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40 CFR Part 197
BACKGROUND INFORMATION DOCUMENT
FOR 40 CFR 197
PUBLIC HEALTH AND ENVIRONMENTAL
RADIATION PROTECTION STANDARDS
FOR YUCCA MOUNTAIN, NEVADA
June 2001
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
Washington, DC 20460
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TABLE OF CONTENTS
List of Acronyms
Executive Summary ES-l
CHAPTER 1 INTRODUCTION l_l
1.1 Purpose and Scope of the Background Information Document l-l
1.2 EPA's Regulatory Authovity for the Rulemaking 1_3
1.3 The National Academy oi Sciences Recommendations ; 1-3
1.4 History of EPA's Rulemaking 1-6
1.4.1 Legislative History j_g
1.4.2 The Development of EPA's Role in the Federal Program 1-11
1.4.3 Early Federal Aciion j_j2
1.4.4 40 CFRPart 191 ...............(......... 1-14
References j_jg
CHAPTER 2 HISTORY OF RADIATION PROTECTION IN THE UNITED STATES
AND CURRENT REGULATIONS 2-1
2.1 Introduction 2-1
2.2 The International Commission on Radiological Protection, the National Council on
Radiation Protection and Measurements, and the International Atomic Energy
Agency 2-2
2.3 Federal Radiation Council Guidance ." 2-7
2.4 Environmental Protection Agency 2-9
.2.4.1 Environmental Radiation Exposure 1 2-10
2.4.2 Environmental Impact Assessments 2-10
2.4.3 Ground Water Protection 2-11
2.4.4 Radionuclide Air Emissions 2-13
2.4.5 Disposal of High-level Radioactive Waste and Spent Nuclear Fuel 2-15
2.4.6 Evaluation of Radiation Dose 2-17
2.5 Nuclear Regulatory Commission 2-18
2.5.1 Fuel Cycle Licensees 2-19
2.5.2 Radioactive Waste Disposal Licenses 2-19
2.5.3 Repository Licensing Support Activities 2-21
2.6 Department of Energy 2-23
References 2-25
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TABLE OF CONTENTS (Continued)
Pas
CHAPTER 3 SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE
DISPOSAL PROGRAMS IN OTHER COUNTRIES 3-1
3.1 Belgium 3-3
3.1.1 Nuclear Power Utilization 3~3
3.1.2 Disposal Programs and Management Organizations 3-5
3.1.3 Regulatory Organizations and Their Regulations 3-7
3.2 Canada • 3~7
3.2.1 Nuclear Power Utilization 3-7
3.2.2 Disposal Programs and Management Organizations 3-8
3.2.3 Regulatory Organizations and Their Regulations 3-10
3.3 Finland 3"11
3.3.1 Nuclear Power Utilization 3-11
3.3.2 Disposal Programs and Management Organizations 3-11
3.3.3 Regulatory Organizations and Their Regulations 3-12
3.4 France 3"13
3.4.1 Nuclear Power Utilization 3-13
3.4.2 Disposal Programs and Management Organizations 3-13
3.4.3 Regulatory Organizations and Their Regulations - 3-15
3.5 Germany • • 3"16
3.5.1 Nuclear Power Utilization 3"16
3.5.2 Disposal Programs and Management Organizations : 3-16
3.5.3 Regulatory Organizations and Their Regulations 3-18
3.6 Japan 3~19
3.6.1 Nuclear Power Utilization -3-iy
3.6.2 Disposal Programs and Management Organizations 3-19
3.6.3 Regulatory Organizations and Their Regulations 3-22
^ 00
3.7 Spain ^
3.7.1 Nuclear Power Utilization J-zz
3.7.2 Disposal Programs and Management Organizations 3-23
3.7.3 Regulatory Organizations and Their Regulations 3-24
3.8 Sweden 3"24
3.8.1 Nuclear Power Utilization 3-24
3.8.2 Disposal Programs and Management Organizations 3-25
3.8.3 Regulatory Organizations and Their Regulations 3-26
3.9 Switzerland 3"27
3.9.1 Nuclear Power Utilization 3-27
3.9.2 Disposal Programs and Management Organizations 3-28
11
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TABLE OF CONTENTS (Continued)
3.9.3 Regulatory Organizations and Their Regulations 3-29
3.10 United Kingdom 3-30
3.10.1 Nuclear Power Utilization 3-30
3.10.2 Disposal Programs and Management Organizations 3-31
3.10.3 Regulatory Organizations and Their Regulations 3-32
References 3-33
CHAPTER 4 U.S. PROGRAMS FOR THE MANAGEMENT AND DISPOSAL OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE
WASTE AND THE EVALUATION OF YUCCA MOUNTAIN 4-1
4.1 Introduction 4-1
4.2 The Department of Energy 4-1
4.2.1 DOE's Office of Civilian Radioactive Waste Management (OCRWM)
4-3
4.2.2 DOE Management and Disposal of Defense Wastes '. 4-4
4.3 The Nuclear Regulatory Commission : 4-5
4.3.1 Legislative Requirements and Regulatory Framework 4-5
4.3,2 Status of NRC's Program 4-6
4.4 Nuclear Waste Technical Review Board 4-7
4.5 State and Local Agencies 4-8
4.6 Native American Tribes 4-9
References 4-11
CHAPTER 5 QUANTITIES, SOURCES, AND CHARACTERISTICS OF SPENT NUCLEAR
FUEL AND HIGH-LEVEL WASTE IN THE UNITED STATES 5-1
5.1 Introduction 5-1
5.2 Spent Nuclear Fuel 5-1
5.2.1 Commercial Spent Nuclear Fuel Inventory and Projection 5-3
5.2.2 DOE Spent Nuclear Fuel 5-4
5.3 Defense High-level Radioactive Waste 5.7
5.3.1 High-level Waste Inventories at the Hanford Site 5-11
5.3.2 High-level Waste Inventories at INEEL 5-12
5.3.3 High-level Waste Inventories at the Savannah River Site 5-13
5.3.4 High-level Waste Inventories at the West Valley
Demonstration Project 5-13
5.4 Significant Radionuclides Contained in Spent Nuclear Fuel
and High-level Waste 5-14
References 5-16
111
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TABLE OF CONTENTS (Continued)
CHAPTER6 DOSE AND RISK ESTIMATION 6-1
6.1 Introduction 6-1
6.2 Dose Estimation 6-1
6.3 Cancer Risk Estimation 6-3
6.4 Genetic Effects 6-4
6.5 Developmental Effects 6-7
6.5.1 In Utero Carcinogenesis 6-7
6.5.2 Brain Teratology 6-7
6.5.3 Other Effects of Prenatal Irradiation 6-8
6.5.4 Summary of Developmental Effects 6-9
References 6-10
CHAPTER 7 CURRENT INFORMATION CONCERNING A POTENTIAL WASTE
REPOSITORY AT YUCCA MOUNTAIN 7-1
7 1 Principal Features of the Natural Environment 7-1
7.1.1 Geologic Features 7-1
7.1.2 Hydrologic Features 7-57
7.1.3 Climate Considerations 7-113
7.2 Repository Concepts under Consideration for Yucca-Mountain 7-121
7.2.1 Conceptual Repository Systems 7-121
7.2.2 Design Concepts for Engineered Features of the VA Repository... 7-123
7.3 Repository System Performance Assessments 7-138
7.3.1 DOE's Historic Performance Assessments 7-139
7.3.2 DOE's TSPA for the Viability Assessment (TSPA-VA) 7-143
7.3.3 TSPA-VA Results 7-156
7.3.4 Reviews of the TSPA-VA 7-172
7.3.5 NRC Total System Performance Assessments 7-181
7.3.6 EPRI Total System Performance Assessments 7-192
7.3.7 Comparison of DOE, NRC, and EPRI TSPA Results
for the VA Repository 7-200
7.3.8 Performance Assessments in the Yucca Mountain DEIS 7-202
7.3.9 Preliminary TSPA Results for the EDA II Design 7-209
7.3.10 Performance Evaluation for the Site Recommendation 7-217
7.3.11 Uncertainties in Projecting Repository Performance over Very
Long Time Periods 7-234
References 7-247
IV
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TABLE OF CONTENTS (Continued)
Section
CHAPTER 8 RADIOLOGICAL PATHWAYS THROUGH THE BIOSPHERE .8-1
8.1 Introduction 8-1
8.2 Past, Current, and Potential Use of the Yucca Mountain Region 8-3
8.2.1 Past Use of the Yucca Mountain Region 8-3
8.2.2 Current Demographics and Land Use 8-13
8.2.3 Factors Affecting Future Use of the Region 8-16
8.3 Radiation Protection of Individuals 8-46
8.3.1 The Critical Group Concept 8-47
8.3.2 Probabilistic Scenario Modeling 8-49
8.3.3 Exposed Individuals and Exposure Scenarios for Yucca Mountain .. 8-52
8.3.4 Details and Analyses for the Subsistence Farmer Scenario 8-54
