may be
applied, in an attempt to offset the inability to meet these unrealistic data objectives.
The analyses should be based on reasonable models and reasonable parameter values, not biased
toward extremes by unrealistically conservative assumptions and parameter value selections. This
approach recognizes that uncertainty encompasses the high-end aspects of performance potential,
as well as the worst-case potential.
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5.4 Use of Reasonable Expectation for Yucca Mountain ',
Given the long time frame of the regulatory period for geologic disposal, the possibility |hat
changes in the repository system will occur over time, and the fact that, unlike reactors, prediction
with certainly of such changes and ability to remedy them is not possible, assumptions concerning
the agents and means of change are necessary. Similarly, assumptions are needed concerning
performance factors that are difficult or impossible to characterize reliably, such as the e,xtent to
which dripping water will wet the surface of a waste package. Reasonable expectation requires
that assumptions are reasonable, rather than purely biased toward conservatism, and that
performance factors that can be identified and potentially have a significant impact on
performance be reasonably valued and not omitted from the models and evaluations simply
because they are difficult to characterize. Consistent selection of conservative parameter values,
and omission of beneficial aspects of performance, because accurate characterization is difficult,
would result in unduly conservative performance assessments that represent situations of very low
probability. Decision-making using such analyses would be unavoidably biased.
It is reasonable to expect, for example, that climate conditions in the future can be estimated and
bounded on the basis of evidence of past and present climate conditions. It would be unreason-
able, however, to assume that future climate conditions will be extreme in comparison with the
past Also, in implementing the NAS finding that future performance of geologic features can be
bounded for periods up to one million years (NAS95), it would be reasonable to base the
assumptions on reasonable, not extreme, interpretations of past processes and events. Similarly, it
is not reasonable to assume that long-term changes will always be in the direction of worsening
performance, and to exclude positive aspects of such changes.
One of the most important aspects of reasonable expectation is to make reasonable assumptions
concerning performance factors that are difficult to quantify with confidence. There are numerous
performance parameters that can contribute significantly to system performance, but are difficult
to quantify accurately, such as the area of a waste package wetted by dripping water and the area
of spent fuel exposed in a breached fuel rod. To establish a realistic characterization oflthe
performance capability of the engineered barrier system, it is necessary to make reasonable
estimates for these factors and to include them in the performance models. As discussed in
Section 4.3, DOE used highly conservative assumptions for such factors in the TSPA-VA
evaluations or omitted them from the models because they were difficult to quantify. Our studies
have convinced us that the TSPA-VA results were consequently highly conservative and
understated the performance potential of the disposal system by several orders of magnitude.
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The effects of some of the TSPA-VA conservative assumptions on results of the TSPA-VA
evaluations can be estimated as follows:
• Assumption that the waste package is as wide as the drift: conservative by a factor of
three, since the package diameter is about one-third that of the drift.
» Assumption that the AHoy 22 is penetrated rapidly as a result of crevice corrosion:
conservative by a factor of 25, since the crevice corrosion rate was assumed to be
25 times higher than the general corrosion rate. This assumption was subsequently
modified to reflect the updated EDA II design, and this mode of degradation was
eliminated from TSPA-SR.
• Assumption that stainless-steel clad fuel rods are distributed among all packages and
fail immediately when the package is penetrated by water: conservative by about a
factor of about 10, since these rods can be packaged together in about one percent of
the total number of packages, and radionuelide releases were assumed to occur from
Zirealoy-elad fuel rods as well as the stainless-steel clad rods.
» Assumption that 0.1 percent of the Zircaloy-clad rods are failed at the time of
emplacement: conservative by a factor of 5-10, since an extensive database shows that
0.05-0.01 percent are failed.
• Assumption that the entire waste form area in a failed fuel rod is exposed and leaches
radionuclides when contacted by water: factor of 100 to 1,000; data show breaches of
cladding are primarily small hairline cracks, and all evidence shows that no significant
deterioration of cladding is expected after disposal.
Overall, many of the assumptions used in the TSPA-VA analyses can be shown, as illustrated
above, to have understated the reasonably expected performance of the repository by at least three
to four orders of magnitude. These arguments apply to the TSPA-VA, as a mechanism for
illustrating the concept of reasonable expectation.
Consideration of reasonable expectation in the TSPA-SR evaluations for the EDA II repository
design included the following:
« Use of a base case that is based on expected performance of the drip shields and the
waste package. As shown in the TSPA-SR, under these conditions, no radionuelide
releases would be expected for more than 10,000 years. Early waste package failures
were treated as possible, but their likelihood evaluated probabilistically and shown to
be unimportant in 10,000 years.
* Realistic estimates of seepage rates, the fraction of seeps that drip onto the drip shields
and subsequently onto waste packages, and the fraction of waste package surfaces that
is wetted. Realistic estimates can be based on emerging data which show that the
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seepage threshold may be as high as 200 mm/yr, i.e., 20 times higher than the estimate
of current infiltration rates. These were not used in the TSPA-SR (a mean threshold of
only 10 mm/yr was used), but may be included in future iterations of the TSPA.
Realistic estimates of the rate and mechanisms of penetration of the waste package wall
by corrosion. A rapidly growing database for corrosion of the wall materials'replaced
the assumed values used in the TSPA-VA that were based on expert elicitation results.
Improved estimates of the rate at which water can enter the waste package interior
through wall penetrations were not used in TSPA-SR, but could be adopted for future
iterations. Models of penetration blockage that were recognized for the TSPA-VA and
TSPA-SR evaluations but not included in the models can be adopted. Modified
assumptions for these effects would likely results in releases occurring at significantly
later times than found in the current model.
Realistic estimates of the time required to achieve contact between water that enters a
waste package and the exposed waste form. As a result of low seepage rates :and
limited entry pathways, the elapsed time to fill the package interior and achieve
water/waste contact can be tens of thousands of years. '
Realistic estimates of the duration and means for radionuclides mobilized from the
waste form to transport within, and exit, the interior of the waste package. As
discussed in Section 6.3, release of radionuclides from the package interior would be
expected to be controlled by extremely slow diffusional processes. By contrast, the
diffusional model included in the TSPA-SR is highly conservative, to the extent that
the majority of the releases are by diffusion. Modification of these assumptions would
lead to a qualitatively different type of release rate, in which significant releases would
not occur until substantial breaching of the waste container would permit advective
flow to dominate. Accommodating these alternative assumptions would likely delay
releases from the facility for tens of thousands of years.