8.3.5 Alternative Exposure Scenarios for Consideration
at Yucca Mountain 8-70
8.4 The Repository Intrusion Scenario: a Special Case 8-74
8.4.1 Site Resources as Potential Cause for Intrusion ,-•••' 8-75
8.4.2 Types of Human Intrusion 8-81
8.4.3 Parameters and Assumptions Associated with
Ground Water Withdrawal ,. 8-87
8.4.4 Parameters and Assumptions Associated with Human Intrusion .... 8-88
References 8-98
CHAPTER 9 YUCCA MOUNTAIN EXPOSURE SCENARIOS
AND COMPLIANCE ASSESSMENT ISSUES 9-1
9.1 Introduction 9-1
9.2 Gaseous Releases: a Secondary Pathway for Human Exposure 9-3
9.2.1 Production and Early Containment of Carbon-14 9-4
9.2.2 Impacts of Thermal Loading on Gaseous Releases and Transport .... 9-4
9.2.3 Estimates of Travel Time 9-5
9.2.4 Dose Modeling and Exposure Estimates 9-7
9.2.5 Dose Estimates from Repository Releases 9-8
9.2.6 Potential Non-radiological Impacts of C-14 9-10
9.3 Development of Performance Scenarios and Compliance Issues 9-11
9.3.1 Identification of Improbable Phenomena 9-11
9.3.2 Screening of Events and Processes 9-12
9.3.3 Compliance With a Standard 9-14
9.3.4 Development of Site Performance Issues 9-17
References 9-27
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TABLE OF CONTENTS (Continued)
CHAPTER 10 RADIOLOGICAL RISKS FOR DEEP GEOLOGICAL DISPOSAL
AND SURFACE STORAGE OF SPENT NUCLEAR FUEL 10-1
10.1 Background Information 10-1
10.2 Regulatory Limits 10-2
10.2.1 Power Reactors 10-3
10.2.2 Research Reactors 10-4
10.2.3 Independent- Spent Fuel Storage Installations (ISFSIS) 10-4
10.2.4 DOE Facilities 10-5
10.2.5 Summary of Regulatory Limits 10-5
10.3 Report by the Monitored Retrievable Storage Review Commission 10-6
10.3.1 At-Reactor Storage Options '..... 10-6
10.3.2 Radiation Exposure Modeling Assumptions
for At-Reactor Storage of SNF ..'... 10-8
10.3.3 • Model Assumptions for MRS Storage of SNF 10-10
10.3.4 Transportation Models for SNF With and Without MRS 10-10
10.3.5 Public Exposure from SNF Storage 10-11
10.4 Other Information Sources 10-14
10.4.1 "An Assessment ofLWRS Spent Fuel Disposal Options" 10-15
10.4.2 "Generic Environmental Impact Statement, Management of
Commercially Generated Radioactive Waste" 10-16
10.4.3 "Review of Dry Storage Concepts Using Probabilistic Risk
Assessment" _.. 10-16
10.4.4 "Requirement for the Independent Storage of Spent Fuel and
High-Level Radioactive Waste" 10-17
10.4.5 "Environmental Assessment Related to the Construction and Operation of
the Surry Dry Cask Independent Spent Fuel Storage Installation" . . 10-18
10.4.6 "Environmental Assessment Deaf Smith County Site, Texas" 10-18
• 10.4.7 "Preliminary Assessment of Radiological Doses in Alternative Waste
Management Systems Without an MRS Facility" 10-19
10.4.8 "Monitored Retrievable Storage Submission to Congress" 10-20
10.4.9 "The Safety Evaluation of Tunnel Rack and Dry Well Monitored
Retrievable Storage Concepts" 10-21
10.4.10 Summary Assessment of Available Data 10-22
References 10-25
VI
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TABLE OF CONTENTS (Continued)
Section
GLOSSARY
age
G-l
APPENDICES
I.
H.
III.
IV.
V.
VI.
i
Demography and Ecosystems ..............................
Radionuclide Exposures to Persons in the Vicinity of the Nevada Test Site/Yucca
Mountain Site .................................
Soil Types Found in the Yucca Mountain Area ...................... '
Well Drilling and Pumping Costs ................................ jy_j
New and Unusual Farming Practices ................................. y_j
Current Information Regarding Ground Water Flow and Radionuclide Transport in the
Unsaturated and Saturated Zones . . .............. . ................ yj_ j
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TABLES
-
ES-1 Programs for HLW and SNF Disposal in Other Nations ES-11
1-1 Significant Events in the History of High-Level Radioactive Waste and Spent Nuclear
Fuel Disposal :
3-1 National and International Criteria and Objectives for the Disposal of Long-Lived
Radioactive Wastes 3"4
5-1 Historical and Projected Mass and Radioactivity of Commercial Spent Nuclear Fuel . 5-3
5-2 Historical and Projected* Installed Nuclear Electric Power Capacity 5-4
5-3 DOE Spent Nuclear Fuel Inventory 5"8
5-4 Historical and Projected Cumulative Volume and R adioactivity of High-Level Waste
Stored in Tanks, Bins, and Capsules By Site 5-10
5-5 Radionuclide Inventories in Spent Nuclear Fuel and High-Level Wastes
Expected to be Disposed in a Yucca Mountain Repository 5-15
6-1 Estimated Frequency of Genetic Disorders in a Birth Cohort Due to Exposure of
Each of the Parents to 001 Gy (1 rad) per Reproductive Generation (30 yr) 6-6
6-2 Possible Effects of In Utero Radiation Exposure 6-9
7-1 Stratigraphy of the Southern Great Basin -. 1~11
7-2 Principal Stratigraphic Units 7"15
7-3 Known or Suspected Quaternary Faults within 20 km of the Proposed Repository
Site • 7-29
7-4 Significant Earthquakes within 320 km of Yucca Mountain Site Since 1850 7-35
7-5 Hydraulic Conductivities Calculated from Pumping Test Data 7-85
7-6 Borehole Location and Depth Data for Wells Drilled to the Lower Carbonate
Aquifer in the Vicinity of and Downgradient of the Yucca Mountain Area 7-91
7-7 Design Parameters for the Enhanced Design Alternatives ,. • 7-135
7-8 Principal Results of Enhanced Design Alternative Analyses 7-135
7-9 Comparison of EDA II and Viability Assessment Design Features 7-137
7-10 Principal Performance Factors for TSPA-VA Modeling 7-149
7-11 DEIS Estimates of Waste Emplacement Areas 7-203
7-12 Peak Dose Rates at 10,000 Years for the Proposed Action Inventory
and Alternative Distances and Thermal Loads 7-206
7-13 Comparison of DEIS Ground Water Concentrations With MCLs 7-209
7-14 Impact of EDA II Design Features on VA Performance Uncertainties
Identified by Reviewers of the TSPA-VA • 7-211
7-15 Factors Potentially Important to Postclosure Safety 7-214
7-16 Comparison of Major Dose Milestones for the VA and EDA H Repositories 7-217
vm
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TABLES (Continued)
No.
7-17 Comparison of Rev 3 and Rev 4 Repository Safety Strategies 7-218
7-18 Potential Performance Assessment Vulnerabilities and Mitigation Measures 7-219
7-19 Implementation of Regulatory Requirements in the TSPA-SR for Regulatory
Requirements (Excerpted from TRWOOb) 7-221
7-20 Technical Assumptions Implemented in the Human Intrusion Scenario in TSPA-SR
(Table excerpted from TRWOOa) 7-223
8-1 Range in Concentration of Dissolved Constituents in Ground Water in the
AmargosaDesert g_jo
8 -2 Hydrographic Basins in the Vicinity of Yucca Mountain g_20
8 -.1 Water Appropriations by Hydrographic Basin in the Study Are? 8-23
8-4 1993 Ground Water Pumpage Inventory for Basin No 230 \\" g-23
8-5 Wells and Boreholes in the Amargosa Valley ' 3.27
8-6 Ground Water Budget for Hydrographic Basins in Study Area g-36
8-7 Estimates of Acreage Under Cultivation for Feedstock g-40
8-8 Ground Water Storage Values for Relevant Hydrographic Basins ['.'. 8-41
8-9 Concentration Ratios and Transfer Coefficients By Element 8-61
8-10 Dose Conversion Factors for a Resident Farmer in Current Biosphere By Exposure
Pathway and Radionuclide for Ground Water Source g_66
8-11 Summary of Mean TEDE Results From CNWRA .
Unit Concentration Evaluations for Water g_gy
8-12 Comparison of Inhalation, Drinking Water and Food Consumption Rate
Parameter Values From Various Sources g_gg
8-13 Likely Human Intrusion Scenarios for Different Types of Resources '. 8-82
8-14 Typical Borehole Characteristics g_90
9-1 Annual Average Doses Resulting from the Release of 100 Ci 14CO2
for Distances Out to 50 Miles g_ JQ
9-2 Potentially Disruptive Events and Processes 9_13
9-3 Techniques for Quantifying or Reducing Uncertainty in the Performance Assessment 9-16
10-1 Spent Fuel Accumulation at Shutdown Commercial Light Water Power Reactors 10-9
10-2 Reduction in Dry Storage Needs At Reactor Facilities with Linked MRS 10-10
10-3 Life-Cycle Transportation Risk Measures 10-11
10-4 Total Life-Cycle Doses in Person-Rem from Spent Nuclear Fuel Management With
and Without MRS 10-12
10-5 Location of Spent Fuel With MRS in 2010 and Repositoryin 2013 (MTUs) ...... 10-13
10-6 Comparison of Public Exposures Resulting from Three SNF Storage Alternatives 10-14
10-7 Public Doses for Normal Repository Operation and From Shaft-Drop Accident ... 10-15
IX
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TABLES (Continued)
No.
10-8 At-Reactor Storage Accidents: Summary of Results (ORV84) 10-17
10-9 Preclosure Exposure Associated with a Reference Salt Repository 10-19
10-10 Public-Dose Estimates for Reference Reactor and Repository Surface Facility
(Based on data from SCH86) 10-20
10-1 la Public Doses From Routine Operations at MRS and Repository 10-21
10-1 Ib Public Doses from Accidental Releases at MRS and Repository 10-21
10-12 Normalized Population Doses 10-22
10-13 Summary Data of Public Doses Associated with SNF Storage At-Reactor, MRS,
and Repository
10-24
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FIGURES
No.
ES5-1 Schematic North/South Cross-Sectional Illustration of Thinning of Volcanic Units
Beneath the Amargosa Desert ES-25
ES5-2 Repository Layout for the TSPA-VA Design ES-31
ES5-3 Waste Package for 21 PWR Uncanistered Fuel Assemblies ES-32
ES6-1 Schematic Illustration of the Major Pathways from a Repository at Yucca Mountain
to Humans ES-38
7-1 Location of Yucca Mountain 7_2
7-2 Boundaries and Larger Subdivisions of the Basin and Range Physiographic Province 7-3
7-3 Physiographic Features in the Yucca Mountain Site Area 7-5
7-4 Generalized Regional Stratigraphic Column Showing Geologic Formations and
Hydrological Units in the Nevada Test Site Area 7.7
7-5 Late Precambrian Through Mid-Paleozoic Paleography of the Great Basin 7--8
7-6 Late Devonian and Mississippian Paleogeography of the Great Basin 7-9
7-7 Simplified Geologsc Map Showing the Distribution of Major Lithostratigraphic
Units in the Yucca Mountain Area l-{2
7-8 East-West Geologic Cross Section forthe Yucca Mountain Site 7-14
7-9 The Walker Lane Belt and Major Associated Faults 7-23
7-10 Major North-Trending Faults in the Vicinity of Yucca Mountain 7-25
7-11 Index Map of Faults at and near Yucca Mountain 7-26
7-12 Index Map of Known or Suspected Quaternary Faults in the Yucca Mountain Region 7-28
7-13 Sketch Map of the Western United States Showing Some Major Structural Features 7-31
7-14 Magnitude 3 or Greater Earthquakes Within 320 Km (200 Miles) of Yucca Mountain
from 1850 to 1992 7.37
7-15 Index Map Showing Outlines of Calderas in the Southwestern Nevada Volcanic Field and
the Extent of the Tiva Canyon and Topopah Spring Tuffs of the Paintbrush Group.. 7-43
7-16 Distribution of Basalts in the Yucca Mountain Region with Ages Less Than 12 MA ..7-45
7-17 Unsaturated Zone Hydrogeologic Units (USG84a) 7.59
7-18 Locations of Deep Boreholes in the Vicinity of Yucca Mountain 7-62
7-19 Early Conceptual Model of Ground Water Flow in the Unsaturated Zone at Yucca
Mountain 7-70
7-20 Current Conceptual Model of Ground Water Flow in the Unsaturated Zone at
Yucca Mountain 7.71
7-2.1 Saturated Zone Hydrostratigraphy of Volcanic Rocks 7-80
7-2,2 Schematic North/South Cross-Sectional Illustration of Thinning of Volcanic Units
Beneath the Amargosa Desert 7-83
7-23 Schematic Illustration of Ground Water Flow System in the Great Basin 7-99
7-24 Death Valley Ground Water Flow System 7-100
7-25 Alkali Flat-Furnace Creek Ranch Ground Water Subbasin 7-102
7-26 Potentiometric Surface in the Death Valley Ground Water Flow System 7-105
XI
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FIGURES (Continued)
No.
7-27 Potentiometric Surface in the Amargosa Desert 7-108
7-28 Future Climates, Expressed in Terms of Overall Global Temperature Change
7-117
7-29 Model Simulations of Past and Future Climate Conditions 7-119
7-30 Repository Layout for the VA Reference Design 7-124
7-31 Repository Location Within Yucca Mountain 7-125
7-32 North Portal Facilities Layout for the VA Reference Design 7-126
7-33 21-PWR Waste Package Design for the VA Reference Design 7-128
7-34 Defense HLW Package Design for the VA Reference Design 7-129
7-35 Drift Cross-Section for the VA Reference Design 7-130
7-36 Computer Code Configuration for the TSPA-VA 7-151
7-37 TSPA-VA Base Case Dose Rates for Periods Up to 10,000 Years (DOE98)
7-158
7-38 TSPA-VA Base Case Dose Rates for Periods Up to 100,000 Years 7-158
7-39 TSPA-VA Base Case Dose Rates for Periods Up to One Million Years 7-160
7-40 Uncertainties in the TSPA-VA Base Case Results 7-162
7-41 Structure of Performance Factors for NRC Performance Assessments
7-182
7-42 Structure of NRC Computer Codes for Performance Assessments 7-183
7-43 NRC TSPA Results for Alternative Conceptual Models 7-188
7-44 NRC'TSPA Results for Mean-Values Data Set • 7-190
7-45 EPRTs IMARC Logic Tree 7-193
7-46 Results of EPRI's IMARC-4 Dose Evaluations 7-199
7-47 Comparison of DOE, NRC, and EPRI Performance Assessment Results 7-201
7-48 Emplacement Block Layout for DEIS Disposal Option 7-204
7-49 Time History of Projected Dose to 10,000 Years, VA and
DEIS Evaluations 7-207
7-50 DEIS Dose Rate Time Histories for Periods Up to One Million Years 7-207
7-51 Barriers Importance Analysis to Assess Natural Barriers of the
Repository System-Early Waste Package Failure Scenario 7-215
7-52 Comparison of VA and EDA 10,000-Year Doses 7-215
7-53 Comparison of VA and EDA Million-Year Doses 7-216
7-54 Comparison of Proposed Radiation Protection Standards with Expected Values of
TSPA-SR Calculations for a Repository at Yucca Mountain for Nominal and
Igneous Scenarios ( Figure adapted from TRWOOb) 7-226
7-55 Expected Values of TSPA-SR Calculations for a Repository at Yucca Mountain
for the Inadvertent Human Intrusion Scenario (Figure adapted from TRWOOb) ... 7-227
7-56 Summary of Groundwater Protection Performance Results of the TSPA-SR:
Combined Beta and Photon-Emitting Radionuclides (Figure adapted from TRWOOb)
7-228
xn
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No.