Realistic estimates of radionuclide transit times and concentrations for migration from
the repository to the dose receptor location. The expanding database for the UZ and SZ
regimes should enable databased estimates of UZ and SZ flow and transport.
Realistic estimates of radionuclide concentration dilution associated with pumping by
the dose receptor. As previously noted, this performance factor was omitted! from the
TSPA-VA and TSPA-SR evaluations. Realistic studies including those done by the
NRC staff for the Issue Resolution Status Reports, indicate that the dilution factor for
this performance factor could be in the range 10-50.
Realistic estimates of the type of igneous activity expected in the Yucca Mountain
region rather than extreme strombolian events could be incorporated in future TSPAs.
Changing this assumption, by itself, may eliminate or greatly reduce the consequences
of an entire scenario (eruption) from the dose results of the first 10,000 years, although
not necessarily eliminating the occurrence of the igneous event. !
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• Realistic models of the contact between magma and waste packages, which account for
temperature decreases, may eliminate all consequences from igneous scenarios. By
accounting for these effects, the potential exists for the repository to be a zero-release
facility during 10,000 years.
Implementation of these applications of reasonable expectation would be expected to predict that
no radionuclide releases will occur until more than 10,000 years after disposal. In addition, long-
term dose rates would occur at much later times, and be significantly lower than those published in
the TSPA-SR.
In the TSPA-SR, as discussed in Chapter 4, DOE has introduced a variety of "uncertainty
importance" analyses, intended to investigate the extreme ends of output distributions. These
analyses include regression analysis and classification tree analysis (TRWOOa). Regression
analysis involves conducting stepwise linear rank regression between total dose and all input
parameters, to determine the strength of the relationship between parameters and the output they
produce. Classification tree analysis is a method for determining which variables or groups of
variables produce a particular category of results. In particular, this approach is used to look at
extremes in the output range, and to categorize which input parameters are associated with those
• extremes.
Since these uncertainly importance analysis techniques are focused purely on parameter
uncertainty, the degree to which they are consistent with the concept of reasonable expectation
depends on the conservatism of the underlying models and scenarios expressed by the
parameterizations they represent For scenario and model representations that are reasonable
representations of fee expected phenomena, it may well be appropriate to investigate and act upon
the boundaries of the output distributions. However, if the scenario and model descriptions
themselves are highly conservative, then making decisions based on the extrema of the parameter
distributions compounds the conservatism, and is not consistent with reasonable expectation. As
discussed in Chapter 4, several examples of models in the TSPA-SR appear to be so conservative
that they fall outside of the realm of expected system behavior, and the tails of the parameter
distributions appear to compound these conservatisms.
The igneous scenarios in the TSPA-SR appear to be an example of compounding conservatisms.
The annual probability of occurrence is highly uncertain, and one must look to the high end of the
possible values for the probability to consider the scenario at all, based on NRC guidance on
probability of scenarios (NRC99). The scenario description itself is for an extreme type of
volcanic event in a location in which such events are highly unlikely. The model for magmatic
interaction with the waste packages also takes extremely conservative assumptions, so that waste.
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packages are entirely destroyed, and the radionuclides are mobilized as an extreme finely ground-
up, easily dispersed powder. Despite these extreme assumptions, the central tendency of the
output distribution associated with parameter uncertainty provides a probability-weighted mean
dose of around 10~2 mrem/yr (see Figure 3-3). However, the distribution that produces this value
includes a few realizations of very low probability with substantial doses. Figure 6.1-2 of
TRWOOa illustrates that a few realizations produce doses in excess of 10 mrem/yr in the first
10,000 years. The potential exists to use uncertainty importance analysis methods to identify
conditions (input parameters) that lead to these high doses, and to use that information in decision
making: for example, to seek design modifications to the repository to mitigate them. However,
the concept of Reasonable Expectation would recognize that it is inappropriate to use the results of
extreme values of parameters applied in an extremely conservative model in an extremely
conservative scenario for prudent decision making. Similar, though less extreme, examples are
possible to elaborate for the nominal scenario of TSPA-SR as well.
i
5.5 Impact of Implementation of Reasonable Expectation for Yucca Mountain
The concept of Reasonable Expectation was developed by EPA to recognize that "absolute proof
of repository performance projections can not be obtained in the commonly understood meaning of
the term, because of the long time frames and inherent uncertainties of the extrapolations involved
in projecting repository performance. The approach, however, is intended to encourage realistic
assumptions and assessments of repository performance, which recognize these inherent
limitations. "Bounding" approaches that exclude important processes which will affect
performance because these processes are not readily quantified with high precision and accuracy,
or that frame performance scenarios unrealistically, have the danger of disguising important
aspects of the site performance. The effect of overly conservative analyses can be to drive
repository design efforts to unnecessary extremes or to set performance expectations beyond what
can be reasonably demonstrated with conservative but reasonable analyses. '
As discussed above, the EPA standards for Yucca Mountain were developed under the concept of
reasonable expectation. In examining the conservative basis for the TSPA-SR results, a reasonable
expectation approach to framing the performance scenarios and assumptions indicates that
expected performance would be orders of magnitude better than the TSPA-SR results. This
difference would be more than enough to compensate for the uncertainties in the assessments.
F
We believe the reasonable expectation approach is more appropriate for repository compliance
evaluations and provides a more realistic link between design and anticipated performance hi the
iterative process of developing a repository design for licensing.
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6.0 COST IMPACTS OF THE STANDARDS IN THE RULE
Preceding sections of this EIA have provided perspectives on the evolution of engineered design
features for a repository at Yucca Mountain; the evolution of understanding of the performance
potential for natural features of the Yucca Mountain site; the relationship between engineered and
natural barrier contributions to repository system performance; and the series of repository system
performance assessments that have provided insights leading to the current repository design
concept.