FIGURES (Continued)
7-57 Summary of Groundwater protection Results for TSPA-SR for Gross Alpha Activity
(Figure adapted from TRWOOb) 7-228
8-1 Schematic Illustration of the Maj or Pathways from a Repository
at Yucca Mountain to Humans 8-2
8-2 Yucca Mountain and Surrounding Land Use 8-4
8-3 Winter Sites Near Beatty and Belted Range 8-7
8-4 Major Winter Sites hi Ash Meadows and Pahrump Valley 8-8
8-5 Major Winter Sites in Northern and Central Death Valley 8-9
8-6 Map Showing Boundaries of Ground Water Subbasins in the Study Area 8-20
8-7 Ground Water Usage in the Amargosa Desert 8-21
8-8 Locations of Water Wells in the Amargosa Farms Area 8-26
8-9 Wells and Boreholes in the Amargosa Valley
Only 15 Persons Currently live at the 20 km Distance 8-33
8-10 Examples of Current Agriculture Activities in the Yucca Mountain Region 8-45
8-11 Ground Water Pathway Model for Subsistence Farmer 8-56
9-1 Schematic Illustration of the Major Pathways from a Repository
at Yucca Mountain to Humans (NAS95) 9-2
9-2 Retarded Travel Times of C-.14 from the Repository to the Atmosphere for Particles
Released at 1,000 Years 9-6
9-3 Retarded Travel Time of C-14 Particles from the Repository to the Atmosphere for
Particles Released at 10,000 Years , 9-6
9-4 Annual Average Concentration for Uniform Continuous Source and Specific Activity
(in Parentheses) for 100 Ci/year 9-9
9-5 An Illustration of Hypothetical Individual Dose Rates Associated with a Disruptive
Event Happening at Two Different Tunes after Disposal of Radioactive Waste 9-15
Xlll
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LIST OF ACRONYMS
ACHP Advisory Council On Historic Preservation
AEA Atomic Energy Act
AEC Atomic Energy Commission
AECB Canadian Atomic Energy Control Board
AECL Atomic Energy of Canada Limited
AFCN Belgian Nuclear Inspection Agency
AGNEB Swiss Interagency Working Group on Licensing of Nuclear Waste Facilities
AGR Advanced Gas-Cooled Reactor
AIRFA American Indian Religious Freedom Act
ALARA As Low As is Reasonably Achievable
ALI Annual Limit on Intake
ANDRA French Radioactive Waste Management Agency
ANL-W Argonne National Laboratory - West
Bq Becquerel
BEW Swiss Energy Office
BfS German Institute for Radiation Protection
BID Background Information Document
BMFT German Ministry for Research and Technology
BMU German Ministry for Environment, Protection of Nature and Reactor
Safety
BNFL British Nuclear Fuels Limited
xiv
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BRGM French Bureau of Geological and Mineral Research
BSS Basic Safety Standards
BWR Boiling Water Reactor
CAA Clean Air Act
CCDF Complementary Cumulative Distribution Function
CEA French Atomic Energy Commission
CEC Council of the European Communities
CEDE Committed Effective Dose Equivalent
CEN Belgian Nuclear Research Center
CFR Code of Federal Regulations
Ci Curie
CLAB Swedish Centralized Spent Fuel Storage Facility
CNWRA Center for Nuclear Waste and Regulatory Analysis
CRPPH Committee on Radiation Protection and Public Health
CRWM Committee on Radioactive Waste Management
DACs Derived Air Concentrations
DCF Dose Conversion Factors
DDREF . Dose, Dose Rate Effectiveness Factor
DOD U.S. Department of Defense
DOE U.S. Department of Energy
DSIN French Directorate for the Safety of Nuclear Installations
EBS Engineered Barrier System
xv
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EDE Effective Dose Equivalent
EDI Swiss Department of Interior
EIA Environmental Impact Assessment
EIR Swiss Institute for Reactor Research
EIS Environmental Impact Statement
EMSL-LV Environmental Monitoring Systems Laboratory - Las Vegas
EnPA Energy Policy Act
EPA U.S. Environmental Protection Agency
EPRI Electric Power Research Institute
ERA Energy Reorganization Act
EURATOM. European Atomic Energy Community
EVED Swiss Department of Transport, Communications, and Energy
FEIS • Final Environmental Impact Statement
FERC Federal Energy Regulatory Commission
FFTF Fast Flux Test Facility
FRC Federal Radiation Council
GTCC Greater-Than-Class-C
Gy Gray
GW (e) Gigawatt - Electric
HEU Highly Enriched Uranium
HI Human Intrusion
HLW High-Level Waste
xvi
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HSK Swiss Nuclear Safety Division
HTGR High-Temperature Gas-Cooled Reactor
IAEA International Atomic Energy Agency
ICPP Idaho Chemical Processing Plant
ICRP International Commission on Radiological Protection
IMARC Integrated Multiple Assumptions and Release Calculations
INEEL ' Idaho National Engineering and Environmental Laboratory
IPA Iterative Performance Assessment
IPSN French Institute for Nuclear Protection and Safety
IRG Interagency Review Group
IRSR Issue Resolution Status Report
JAERI Japan Atomic Energy Research Institute
JNFL Japan Nuclear Fuel Services Limited
KASAM Swedish Consultative Committee for Nuclear Waste Management
KSA Swiss Commission for the Safety of Nuclear Installations
KTI Key Technical Issue
LET Linear Energy Transfer
LMFBR Liquid-Metal Fast-Breeder Reactor
LWR Light Water Reactor
MCLs Maximum Contaminant Levels
MCLG Maximum Contaminant Level Goal
MFRP Midwest Fuel Recovery Plant
xvii
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MITI Japanese Ministry of International Trade and Industry
MFC Multi-Purpose Canister
mrem Millirem
MRS Monitored Retrievable Storage
mSv Millisieverts
MTHM Metric Tons of Heavy Metal
MTIHM Metric Tons of Initial Heavy Metal
MWd Megawatt Days
NAGRA Swiss Cooperative for the Storage of Radioactive Waste
NAS National Academy of Sciences
NCI National Cancer Institute
NCRP National Council on Radiation Protection and Measurements
NEPA National Environmental Policy Act
NHPA National Historic Preservation Act
NIREX British Nuclear Industry Radioactive Waste Executive
NPRM Notice of Proposed Rulemaking
NRC U.S. Nuclear Regulatory Commission
NRPB . National Radiological Protection Board of the United Kingdom
NSC Japanese Nuclear Safety Commission
NTS Nevada Test Site
NUCEF Japanese Nuclear Fuel Cycle Engineering Facility
NWPA Nuclear Waste Policy Act
xviii
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NWPAA Nuclear Waste Policy Amendments Act
NWPO Nuclear Waste Project Office
NWTRB Nuclear Waste Technical Review Board
OECD/NEA Organization for Economic Cooperation and Development/Nuclear Energy Agency
OCRWM Office of Civilian Radioactive Waste Management
OMB Office of Management and Budget
ONDRAF Belgian Agency for Radioactive Waste and Fissle Materials
ORERP Off-Site Radiation Exposure Review Project
ORNL Oak Ridge National Laboratory
ORSP Off-Site Radiological Safety Program
PA Programmatic Agreement
PARCLAY Belgian Preliminary Demonstration Test for Clay Disposal
PBF Power Burst Facility
PICs Pressurized Ion Chambers
PNC Japanese Power Reactor and Nuclear Fuel Development Corporation
PPA Project Programmatic Agreement (Yucca Mountain)
PUREX Plutonium-Uranium Extraction
PWR Pressurized Water Reactor
QAP Quality Assurance Plan
R Roentgen
R&D Research and Development
RADWASS Radioactive Waste Safety Standards Radioactive Waste Management
xix
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RBE Relative Biological Effectiveness
rem Roentgen Equivalent Man
RBOF Receiving Basin for Off-site Fuels
RETF Japanese Recycling Equipment Testing Facility
RMEI Resaonably, Maximally Exposed Individual
ROD Record of Decision
RSK • German Reactor Safety Commission
SAB Science Advisory Board
SAR Safety Analysis Report
SCP Site Characterization Plan
SDWA Safe Drinking Water Act
SON French Agency Providing Architectural and Engineering Services
SHP Japanese Steering Committee on High-Level Radioactive Waste
'8KB Swedish Nuclear Fuel and Waste Management Company
SKI Swedish Nuclear Power Inspectorate
SKN Swedish Board for Spent Nuclear Fuel
SNF Spent Nuclear Fuel
SNTZ Southern Nevada Transverse Zone
SRS Savannah River Site
SSI Swedish Institute for Radiation Protection
SSK German Committee on Radiological Protection
STA Japanese Science and Technology Agency
xx
-------
Sv
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IBM
TEDE
THORP
TLD
TRU
TSPA
TSPAI
UK
UNSCEAR
URL
USDW
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WL
WLM
WIPP LWA
WVDP
YMS
ZWILAG
Sievert
Saturated Zone
Tunnel Boring Machine
Total Effective Dose Equivalent
Thermal Oxide Reprocessing Plant
Thermoluminescent Dosimeter
Transuranic
Total System Performance Assessment
Total System Performance Assessment and Integration
United Kingdom
United Nations Scientific Committee on the Effects of Atomic Radiation
Underground Research Laboratory
Underground Sources Drinking Water
Unsaturated Zone
Viability Assessment
Working Level
Working Level Month
Waste Isolation Pilot Plant Land Withdrawal Act
West Valley Demonstration Project
Yucca Mountain Site
Swiss Cooperative of Nuclear Utility Operators
xxi
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EXECUTIVE SUMMARY :
ES.l INTRODUCTION AND BACKGROUND
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 from waste stored or disposed in Yucca
Mountain, Nevada. The Yucca Mountain site is located in Nye County, approximately
150 kilometers (90 miles) northwest of Las Vegas, Nevada, and on the southwestern boundary of
the Nevada Test Site. Yucca Mountain is an irregularly shaped elevated land mass six to 10 km
wide (four to six miles) and about 40 km (25 miles) long. A waste repository would be about
300 meters (one thousand feet) below the crest of Yucca Mountain and about the same distance
above the water table under the mountain.
The EPA is promulgating, in 40 CFR Part 197, site-specific environmental standards to protect
the public from releases of radioactive materials disposed of or stored in the potential repository
to be constructed at Yucca Mountain.1 These standards provide the basic framework to control
the long-term storage and disposal of three types of radioactive waste:
• Spent nuclear fuel (SNF), if disposed of without reprocessing
• High-level radioactive waste (HLW) from the reprocessing of spent nuclear fuel
• Other radioactive materials that may be placed in the potential repository
The 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. However, the plans for
placement of these materials are uncertain and therefore, for the purpose of the present
rulemaking, the information presented in this Background Information Document (BID) is
limited to spent nuclear fuel and high-level radioactive waste.
It is important to note that no decision has been made regarding the acceptability of Yucca Mountain for
storage or disposal. However, for the purposes of this document, the description of Yucca Mountain as "potential"
will generally not be used but is intended.
ES-1
-------
ES.1.1 Purpose And Scope of The Background Information Document
The BED presents the technical information used by EPA to understand the characteristics of the
Yucca Mountain site and to develop its rule, 40 CFR Part 197. Most of the technical information
discussed in the BID is derived from investigations sponsored by the Department of Energy
(DOE). However, where appropriate, information from other sources, such as the Electric Power
Research Institute (EPRI) and U.S. Nuclear Regulatory Commission (NRC), and Nevada state
and local agencies is presented to supplement the DOE data base, to fill data gaps, and to
illustrate alternative conceptualizations of geologic processes and engineered barrier
performance.
The scope of the BID encompasses the conceptual framework employed by the Agency for
assessing radiation exposures and associated health risks. In general terms, this assessment
discusses the radioactive source term characterization, movement of radionuclides from the
repository at Yucca Mountain through the appropriate environmental exposure pathways, and the
estimates of potential doses to members of a representative group of people living in the region
around the repository site. It is not intended to be a technical critique of the investigations
conducted by DOE and other parties. Nor is it a regulatory compliance or criteria document.
The BID is simply a summary of the technical information considered by EPA in developing the
rationale for, and specifics in, 40 CFR Part 197.
This executive summary highlights key chapters of the BID, particularly information concerning
efforts in other nations to develop deep geological repositories (Chapter 3); current efforts to
develop a repository at Yucca Mountain (Chapter 4); the types and inventories of waste likely to
be disposed in Yucca Mountain (Chapter 5); geologic and hydrogeologic characteristics of the
repository site and anticipated repository performance (Chapter 7); and pathways for human
exposure to radionuclides potentially released from the site (Chapter 8). The reader is referred to
the full text of the BID for information regarding ways in which radiological dose and risk are
estimated (Chapter 6); potential exposure scenarios and compliance assessment issues for the
Yucca Mountain repository (Chapter 9); and the comparative radiological risks associated with
deep geological disposal and'surface storage of spent nuclear fuel (Chapter 10).
ES-2
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ES.1.2 EPA's Regulatory Authority For The Rulemaking
The standards governing environmental releases from the potential Yucca Mountain repository
have been developed pursuant to the Agency's responsibilities under the Energy Policy Act
(EnPA) of 1992 (Public Law 102-486). Section 801 of this Act directed EPA to promulgate
standards to ensure protection of public health from releases of radioactive material in a deep
geologic repository to be built in Yucca Mountain by setting standards to protect individual
members of the public. The EnPA also required EPA to contract with the National Academy of
Sciences (NAS) to advise the Agency on the technical bases for the Yucca Mountain standards.
The EnPA directed that these standards will apply only to the Yucca Mountain site and are to be
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.
ES.1.3 The National Academy of Sciences Recommendations
In the EnPA, the Congress asked 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 of10,000 years.
ES-3
-------
To address these questions, the Academy assembled a committee of 15 persons 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 NRC,
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. In this
report, the committee addressed the key issues posed by Congress and reached the following
conclusions:
• ...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.
• ...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 of individual members of
the public to radiation beyond allowable limits.
• ...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.
In addition, the report offered the Agency several general recommendations as to the approach
EPA should take in developing 40 CFR Part 197. Specifically, the NAS recommended:
...the use of a standard that sets a limit on the risk to individuals of
adverse health effects from releases from the repository.
• ...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.)
• ...that the consequences of an intrusion be calculated to assess the
resilience of the repository to intrusion.
ES-4
-------
The EPA does not believe it is bound to adopt all of the positions advanced by the NAS in the
Yucca Mountain rulemaking. The Agency has used the NAS report as the foundational starting
point for the rulemaking. The Agency has carefully considered the recommendations of the
NAS, but the role of the NAS recommendations is not to replace the rulemaking authority of the
Agency. The Agency will tend to accord greatest deference to the judgements of NAS about
issues having a strong scientific component, the area where NAS has its greatest expertise. The
EPA will reach final determinations that are congruent with the NAS analysis whenever it can do
so without departing from the Congressional delegation of authority to promulgate, by rule,
health and safety standards for protection of the public. The Agency believes that such
determinations require the consideration of public comments and the Agency's own expertise
and discretion.
ES.1.4 Prior Agency Action
In December 1976, EPA announced its intent to develop environmental radiation protection
criteria for radioactive waste to ensure the protection of public health and the general
environment. These efforts resulted in a series of radioactive waste disposal workshops, held in
1977 and 1978. Based on issues raised during workshop deliberations, EPA published a Federal
Register notice on November 15, 1978 of intent to propose criteria for radioactive wastes and to
solicit public comments on possible recommendations for Federal Radiation Guidance. In March
1981, EPA withdrew the proposed "Criteria for Radioactive Wastes" because it considered the
implementation of generic disposal guidance too complex given the many different types of
radioactive waste.
In 1982, Congress enacted the Nuclear Waste Policy Act (NWPA), which established the current
national program for the disposal of SNF andHLW. The Act assigned to DOE the responsibility
of siting, building, and operating an underground geologic repository for the disposal of these
wastes and directed the EPA to "promulgate generally applicable standards for the protection of
the general environment from off-site releases from radioactive material in repositories." In that
same year, under the authority of the Atomic Energy Act (AEA), the EPA proposed a set of
standards under 40 CFR Part 191, "Environmental Standards for the Management and Disposal
of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes." After a number of
public hearings and comment periods, the EPA issued the final rule under 40 CFR Part 191 on
August 15, 1985. Sections of this rulemaking were remanded by a Federal Court in 1987 and
repromulgated by EPA in 1993.
ES-5
-------
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 the Yucca Mountain site as the location for the first SNF and HLW repository.
Activities at all other potential sites were to be phased out.
In October 1992, the Waste Isolation Pilot Plant Land Withdrawal Act (WIPP LWA) was
enacted. While reinstating certain sections of the Agency's 1985 disposal standards, the Act had
the effect of exempting the Yucca Mountain site from these generic disposal standards.