This section discusses the impact of the EPA's individual-protection standard, human-intrusion
standard, and ground water protection standard on the costs of the Yucca Mountain program and
the costs of the repository. Section 6.1 underscores the fact that individual-protection standards
are fundamental to radiation protection, and that the costs for the Yucca Mountain program and
repository design have evolved independent of the EPA IPS. Section 6.2 notes that the HIS is the
same as the IPS, and that it imposes no incremental costs. Section 6.3 demonstrates that the GWS
also imposes no incremental cost impacts.
In sum, the data and analysis requirements are the same for evaluating compliance with the IPS,
HIS, and GWS standards, and the Yucca Mountain program, repository design, and costs have
evolved without having been driven by the EPA standards.
6.1 The Individual-Protection Standard
As previously noted, the need for an individual-protection standard is fundamental to radiation
protection in general and to protection of health and safety for deep geologic disposal of
radioactive wastes. The issue here is not whether or not to have a standard; the issues are, what
level of protection should be required, and is there a cost impact of a standard that is more
stringent than an alternative? The choices under consideration are the 15 mrem/yr (CEDE)
standard selected by EPA and the 25 mrem/yr standard advocated by the NRC.
The issue concerning incremental cost for the more stringent standard can be addressed by
determining if there are any data collection requirements or design improvements imposed by the
more stringent standard. For Yucca Mountain, the basis for assessing the need for incremental
cost is provided by considering the information presented above in Sections 3 through 5
concerning the design features and projected performance for the EDA II design. As discussed in
Sections 3 and 4, the TSPA-SR shows mean doses two orders of magnitude less than the
15 mrem/yr standard for the reference individual at 10,000 years and 20 km, and the reference ,
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individual corresponds to the EPA's proposed RMEI. In addition, these doses are only realized for
a highly catastrophic and unlikely volcanic event. '
The assessment results are based on highly conservative assumptions, to the point that some of
them are highly unrealistic. Despite the conservatism, the results still showed potential to
demonstrate compliance with EPA's proposed individual-protection standard. If the EPA's
approach of "reasonable expectation" was used to frame the igneous scenario and assumptions, the
projected dose results may have been negligible during the first 10,000 years. \
The spread of the dose curves associated with parameter uncertainty shows that uncertainty in the
peak dose covers at least 5 orders of magnitude during the first 10,000 years (see Figure'6.1-2,
TRWOOa). A very few of the realizations appear to have extreme consequences, to the extent that
mean value is strongly biased by the high dose results. Indeed, for a portion of the curve this bias
is so strong that the mean value exceeds the 95th percentile of the dose curves. This suggests that
the mean dose curve is strongly influenced (perhaps unduly influenced) by a few realizations
representing the extreme tails of the distributions. In viewing these results, it must also be kept in
mind that the curves are themselves the result of the scenario and model assumptions discussed
above. Modification of these assumptions to reflect more reasonable system behavior would likely
decrease all of the output dose curves to negligible values in the first 10,000 years.
Demonstration of compliance with individual-protection standards for Yucca Mountain requires
detailed, in-depth characterization of engineered and natural barriers and analysis of performance
potential that assures a high degree of confidence in results presented for licensing reviews, and
results that indicate that the predicted performance is substantially better than the required
performance. As demonstrated clearly by Figure 3-2, current estimates of performance; are
significantly better than either the 25 mrem/yr standard or the 15 mrem/yr standard, and there is no
need for increased costs for design improvements or data acquisition to demonstrate compliance
with the 15 mrem/yr standard in comparison with the 25 mrem/yr standard. Indeed, it can be
argued that adoption of the EDA II design, with an incremental cost of only $0.8 billion out of a
total forward cost of nearly $22 billion (Section 3.8), is an effective time and cost saving strategy.
It reduces the uncertainties and issues that were of concern for the VA design, and it improves the
expected performance of the repository by several orders of magnitude, without facing the costs
and time involved in trying to reduce uncertainties in the performance of the natural barriers,
perhaps without definitive results. It can also be argued that more realistic treatment of juvenile
waste package failures in the TSPA-VA, together with a relatively minor design change to switch
the corrosion-resistant layer to the outside of the package, would by themselves have sufficiently
improved performance. By this line of argument, the full change to the EDA II design may have
6-2
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been unnecessary, so that even a simply modified VA design may have been able to meet the IPS
standards by orders of magnitude. In contrast to the several orders of magnitude improvement in
performance for the EDA n design as shown in the TSPA-SR, the NRC and EPA individual-
protection standards differ by less than a factor of two. The practical implication of tihis
observation is that the proposed design can be expected to protect the public far better man is
required by either of the slightly different standards.
6.2 Cost Impacts of the HIS Requirements
The standards for human intrusion, a performance standard unique to long-term geologic disposal,
are the same as those for the IPS. All parties to evaluation of factors important to this
demonstration of compliance concur with the NAS finding that a stylized scenario of intrusion and
its consequences is needed because circumstances of intrusion cannot be predicted on the basis of
scientific evidence. Therefore, the issues to be addressed in licensing reviews and compliance
evaluations are:
» whether the intrusion scenario considered for licensing is reasonable, and
• what are the dose consequences of the appropriate scenario.
The EPA's standard for individual exposure limits for human intrusion (15 mrem/yr) is no
different from the individual exposure limits applicable to gradual processes that will eventually
degrade the repository's functional capability. Protection of human health is independent of the
means by which it might be threatened. It is therefore appropriate and necessary for the EPA to
prescribe that the standard for human intrusion be no different than mat for the RMEI under
undisturbed performance of the repository. The EPA is concerned only with the fact that
individuals potentially affected by human intrusion be protected to the same extent as others.
Details of the stylized intrusion scenario given in the rule are based on the recommendations of the
NAS to EPA for the rulemaking (NAS95). EPA has adopted those recommendations it agrees
with, to make clear to DOE and NRC the intent of the standard. However, EPA has not prescribed
the scenario in excessive detail, thus allowing DOE and NRC to exercise their appropriate roles as
applicant and regulator in implementing the EPA standard. Considerable flexibility has been left
in the standard to explore the effects of alternative processes associated with releases from the
repository and transport through natural barriers.
As discussed above, it is apparent that the HIS requirements have no impact on the costs of the
DOE program for Yucca Mountain because, in fact, they are no different than the IPS
requirements, as should be the case (i.e., protection is independent of the circumstances that
require protection). Program schedules and costs for DOE have been established on the basis that
6-3
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demonstration of compliance with the IPS is needed and crucial; demonstration of compliance
with HIS requirements can be developed independently through the intrusion scenario characteris-
tics accepted for the basis for licensing reviews. Parameter values needed for the HIS analyses are
available either from the parameters used in the IPS analyses, or may be based on straightforward
assumptions without the need to collect additional field or laboratory data, as shown in Table 6-1.
i
Table 6-1. Data and Analysis Requirements for Assessing Compliance With ;
the Human-Intrusion Standard
Data/Analysis Requirement
Nature of the intrusion to be modeled
Probability of the intrusion
Time frame for the intrusion
Mechanism for release of radionuclides from the
penetrated waste package:
• direct fall down borehole
• leaking package or diffusion release
Transport of radionuclides through the saturated
zone to the compliance point
Doses to the receptor:
» definition of the receptor
* path through the biosphere
.,:;.;^'; . •. '•:.- ',-^:\^Sdurce6Fthelh!brmati6iiv-r:'; ^S>;'-"'-!;- -..
Defined in the standard: waste package penetration by water
well drilling with current technology; connection to the
saturated zone
Defined as unity (1 .0) in the standard •
Derived from corrosion modeling done for the IPS assessments
Assumptions for the analysis; no testing required ;
Required to use the same methods as for the IPS assessments
Same definition and analyses as for the IPS assessments
The key point is that the EPA standard is designed to assure that future populations are afforded
the same protection as present populations. DOE programs and projected costs have been
developed on the basis of the Department's expectations with regard to general licensing review
requirements for demonstration of compliance with applicable standards. They have not been
based on an assessment of the impact-of compliance with specific regulatory standards.
<
TSPA-SR estimates of the impact of inadvertent human intrusion to be about 3 orders of
magnitude below the standard, as shown in Figure 3-3. Differences are negligible between the
proposed NRC approach to assume intrusion at 100 years, and the reasonable expectation
approach, which would suggest that the waste package will be identifiable for much longer times.
It can be concluded that neither the HIS requirement nor its timing have any impact on repository
cost :
Data and analysis requirements for assessing compliance with the human-intrusion standard,
which fall within the framework of requirements for assessing compliance with the individual
6-4
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protection standard, are summarized in Table 6-1. From this table, it is apparent that parameters
and data necessary to analyze exposures are either defined in the rule or are already available from
the IPS assessments. Consequently, no additional demands for data collection are imposed by the
HIS. As a result, no additional significant program costs are imposed by the HIS requirements.
6.3 Cost Impact of the GWS Requirements
The Ground Water Protection Standards do not impose any additional costs on the program. The
information required to evaluate compliance with the GWS is radionuclide concentration in the
ground water as a function of distance from the repository. This is the same information as is
required for assessment of compliance with the IPS, and no incremental costs or effort to assess
ground water concentrations with a higher degree of certainty for the GWS in comparison with the
IPS is appropriate or necessary. As shown in Figure 3-2, the GWS is of the same order of
magnitude as the IPS, and the characteristics of the database that are needed for licensing reviews
are the same for the GWS and the IPS.
As shown in Figures 4-1 and 4-2, the TSPA-SR analysis indicates no potential for impact of the
GWS within the performance period, as there are no releases in the nominal scenario during this
period. As noted in Section 4.2, concentrations were calculated in the period out to 100,000 years
to demonstrate that no significant degradation occurs even after the 10,000-year time period is
ended.
Data and analysis requirements for assessing compliance with the ground water protection
standards are summarized in Table 6-2.
6-5
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Table 6-2. Data and Analysis Requirements for Assessing Compliance With
the Ground Water Protection Standards
• • • I .-...-.. v~:- •...•'- V. ' . •'--,-. - ..V
Data/Analysis Requirement ^J : K Scarce of the Information , ;
Water flux through the unsaturated zone above and into
the repository (precipitation, infiltration, seepage into
drifts, etc.)
Source term for radionuclide releases from the
repository (container failure profiles, exposed waste
form areas, radionuclide leach rates, solubilities, etc.
Characterization of saturated zone flow and
radionuclide transport (hydropoaie conditions down-
gradient to the compliance point; only average values
are required by the GWS )
Methods for calculating radionuclide concentrations in
the Representative Volume
Characterization data, models, and analyses for the IPS
compliance evaluations
Engineered barrier system characterization, testing and
modeling as required for the IPS compliance Devaluations
Characterization data, flow and transport models, and
analysis of the type required by the IPS compliance
evaluations, but GWS requires less detail
Methods defined in the standard; no further effort
required ' !
6-6
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7.0 SUMMARY DEMONSTRATION THAT THE EPA STANDARDS HAVE NO COST
IMPACTS ON THE YUCCA MOUNTAIN PROGRAM AND REPOSITORY
7,1 Principal Bases for Findings of No Cost Impacts
This Economic Impact Assessment (EIA) has demonstrated that DOE's strategy for development
and design of a possible repository at Yucca Mountain has evolved to the point that EPA's 40 CFR
Part 197 standards will have no impact on the total life-cycle costs of the repository. This has been
demonstrated through an examination of the factors that influenced evolution of repository design
and a review and analysis of DOE's performance assessments. The principal factors that provide
the basis for a finding of no-cost impact of the standards are:
• The DOE plans for repository design strategy, data acquisition, and budget'allocations
and requirements have been established independent of the EPA standards. DOE's
plans and cost estimates reflect, as suggested above, expenditures and activities not
needed as a direct consequence of the EPA standards.
• Earlier performance assessment results (TSPA-VA), which are based on highly
conservative assumptions that would not be used under principles of Reasonable
Expectation, suggest expectation of compliance with EPA's IPS, HIS and GWS limits.
More recent performance assessment results (TSPA-SR) show even greater margins for
compliance with the EPA standards than the TSPA-VA results. The newer design
(EDA II) is augmented to produce improved expected performance for me nominal
case, and design features have been selected to reduce the potential for significant
issues during licensing reviews. Figure 3-2 demonstrates dramatically the assertion
that EPA's standards have no impact on Yucca Mountain program costs. Under the
nominal scenario there is no release during the time period over which the IPS, HIS,
and GWS would apply. Releases may only be expected to occur if violent volcanic
activity occurs at the site, and this is unlikely considering the volcanic history of the
site. The magnitude of releases associated with volcanic activity are very
conservatively estimated in the TSPA-SR in comparison to reasonably expected
conditions.