However, also in October 1992, the EnPA directed the EPA to promulgate site-specific radiation
protection standards for the Yucca Mountain disposal system.
ES.2 CURRENT U.S. PROGRAMS FOR YUCCA MOUNTAIN
The DOE, NRC, and EPA each have legislatively defined roles in the management and disposal
of SNF and HLW at the proposed Yucca Mountain disposal site. As stated in the NWPA, DOE
is responsible for developing, constructing, and operating repositories for disposal of these
wastes. The NRC has responsibility to license the repository and related facilities, and EPA is to
promulgate radiation protection standards which the NRC is to adopt as the basis for its licensing
actions. Affected state and local governments and Native American tribes have an oversight role
in the program. The NWPAA designated the Yucca Mountain site in Nevada as the only site to
be evaluated by DOE as a potential location for disposal of SNF and HLW, and established the
Nuclear Waste Technical Review Board (NWTRB) to provide oversight of the DOE program.
ES.2.1 The Department of Energy
The DOE's Office of Civilian Radioactive Waste Management (OCRWM) was established by
Congress in the NWPA specifically to provide management for the disposal of SNF from
commercial nuclear power reactors. Under a 1985 Presidential Executive Order, the repository
established by DOE is also to be used for disposal of HLW from DOE operations. The OCRWM
charter includes responsibility for receipt of SNF from reactors at the reactor sites; interim
storage of received SNF, as necessary, prior to disposal; transport of SNF to the site(s) for
interim storage and disposal; and siting, design, licensing, and operation of a central interim
storage facility and disposal facilities. In addition to its work at Yucca Mountain, DOE has
developed alternative designs for a central interim storage facility (known historically as a
ES-6
-------
Monitored Retrievable Storage (MRS) facility), but, as of March 2000, the Department has not
established a site for such a facility.
In accord with the NWPAA, DOE has been evaluating Yucca Mountain as the disposal site for
SNF and HLW. Characterization of the site is proceeding with surface-based and sub-surface
activities. Design concepts for the engineered features of the repository are being developed.
Recent DOE activities produced the Viability Assessment (VA) report, which is a
Congressionally mandated appraisal of the viability of the Yucca Mountain project for geologic
disposal of nuclear wastes. The VA report contains: ;
. A site description and a design for engineered features of the repository and waste
package
A Total System Performance Assessment, based on available data, describing the
probable safety performance of the VA reference design (TSPA-VA)
A plan and cost estimate for completing the license application (LA) to NRC for
repository construction
Cost estimates for constructing and operating the repository
The VA report was published in December 1998. It was followed by a draft environmental
impact statement (DEIS) in 1999 and a final EIS is planned for late in 2000. The site-suitability
recommendation, required by the NWPA, is planned to be submitted to the President in 2001 and
a License Application (LA) would be submitted to NRC in 2002 if the site is found suitable for
disposal. Significant recent accomplishments of the DOE program include:
Completion of the Exploratory Studies Facility (ESF) and Cross Drift tunnels for
gathering experimental data at the proposed repository horizon
Initiation of various types of experiments in the tunnel alcoves and niches
Completion of the Viability Assessment in December 1998
Issuance of the DEIS in August 1999
Selection of an improved repository design, EDA H, based on TSPA-VA results
ES-7
-------
ES.2.2 The Nuclear Regulatory Commission
The NRC is responsible for licensing and regulating the receipt and possession of SNF and
HLW, at privately owned facilities and at certain facilities managed by DOE. This licensing
responsibility includes the waste management and disposal facilities at Yucca Mountain. The
NRC currently licenses temporary storage facilities at reactor sites, as well as commercial storage
facilities at West Valley, New York, and Morris, Illinois.
NRC licensing of a repository at Yucca Mountain will be accomplished through review of a
License Application (LA) submitted by DOE after completion of site approval procedures set
forth in the NWPA. If the LA is found acceptable for review, NRC would review it to determme
if there is reasonable assurance of compliance with regulatory standards. If expectation of
compliance is established, DOE will be authorized to construct the repository. Subsequently, the
LA will be amended to seek approval to receive and emplace wastes for disposal. Confirmatory
testing is expected to continue throughout construction and disposal operations. After disposal is
completed (a process expected to span about 50 years), the LA would be amended to request
closure of the repository. After closure is authorized, post-closure monitoring would be expected
to be required.
The NWPA requires both EPA and NRC to publish radiation-protection standards and
regulations, for the storage and disposal of HLW. As previously noted, the EnPA directed the
EPA to develop radiation protection standards for the Yucca Mountain site and for the NRC to
develop implementing regulations that conform to the EPA's Yucca Mountain standards. The
NRC's proposed (February 1999) 10 CFR Part 63 regulations require use of multiple barriers
(natural and engineered) to achive compliance with regulatory standards, and implement the
Commission's principles of defense-in-depth and risk-informed regulation. The proposed rule
addresses licensing procedures, radiation exposure standards, criteria for public participation,
records and reporting, monitoring and testing programs, performance confirmation, quality
assurance, personnel training and certification, and emergency planning.
The NRC's proposed 10 CFR Part 63 regulations would be modified as necessary to conform to
EPA's 40 CFR Part 197 standards after they are established.
ES-8
-------
ES.2.3 Nuclear Waste Technical Review Board
The NWPAA established the Nuclear Waste Technical Review Board comprised of eleven
members recommended by the NAS and appointed by the President. These individuals are
experts in the fields of science, engineering, or environmental sciences and represent a broad
range of scientific and engineering disciplines, including hydrology, underground construction,
hydrogeology, and physical metallurgy. No member of the Board may be employed by DOE, its
contractors, or the National Laboratories. The current Board is composed of individuals with
academic and public and private sector experience. The Board's mandate is to evaluate the
technical and scientific validity of activities undertaken by DOE, regarding various aspects of the
U.S. SNF and HLW management. For example, the NWTRB provided comments in April 1999
on DOE's Viability Assessment in a report entitled Moving Beyond the Yucca Mountain Viability
Assessment - A Report to the U.S. Congress and the Secretary of Energy.
The NWTRB meets periodically in open public meetings. The Board reports to Congress and to
the Secretary of Energy at least twice a year on technical issues associated with the Nation's SNF
and HLW disposal program.
ES.2.4 State Governments and Native American Tribes.
In both the NWPA and the NWPAA, the Congress provided for active State and Native
American tribe participation in the Yucca Mountain site evaluation process. The legislation
provides for financial assistance to the State of Nevada, and for affected tribes and units of local
government, for participation in program activities. The State of Nevada and affected tribes or
units of local government may also request assistance to mitigate any economic, social, public
health and safety, and environmental impacts that are likely to result from site characterization
activities at Yucca Mountain.
The Nevada legislature created the State's Nuclear Projects/Nuclear Waste Project Office
(NWPO) in 1985 to oversee Federal high-level nuclear waste activities in the State. Since then,
the NWPO has dealt primarily with the technical and institutional issues associated with DOE's
efforts to characterize the Yucca Mountain site. In addition, the counties contiguous to Nye
County, the host county for Yucca Mountain, have been determined to be affected parties and are
participating in program oversight.
ES-9
-------
ES.3 SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE DISPOSAL PROGRAMS IN
OTHER COUNTRIES
As in the United States, other countries that use nuclear power are establishing long-term
programs for the safe management and disposal of SNF and HLW. These countries include
Belgium, Canada, France, Germany, Japan, Spain, Sweden, Switzerland, and the United
Kingdom. Management strategies of these countries may include SNF storage at and away from
reactor sites, SNF reprocessing, HLW vitrification and storage, partitioning and transmutation of
the waste into short-lived or stable forms, and disposal in deep geologic media.
Deep geologic disposal is considered by the international scientific community to be the most
promising method for disposing of long-lived nuclear waste. As a consequence, all of the
countries discussed in this document envision emplacing solid radioactive waste in a deep
geologic formation located within their national borders.
Only the United States and Germany have identified candidate locations for disposal of HLW,
i.e., the Yucca Mountain site in Nevada and the Gorleben site in Germany. Other countries are to
varying degrees engaged in technical evaluations of the potential suitability of indigenous
geologic formations for disposal. Some nations, such as France, have several geologic
formations such as clay and granite, that might be used for disposal, and each alternative is being
evaluated. Others, such as Canada, have focused on one type of geologic formation. (Canada is
evaluating a crystalline rock formation in a setting with low seismic activity.) In addition,
several countries, such as Canada and Sweden, have established underground research
laboratories (URL's) and extended their research programs to include participation by other
nations with similar candidate geologies.