• The data and analysis requirements for assessing compliance with the ground water
protection and human-intrusion standards are the same as those required for assessing
compliance with the fundamental and essential individual-protection standard. The
ground water protection standard and the human-intrusion standard therefore impose no
incremental costs.
These factors are discussed in more detail in Sections 7.1.1 through 7.1.3. Section 7.2 discusses
alternative standards and their relationship to repository performance, and Section 7.3 provides an
overall summary and conclusions.
7-1
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7.1.1 Evolution of the Repository Design and Roles of Natural and Engineered Features
The initial repository design concept, described in the Site Characterization Plan (SCP) issued in
1988, anticipated that natural features of the repository system, such as very low rates of water
movement in the unsaturated zone (UZ), would dominate repository performance. Engineered
features would be the minimum necessary to meet the subsystem performance requirements of the
Nuclear Regulatory Commission's (NRC) 10 CFR Part 60 standards, such as substantially
complete containment of radionuclides within the waste package for 300-1,000 years. ,
In contrast to SCP expectations, acquisition and analysis of subsequent site characterization data
revealed that the SCP's performance expectations for the natural system would not be achieved,
e.g., there are paths for rapid movement of water through the UZ and rates of ground water
infiltration were higher than earlier thought. Consequently, the performance capabilities of the
engineered features of the system have been revised from the SCP concept to one in which the
engineered features play the dominant role in disposal system performance during the regulatory
period: more specifically, the use of highly corrosion-resistant waste package wall materials and
drip shields to defer contact of the waste packages by water that drips into the repository. The
design features arc intended to provide defense-in-depth for performance and to minimize
uncertainties and technical issues associated site performance that could become contentious issues
during the licensing process. :
The inversion of performance roles of the natural and engineered features of the repository system
has evolved as a result of site characterization findings, guidance from external reviews ;such as
those of the Nuclear Waste Technical Review Board, and interactions with NRC staff wjiich
provide guidance on licensing requirements. The evolution has been independent of fjhe EPA
standards, the major components of which have remained essentially unchanged since the
1985 promulgation of the generic 40 CFR Part 191 standards for geologic disposal.
7.1.2 DOE's Use of Performance Evaluations I
The Department has used a series of Total System Performance Assessments (TSPA) to guide
selection and prioritization of site characterization activities, to guide selection of engineered
features and parameters, and to make projections of repository safety performance. TSPA models
and methodology have evolved in parallel with the evolution of the site database and engineered
design concepts. !
7-2
-------
The TSPA for the Viability Assessment in 1998 (TSPA-VA) was the first TSPA for a potential
repository system at the Yucca Mountain site. Despite use of conservative models and
assumptions, TSPA-VA results for the base case using average parameter values showed dose
rates at 10,000 years, for a dose receptor at 20 km distance from the repository and with
characteristics comparable to EPA's Reasonably Maximally Exposed Individual (RMEI), that
were two orders of magnitude lower than the EPA's individual-protection standard of 15 mrem/yr
CEDE. More reasonable assumptions in framing these scenarios and the associated conceptual
models would show lower projected doses of at least several orders of magnitude.
In response to reviews of the TSPA-VA which found that there were uncertainties in the models
and results that could produce significant technical issues for licensing reviews, DOE subsequently
adopted the current engineered design, EDA II, which has as principal features use of titanium drip
shields and a highly corrosion resistant waste package outer wall. This engineered barrier design
concept is significantly augmented in comparison with the VA design. TSPA-SR estimates of
performance for this design indicate that, under expected conditions, there will be no radionuclide
releases and no potential for radiation doses for more than 10,000 years after repository closure,
unless the repository is disrupted by volcanic activity. Even in that extreme occurrence, the
repository is shown in the TSPA-SR not to exceed the exposure limits. The performance scenarios
and conceptual models in the TSPA-SR were also developed using conservative assumptions,
although more realistically than the TSPA-VA approaches. Expected releases would be
considerably lower for even more realistic assessments.
All of the above actions were completed or underway by the time NRC put forth its proposed
10 CFR Part 63 regulations in February 1999 and EPA put forth its proposed 40 CFR Part
197 standards in August 1999. In particular, DOE program plans, repository design
concepts, and program cost estimates had all been documented before EPA's proposed
standards were issued for public comment.
7.1.3 Impact of the EPA Standards on Data and Analysis Requirements
The third perspective included in this EIA is an examination of the data and analysis requirements
imposed by the individual-protection, ground water protection, and human-intrusion standards.
Each of these components of the standard requires a quantitative evaluation of projected repository
performance, and a database of performance parameters for the repository's natural and engineered
features, for compliance assessment. This EIA demonstrates that the data and analysis
requirements for assessing compliance with the ground water protection and human-intrusion
standards are the same as those required for assessing compliance with the fundamental and
7-3
-------
essential individual-protection standard. The ground-water-protection and human-intrusion
provisions therefore impose no incremental cost impacts. ;
7.2 Comparative Impacts of Alternative Dose Limits for the Individual-Protection
Standard
An important issue in developing the individual-protection standard has been comparative impacts
of alternative dose limits, e.g., 15 mrem/yr versus 25 mrem/yr. Figure 3-2 (which is the same as
Figure ES-1) shows the performance projections EDA II designs given in TSPA-SR. |
i
As seen in Figure 3-2, the EDA II repository design demonstrates performance such thatprojected
doses are significantly less than either the 15 mrem/yr or the 25 mrem/yr dose limit. Indeed, the
only doses that occur in the first 10,000 years are the result of potential volcanic activity, scenarios
that are very conservative. It is therefore evident that selection of a 15 mrem/yr dose limit rather
than a 25 mrem/yr limit will not impose any additional cost impacts on the repository. This is a
highly significant finding in that the 15 mrem/yr CEDE dose limit is consistent with the :
recommendations of the National Academy of Sciences and regulatory precedents for deep
geologic disposal applications (WIPP).
As noted in Section 4 of this document, the TSPA-VA evaluations of potential VA-repository
performance used highly conservative models and assumptions, such that the actual expected
performance of a VA repository would be at least several orders of magnitude better than was
reported in the TSPA-VA results. Similarly, with the enhanced engineered barrier system design
for EDA n, the performance as evaluated in the TSPA-SR is significantly better than that
projected for the VA. No radionuclide releases are expected to occur for more than 10,000 years,
and even if highly-improbable violent strombolian eruption occurs, the repository design easily
meets either the 15 mrem/yr or the 25 mrem/yr Limit. Performance scenarios in the TSPA-SR
analyses and the models used to evaluate them, although different in many respects from! the
TSPA-VA, are still very conservative. Analyses using more realistic, yet still defensiblej
assumptions would show performance results considerably better than the one presented jin the
TSPA-SR. I
i
The projections of repository performance for the EDA II design are shown in Figures 3;-2 and 3-3
compared to the EPA and proposed NRC regulations. As can be seen in these figures, and as
noted above in the discussion of the alternative dose limits, performance in all cases considered is
significantly better than required by the standards. The highly conservative igneous intrusion and
eruptions considered in the TSPA-SR show dose estimates one to two orders of magnitude below
7-4
-------
the limits imposed by the standards; the expected performance (nominal scenario, excluding
volcanic events) within the regulatory time period for the EDA II repository shows no releases
relevant to the proposed standards.
As discussed in Section 3.4, the EDA II design and the refinement of repository strategy serve
primarily to ease concerns for uncertainties and technical issues that were associated with the
TSPA-VA methodology that could be difficult to resolve in licensing reviews, and to add to the
performance margin with use of drip shields to implement defense-in-depth concepts. The new
design was not driven by requirements in the EPA rule, but rather as a means to compensate for
uncertainties in performance projections.
7.3 Summary and Conclusions
The need to demonstrate compliance with the individual-protection standard is fundamental to
assurance of protection of public health and safety for deep geologic disposal. There is also need,
for geologic disposal, to provide protection in the event of inadvertent future human intrusion and
there is need to protect ground water resources for future generations. Imposition of, and
compliance with, the HIS and GWS standards is essential for consistent and comprehensive
application of EPA policy concerning ground water protection and for appropriate application of
generic principles set forth in 40 CFR Part 191 to the Yucca Mountain setting.
As shown in this document, the evolving understanding of the Yucca Mountain site characteristics,
and the resulting information base needed to provide defense-in-depth and to reduce uncertainties
during licensing reviews has driven the Yucca Mountain program data acquisition program and
evolution of design concepts. Because of site-specific conditions, DOE's strategy for development
and design of a possible repository at Yucca Mountain has evolved so that EPA's 40 CFR Part 197
standards will have no impact on the costs of the repository program. This document has also
shown that EPA's generic 40 CFR Part 191 standards did not influence evolution of the Yucca
Mountain program or the repository design. Moreover, as illustrated by Figures 3-2 and 3-3,
expected performance for the current repository design is significantly better than is required by
the EPA standards for HIS, GWS, and IPS.
The information base required for demonstrating compliance with the HIS and GWS standards is
the same as that required for demonstrating compliance with the individual-protection standard.
Costs and effort above those needed to evaluate compliance with the IPS therefore do not have to
be incurred to evaluate compliance with the HIS and GWS standards.
7-5
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7-6
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8.0 REFERENCES
AEA54 Atomic Energy Act, Public Law 83-703, as amended, 42 USC 2011 et seq., 1954.
BER92 Bernard, R.W. et aL, TSPA 1991: An Initial Total System Performance Assessment
for Yucca Mountain, SAND91-2795, Sandia National Laboratories, My 1992.
BOD97 Bodvarsson, G.S., Bandurraga, T.M., and Wu, Y.S., editors, The Site-Scale
Unsaturated Zone Model of Yucca Mountain, Nevada, for the Viability Asessment,
LBNL^0376, Berkeley, California, 1997.
DAG97 D'Agnese, F.A., Faunt, C.C, Turner, AJL, and Hill, M.C., Hydrogeologic
Evaluation and Numerical Simulation of the Death Valley Regional Ground Water
Flow System, Nevada and California, Water-Resources Investigations Report 96-
4300, U.S. Geological Survey, 1997.
DOE86 U.S. Department of Energy, Issues Hierarchy For A Mined Geologic Disposal
System, DOE/RW-0101, September, 1986.
DOE88 Site Characterization Plan - Overview: Yucca Mountain Site, Nevada Research
and Development Area, Nevada, DOE/RW-0198, U.S. Department of Energy,
December 1988. ;
DOESSa Site Characterization Plan, Yucca Mountain Site, Nevada Research and
Development Area, Nevada, DOE/RW-0199, U.S. Department of Energy.
DOE94 Total System Performance Assessment ~ 1993: An Evaluation of the Potential
Yucca Mountain Repository, BOOOOOOOO-0717-2200-0099-Rev, 01, Prepared by
R.W. Andrews et al. INTERA, Inc., March 1994.
DOE95 Total System Performance Assessment - 1995: An Evaluation of the Potential
Yucca Mountain Repository, BOOOOOOOO-01717-2200-00136, Rev. 01, prepared by
TRW Environmental Safety Systems Inc., November 1995.
DOE97 Unsaturated Zone Flow Model Expert Elicitation Project, CRWMS M&O
Contractor, Las Vegas, Nevada, 1997.
8-1
-------
DOE98 Viability Assessment of a Repository at Yucca Mountain, Volume 3 - Total System
Performance Assessment, DOE/KW-0508, U.S. Department of Energy, December
1998. '
DOE98a Total System Performance Assessment - Viability Assessment (TSPA- VA) -Analyses
Technical Basis Document, BGOOOOOOO-01717-4301-00005 REV 01, U.Sr
Department of Energy, November 13,1998. i
I
DOE99 Draft Environmental Impact Statement for a Repository at the Yucca Mountain Site,
Nye County, Nevada, U.S. Department of Energy, DOE/RW-0250D, August 1999.