The disposal strategies of all nations assume that waste isolation will be maintained by reliance
on a combination of engineered and natural barriers between the emplaced waste and the
environment. Currently the United States is, as a result of site characterization data
interpretations, placing increasing emphasis on the role of engineered barriers in a potential
repository site at Yucca Mountain. This is, in part, due to the unique repository environment
and associated disposal strategy. Other countries, because of the characteristics of their available
geologic formations, are also placing emphasis on engineered barrier systems and are designing
these systems to ensure their long-term performance as a barrier to radionuclide release.
Table ES-1 summarizes the characteristics of disposal programs in other nations.
ES-10
-------
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Various nations and international agencies, in addition to the United States, have begun to give
consideration to regulations and regulatory standards for SNF and HLW disposal. Some nations
have developed broad risk or dose criteria, and some have supplemented such criteria with
additional qualitative technical criteria concerning features of the disposal system. International
organizations, such as the Nuclear Energy Agency, provide opportunities for discussion of
regulatory criteria and also provide programs on issues of common interest.
Although the performance standards and the criteria for the various national regulations are
similar, each nation has established specific requirements to meet its needs. Current information
concerning the provisions of national and international criteria and objectives for the safety of
long-lived radioactive waste disposal is presented in Chapter 3 along with a summary of the
waste management programs of Belgium, Canada, Finland, France, Germany, Japan, Spain,
Sweden, Switzerland, and the United Kingdom.
ES.4 WASTE CHARACTERISTICS
Current national plans call for existing and yet-to be-produced inventories of SNF and HLW to
be disposed of in a Yucca Mountain geologic repository if it is approved for disposal. Each of
these waste forms is described below.
ES.4.1 Spent Nuclear Fuel
Spent nuclear fuel is defined as fuel that has been withdrawn from a nuclear reactor following
irradiation and whose constituent elements have not been separated by reprocessing. Generators
of SNF include: commercial nuclear power reactors, which consist of pressurized water reactors
(PWR) and boiling water reactors (BWR); reactors which are used in government-sponsored
research and demonstration programs in universities and industry; experimental reactors, e.g.,
liquid-metal, fast-breeder reactors (LMFBR) and high-temperature gas-cooled reactors (HTGR);
DOE Naval and nuclear-weapons production reactors; and Department of Defense (DOD)
reactors.
Commercial power reactors are by far the largest source of SNF. Approximately 98 percent of
the SNF from these reactors is stored at the reactor sites where it was generated. Spent nuclear
fuel from government research and production reactors is currently stored at various DOE
ES-12
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facilities. The fuels at these DOE facilities are Government-owned and, like commercial fuels,
there are no plans for reprocessing.
The fuel for commercial nuclear reactors consists of uranium dioxide pellets encased in
zirconium alloy (Zircaloy) or stainless steel tubes. During reactor operation, fission of some of
the uranium produces energy, neutrons, and radioactive isotopes known as fission products. The
neutrons cause further fission reactions and thus sustain the nuclear chain reaction. In time, the
uranium, is depleted to such a level that power production becomes inefficient. Once this occurs,
the fuel bundles are deemed "spent" and are removed from the reactor. Reprocessing of
commercial SNF to recover the unfissioned uranium and the by-product plutonium for reuse as a
fuel resource is currently not taking place in the United States.
The radioactive materials associated with SNF fall into three categories: (1) fission products;
(2) actinide elements (atomic numbers of 89 and greater); and (3) activation products. Typically,
fresh SNF contains more than 100 radionuclides as fission products. Fission products are of
particular importance because of the quantities produced, their high radiological decay rates, their
decay-heat production, and their potential biological hazard. Such fission products include:
strontium-90; technetium-99; iodine-129 and -131; cesium isotopes; tin-126; and krypton-85 and
other noble gases.
Activation products include tritium (hydrogen-3), carbon-14, cobalt-60, and other radioactive
isotopes created by neutron activation of reactor components, fuel assembly materials, and
impurities in cooling water or in the fuel pellets. The actinides include uranium isotopes and
transuranic elements, such as plutonium, curium, americium, and neptunium. The exact
radionuclide composition of a particular SNF sample depends on the reactor type, the initial fuel
composition, the length of time the fuel was irradiated (also known as "burnup"), and the elapsed
time since its removal from the reactor core.
As of January 1999, SNF from commercial reactor operations in inventory at various locations
amounted to 37,700 metric tons of initial heavy metal (MTIHM)2. Based on the DOE/EIA Low
Case assumptions of nuclear power capacity through the year 2030, the SNF inventory is
2 Commercial SNF reported in certain DOE documents is in units of metric tons of initial heavy metal
(MTIHM) to avoid difficulties arising from the need to estimate ranges of varied heavy-metal content that result
from different levels of enrichment and reactor fuel burn up. A metric ton (tonne) is 1,000 kilograms, corresponding
to about 2,200 pounds.
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expected to increase to 87,900 MTIHM. This is the amount that would be produced by existing
commercial reactors under current licenses.
ES.4.2 Defense High-Level Radioactive Waste
High-level radioactive wastes are the intensely radioactive materials resulting from the
reprocessing of SNF, including liquid waste produced directly in reprocessing, and any solid
material derived from such liquid waste. High-level waste is generated by the chemical
reprocessing of spent research and production reactor fuel, irradiated targets, and fuel from U.S.
Naval propulsion reactors. The fission products, actinides, and neutron-activated products of
particular importance are the same for HLW as for SNF assemblies.
Historically, weapons program reactors were operated mainly to produce plutonium.
Reprocessing to recover the plutonium was an integral part of the weapons program. Naval
propulsion reactor fuel elements were also reprocessed to recover the highly enriched uranium
that remained after use. DOE decided in 1992 to phase out the domestic reprocessing of
irradiated nuclear fuel of defense program origin, so only minimal amounts of HLW are expected
to be added to the current inventory.
High-level radioactive waste that is generated by the reprocessing of SNF and targets contains
more than 99 percent of the nonvolatile fission products produced in the fuel or targets during
reactor operation. It generally contains about 0.5 percent of the uranium and plutonium
originally present in the fuel. Most of the current HLW inventory, which is the result of DOE
national defense activities, is stored at the Savannah River Site (126,300 m3), the Idaho National
Engineering and Environmental Laboratory (INEEL) (11,000 m3), and the Hanford Site
(239,000 m3). A limited quantity of HLW is stored at the West Valley Demonstration Project
(2,180 m3). The HLW has, to date, been through one or more treatment steps, e.g.,
neutralization, precipitation, decantation, or evaporation. It is currently planned that this HLW
will be solidified, using a vitrification process, for disposal. Vitrification of HLW is in progress
at West Valley and the Savannah River Site. A vitrification facility for HLW at Hanford is being
designed.
The DOE defense HLW at INEEL results from reprocessing nuclear fuels from naval propulsion
reactors and special research and test reactors. The bulk of this waste has been converted to a
stable, granular solid (calcine). At the Savannah River and Hanford Sites, the acidic liquid waste
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from reprocessing defense reactor fuel is or has been made alkaline by the addition of caustic
soda and stored in tanks. During storage, this alkaline waste separates into three phases: liquid,
sludge, and salt cake. The relative proportions of liquid and salt cake depend on how much water
is removed by waste treatment evaporators during waste management operations.
Both alkaline and acidic HLW was generated at West Valley. The alkaline waste was generated
by reprocessing commercial power reactor fuels and some Hanford N-Reactor fuels. Acidic
waste was generated by reprocessing a small amount of commercial fuel containing thorium.
Projecting DOE defense HLW inventories is based on specific assumptions and may be subject
to change. New treatment methods and waste forms are possible and may affect the future
projections. Since all DOE defense production sites are progressing toward closure, there should
be minimal amounts of waste added to the current inventory. Interim storage of DOE HLW will
be required and will most likely continue to be at the site where the waste is produced. Current
DOE policy states that DOE HLW will not be accepted at the geologic repository until six years
after initial receipt of commercial SNF.