DOE99a U.S. Department of Energy, Yucca Mountain Site Suitability Guidelines, 64 Federal
Register 67054-67089, November 30,1999. :
DOEOO U.S. Department of Energy, Dike Propagation Near Drifts, Analysis/Model Report
ANL-WIS-MD-00001S Rev 00, April 2000. :
I
DOEOOa U.S. Department of Energy, Characterize Framework for Igneous Activity at Yucca
Mountain, Nevada, Analysis/Model Report ANL-MGR-GS-000001 Rev 00, June
2000. :
i
DOE01 U.S. Department of Energy, Yucca Mountain Science and Engineering Report,
Technical Information Supporting Site Recommendation Consideration, DOE/RW-
0539, May 2001. . i
DOE01 a U.S. Department of Energy, Analysis of the Total System Life Cycle Cost of the
Civilian Radioactive Waste Management Program, DOE/RW-0553, May 2001.
i
EDD01 Eddebborh, ALA., What is the Mean and Variance of Transport Time of a
Conservative Species in the SZ, Handout for Presentation to the Nuclear Waste
Technical Review Board Meeting, January 30-31, Amargosa'Valley, Nevada, 2001.
I
EnP92 Energy Policy Act of 1992, Public Law 102-486, October 24,1992. |
EPA76 U.S. Environmental Protection Agency, National Interim Primary Drinking Water
Regulations, EPA 50/9-76-003,1976. '
8-2
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EPA85 U.S. Environmental Protection Agency, final Rule, Environmental Standards for
the Management and Disposal of Spent Nuclear Fuel, High-Level, and Transuranic
Radioactive Wastes, Federal Register, 50 FR 38066-38089, September 19, 1985.
EPA91 U.S. Environmental Protection Agency, 40 CFR Parts 141 and 142, Proposed Rule,
National Primary Drinking Water Regulations; Radionuclides, Federal Register, 56
FR 33050, July 18,1991.
EPA93 U.S. Environmental Protection Agency, 40 CFR Part 191, Environmental Radiation
Protection Standards for the Management and Disposal of Spent Nuclear Fuel,
High-Level, and Transuranic Radioactive Wastes', Final Rule, Federal Register, 58
FR 66398-66416, December 20,1993.
EPA99 U.S. Environmental Protection Agency, 40 CFR Part 197, Environmental Radiation
Protection Standards for Yucca Mountain, Nevada; Proposed Rule, Federal
Register, 64 FR 46976-47016, August 27, 1999.
EPAO1 U.S. Environmental Protection Agency, 40 CFR Part 197, Public Health and
Environmental Radiation Protection Standards for Yucca Mountain, Nevada; Final
Rule, Federal Register, 66 FR 32074-32135, June 13,2001. ',
EPROO Electric Power Research Institute, Evaluation of the Candidate High-Level Waste
Repository At Yucca Mountain Using Total System Performance Assessment, Phase
5, EPRI Report 1000802, November 2000.
FAB98 Fabryka-Martin, J.T. et al., Distribution of Fast Hydrologic Paths in the
Unsaturated Zone at Yucca Mountain, Proceedings of the Eighth Annual
Conference on High-Level Radioactive Waste Management, American Nuclear
Society, May 11-14, 1998.
FLI96 Flint, A.L., Hevesi, J.A., and Flint, L.E., Conceptual and Numerical Model of
Infiltration for the Yucca Mountain Area, U.S. Geological Survey, 1996.
GEO98 Geomatrix Consultants Inc. and TRW, Saturated Zone Flow and Transport Expert
Elicitation Project, January 1998.
8-3
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ICR77 International Commission on Radiation Protection, Recommendations of the
International Commission on Radiation Protection, ICRP Publication 26,,
Pergamon Press, Oxford, 1977. i
KLOZ97 Kozak, M.W., Sensitivity, Uncertainty, and Importance Analyses, Proceedings of
the 18th Annual Department of Energy Low-Level Radioactive Waste Management.
Conference, May 20-22, Salt Lake City, 1997. :
NAS95 National Academy of Sciences, Technical Bases for Yucca Mountain Standards,
Committee on Technical Bases for Yucca Mountain Standards, National Research
Council, National Academy Press, 1995. ;
NIX70 The White House, President R. Nixon, Reorganization Plan No, 3 of 197$, Federal
Register, 35 FR 15623-15626, October 6,1970. ;
NRC99 U.S. Nuclear Regulatory Commission, Disposal ofHigh-Level Radioactive Wastes
in a Proposed Geological Repository at Yucca Mountain, Nevada; Proposed Rule,
64 Federal Register 8640-8679, February 22,1999.
i
NRC99a U.S. Nuclear Regulatory Commission, Issue Resolution Status Report, Key
j
Technical Issue: Igneous Activity, Revision 2, July 1999. j
NWP83 Nuclear Waste Policy Act of 1982, Public Law 97-425, January 7,1983. \
NWP 87 Nuclear Waste Policy Amendments Act of 1987, Public Laws 100-202 and 100-203,
December 22,1987. . ;
NYEOO N. Stellavato, Nye County Drilling: Phase 2, Presentation at the January 25-26,
2000 meeting of the Nuclear Waste Technical Review Board, Las Vegas, Nevada.
PRP99 Total System Performance Assessment Peer Review Panel, Final Report,
February 11,1999.
SCA99 S. Cohen & Associates, Inc, Effectiveness of Fuel Rod Cladding as an Engineered
Barrier in the Yucca Mountain Repository, December 1999.
8-4
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SCAOO S. Cohen & Associates, Inc., Characterization and Comparison of Alternate Dose
Receptors for Individual Radiation Protection for a Repository at Yucca Mountain,
April 2000 (Docket No. A-015-12, V-B-3).
TRB96 U.S. Nuclear Waste Technical Review Board, Report to the U.S. Congress and The
Secretary of Energy, January to December 1996.
TRB99a DOE Presentations to the Nuclear Waste Technical Review Board, June 29-30,
1999, Beatty, Nevada.
TRB99b DOE Presentations to the Nuclear Waste Technical Review Board, September 14-
15, 1999, Alexandria, Virginia.
TRBOOa Stuckless, J., Natural Analog Studies, Presentation at the January 25-26,2000,
meeting of the Nuclear Waste Technical Review Board, Las Vegas, Nevada.
TRBOOb Bodvarrson, G.S., Application of Principal Factors: Seepage Studies, Presentation
at the January 25-26, 2000, meeting of the Nuclear Waste Technical Review Board,
Las Vegas, Nevada.
TRWOO TRW Environmental Safety Systems, Inc., Repository Safety Strategy: Plan to
Prepare the Postclosure Safety Case to Support Yucca Mountain Site Recommen-
dation and Licensing Considerations, TDR-WIS-RL-000001 Rev 03, January 2000.
TRWOOa TRW Environmental Safety Systems, Inc., Total System Performance Assessment
for the Site Recommendation, TDR-WIS-PA-000001 REV 00ICN 01, December
2000.
USC8 7 United States Court of Appeals for the First Circuit, Natural Resources Defense
Council Inc., etal, v. United States Environmental Protection Agency, Docket No.:
85-1915, 86-1097, 86-1098, Amended Decree, September 23,1987.
WIL94 Wilson, Michael et al., Total System Performance Assessment for Yucca Mountain -
SNL Second Iteration (TSPA-1993), Sandia National Laboratories, April 1994.
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— Editorial Changes to Docket Version — i
I
The following changes to the EIA in the Yucca Mountain Rule Docket (A-95-12, V-B-2) were
made in this document to correct typographical and other minor errors in the text.
P. i Item 1.1 changed to read, "EPA Action and Authority" :
Item 1.7 changed to read, "Final 40 CFR Part 197 - Public Health and Environmental
Radiation Protection Standards for Yucca Mountain, Nevada" !
P. ii Item 4.1 changed to read, "Performance in Comparison with the Individual-Protection
Standards" , :
Item 4.2 changed to read, "Performance in Comparison with the Ground Water
Protection Standards" •
P. iv Titles for Tables 3.6 and 3.7 added to the Table of Contents, existing Table 3.7
renumbered as 3.8. ;
I
Figure ES-1 changed to read, "Comparison of Radiation Protection Standards with
Expected Values of TSPA-SR Calculations for a Repository at Yucca Mountain for
Nominal and Igneous Scenarios"
P. v Acronym OCRWM (Office of Civilian Radioactive Waste Management) addfd
i
i
P. 1-1 Changed first sentence in subsection 1.1 to read, "The U.S. Environmental Protection
Agency has issued a rale "
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P. 1-2 Changed the third sentence in subsection 1.3, paragraph 1, to read, "The regulation
contains site-specific environmental standards "
P. 1-11 Changed the subsection heading to read, "Final 40 CFR Part 197 - Public Health and
Environmental Radiation Protection Standards for Yucca Mountain, Nevada.?*
P. 1-12 Changed the reference designation in subsection 1.7.1 italicized text to read (EPA01).
P. 1-13 Inserted the word "proposed" hi the first sentence below the first bulleted text in
subsection 1.7.2, before the text "...§197.26" !
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Changed the first sentence after the second bulleted text to read, "In the final rule, EPA
selected this second alternative in which DOE must project..."
P, 1-14 Modified the final sentence 6f the first paragraph of siibsection 1.7,3 to read,
"However, it is Agency policy, as well as national policy and the policy of many states,
to protect ground water resources."
Deleted the sentence that begins, "If revised MCLs are promulgated ", and the
concluding sentence to the paragraph.
P. 1-15 Revised the italicized text to match the regulatory text in §197.30 of the final rule.
P. 1-16 Revised the italicized text to match the regulatory text in §197.21 of the final rule.
P. 1-18 Changed the first sentence of the second paragraph to read, "Another alternative RV
proposed was 120 acre-ft/yr."
P.l-19 Deleted "has" from the first sentence of the second paragraph.
Corrected the reference (EPA99) to read "(EPA99, 01)."
Modified the second paragraph, last sentence to read "...portion of the Town of
Amargosa Valley..."
P.3-33 In the second paragraph of subsection 3.7.1.3, changed "draft 40 CFR part 197" to read
"proposed 40 CFR Part 197".
P. 3-37 In the first sentence of paragraph one of Section 3.8, changed the text to reference
Table 3-8, rather than 3-7.
P. 3-39 Table 3-7 changed to Table 3-8; Item 4 total cost corrected to 35.4 from 36.3;
Item 6 revised to read, 'Total Repository Cost (1983-2119)";
Item 7 revised to read, " Total Program Cost"
P. 4-3 Changed 20 km to 18 km.
P. 5-1 Modified the last sentence of the second paragraph of Section 5.1 to read, "Reasonable
assurance is a concept that has been used...."
P. 5-2 Inserted the reference designation "(EPA99)" in the second paragraph, first sentence
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P. 6-4 Changed paragraph two, second sentence to read, "...with regard to general licensing
review requirements.... with applicable standards." j
Changed paragraph two, last sentence to read, "...on an assessment of the impact of
compliance with specific EPA regulatory standards." !
P. 8-3 Added reference to EPA final standards (EPA01) |
Deleted the word "proposed" from the following text locations which refer to the final rule:
P. ES-3, Figure ES-1 (and added proposed to the line "NRC proposed 25 mrem/yr IPS Limit");
p. 1-1, section 1.1 text and heading and subsection 1.2; p. 1-2, subsection 1.3, first sentence; p. 1-9,
first paragraph; p. 3-39 (Figure 3-2); p. 3-37, second paragraph, last sentence - inserted "the
relevant"; p. 4-1, heading for Section 4.1; p. 4-2, second paragraph, 4 th sentence; p. 4-3, second
paragraph, first sentence, and the heading for section 4.2; p. 4-16, paragraph two, last sentence;
p. 4-17, paragraph one, first sentence; p. 5-1, section 5.1 - second paragraph, first sentence;
p.5-2, first paragraph, first two sentences; p. 5-3, second paragraph, first sentence; p. 6-1;,
paragraph four, last sentence and inserted the word "selected; p. 6-3, second to last sentence of
paragraph one, the first sentences of paragraphs two and three; p. 6-4, first sentence of paragraph
two; p. 7-1, first sentence of the text in bulleted texts; p. 7-3, second sentence of paragraph two;
p. 7-5, first sentence of paragraph two and inserted the word "proposed" before "NRC ;
regulations."
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