ES.4.3 Significant Radionuclides Contained in Spent Nuclear Fuel and High-Level Waste
Of the 70,000-tonne capacity limit for Yucca Mountain set by the NWPA, about 40,785 MTHM
and 22,210 MTHM represent spent PWR and spent BWR fuel, respectively. About 7,000
MTHM of vitrified defense HLW and SNF represents the balance of the total specified repository
inventory. For the Yucca Mountain site, radionuclide-specific activity levels are estimated by
assuming that all spent nuclear fuel had been removed from the reactors 30 years before
emplacement with burn-ups of 39,651 MWd/MTHM for PWR fuel and 31,186 MWd/MTHM for
BWR fuel. Although the burn-up of SNF from which HLW is derived is generally uncertain, this
is thought to affect the adjustment for decay only marginally, hi addition, the radionuclide
inventories in a repository at Yucca Mountain stemming from defense HLW are expected to be
much less than those from commercial SNF.
The radionuclide inventory of the repository will change with time due to radioactive decay and
ingrowth of radioactive decay products. For example, inventories of the initially prominent
fission products cesium-137 and strontium-90, which have approximately 30-year half-lives, will
decay to insignificant levels within 1,000 years, while some decay products, such as neptunium-
237 with a half life of 2.1 million years, will not contribute significantly to doses until about
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50,000 years after repository closure. Activity levels for very long-lived radioisotopes will be
low but nearly constant for periods on the order of a million years. Overall, the radioisotope
inventory of the wastes placed in the repository will decrease by about five orders of magnitude
during the first 100,000 years after closure, and remain virtually constant thereafter.
ES.5 CURRENT INFORMATION ON A POTENTIAL WASTE REPOSITORY AT YUCCA
MOUNTAIN
ES.5.1 Geologic-Features of the Yucca Mountain Site
In terms of geologic designations, the site is situated in the southern section of the Great Basin,
which is characterized by north-south mountain ranges separating narrow, flat valleys. As is
typical of mountains in the region, Yucca Mountain is essentially a tilted fault block, with the
west side steep and nearly vertical and the east side sloping to the adjacent valley floor. The crest
elevation of the mountain is on the order of 1,500 to 2,000 meters (5,000 to 6,000 feet) above sea
level and is about 650 meters (2,000 feet) above the adjacent valley floors.
The geologic features of the southern Great Basin are highly complex and varied, with rock
formations ranging in age from 500 million to less than 400,000 years. The geologic structure at
Yucca Mountain is dominated by a series of layers of rocks that were produced by explosive
volcanic eruptions and are known as tuffs. The tuff layers have widely varying physical
characteristics and are on the order of 10 to 15 million years old. The host rock of the potential
repository is known as Topapah Spring tuff, which is a hard, fractured rock about 13 million
years old.
Geologic features of the region that are important to the integrity of a radioactive waste
repository in Yucca Mountain include faulting, seismicity, volcanism, and stability of the
geologic regime.
ES.5.1.1 Major Fault Features of the Yucca Mountain Area
The geologic formations, of which Yucca Mountain is a part, contain numerous major faults as a
result of deformation caused by tectonic movement. The faults are indicative of past and
potential movement of the geologic structures and they are potential pathways for water to
transport radioactivity released from the repository to the biosphere. The location of faults and
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the extent of recent movement along the faults are important to the location and design of surface
facilities at the disposal site and to the design of the underground facilities into which the wastes
would be placed for disposal.
There are more that 80 known or suspected Quaternary faults and fault rupture combinations
within 100 kilometers (km) of the Yucca Mountain site. The DOE has determined that 38 of
these faults are capable of generating a peak acceleration of one-tenth the acceleration of gravity
(0.1 g) or greater at the ground surface of the proposed repository site; these are classified as
relevant earthquake sources. An updated compilation of faults prepared by the U.S. Geological
Survey identifies 67 faults with demonstrable or questionable evidence of Quaternary movement
and capability for accelerations of at least 0.1 g at an 84 percent confidence limit. The NRC-
supported program of the Center for Nuclear Waste Regulatory Analyses (CNWRA) has
identified 52 Type I faults within a 100-km radius of the mountain. Eleven known or suspected
Quaternary faults exist within 20 km (12 miles) of Yucca Mountain.
The three major faults in the immediate vicinity of the proposed disposal site are the Ghost
Dance fault, which passes through Yucca Mountain; the Bow Ridge fault, which is just to the
east of the mountain; and the Solitario Canyon fault, which is just to the west of the mountain.
According to DOE's interpretation of available data, no movement on any of these faults has
occurred during the past 10,000 years.
ES.5.1.2
Seismology of the Yucca Mountain Area
The fault systems and the seismic history of the Yucca Mountain area are the result of regional
tectonics, which are dominated by the interaction of the North American and Pacific plates. The
tectonic processes that are stretching the Great Basin and produced its major land forms are the
result of the Pacific Plate moving northwest relative to the North American plate; the typical
geologic structures of the region were developed on the order of 11 million years ago. The
relative plate movements produced the past and recent seismic activity characteristics of the
region, outlined below.
Seismicity in the region of Yucca Mountain is concentrated in several zones. The Southern
Nevada Transverse Zone (SNTZ) is nearest to the Yucca Mountain site and is the most
significant to repository performance. Historic earthquakes in the SNTZ have been of moderate
magnitude with no documented surface rupture. The most recent earthquake in the vicinity of
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Yucca Mountain was the Little Skull Mountain event, of Richter magnitude 5.6, in June 1992.
This earthquake was centered 20 km southeast of Yucca Mountain and was associated with the
Landers, California earthquake earlier that year. It caused minor structural damage to the Yucca
Mountain project field office near the mountain but had no apparent effect on geologic features
near the mountain. It was the largest earthquake ever recorded in the vicinity of the site, based
on nearly 100 years of records.
Assessments of available data indicate that Yucca Mountain has not been subject to ground
accelerations at the surface in excess of 0.2 g for over several tens of thousands of years. At the
proposed waste emplacement depth of about 300 meters, the effects of ground motion are
expected to be insignificant. Empirical evidence of damage to 71 rock tunnels in Alaska,
California, and Japan resulting from earthquake shaking indicates that tunnel damage does not
occur at peak surface accelerations of less than approximately 0.2 g and only minor tunnel
damage occurs when the peak surface acceleration is between 0.2 and 0.5 g. Since ground
acceleration is not expected to impose significant design demands on either the underground
repository or surface facilities at Yucca Mountain, DOE does not consider seismicity to be a
significant factor in repository performance.
The Department also believes that future tectonic events are unlikely to significantly alter the
hydrologic characteristics of the Yucca Mountain site. This position .is based on the assumption
that the current state of faults and fractures at the site is the result of cumulative past tectonic
events. The CNWRA has proposed, however, that a single tectonic event can cause significant
changes in hydrologic characteristics. Currently, there are five alternative tectonic models which
may form the basis for future assessment of relationships between tectonic phenomena and the
hydrologic regime.
ES.5.1.3
Volcanism
Yucca Mountain is composed of layers of volcanic rocks which originated in silica-rich eruptions
at what is now the Timber Mountain volcanic basin complex starting about 10 km north of
Yucca Mountain. The principal eruptions took place approximately 11 to 15 million years ago,
and ceased about 7.5 million years ago. After the silicic volcanism ended, there were two
episodes of basaltic volcanic rock formation. The most recent of these, which produced minor
ash deposits in the Lathrop Wells area to the southwest of Yucca Mountain, ended about 9,000
years ago.
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DOE and NRC agree that a future occurrence of silicic volcanism is highly unlikely; the
consequences of such an event, therefore, do not need to be considered in assessments of a waste
repository at Yucca Mountain. The Department and the NRC have also recently reached
agreement on the likelihood of future basaltic volcanic events and their possible consequences.
One of the phenomena potentially associated with basaltic volcanism is sheet-like intrusion of
molten or liquid basaltic rock along fractures in the overlying rocks, i.e., the formation of dikes.
Given the history of volcanism in the Yucca Mountain region, there is some potential for magma
from a basaltic volcanic event to either intersect the repository footprint and directly affect the
waste or to form a nearby intrusive dike that might affect the waste isolation capability of the
natural system. If such intrusions occur, they could mobilize wastes and/or alter ground water
pathways. DOE and NRC are currently developing a mutually acceptable approach to estimating
the likelihood and consequences of such intrusions.
ES.5.1.4 Geologic Stability
The NAS report, "Technical Bases for Yucca Mountain Standards," recommended that
assessments of compliance with the Yucca Mountain standards be conducted for the time at
which the greatest risk occurs, within the limits imposed by long-term stability of the geologic
environment. The report also stated that long-term geologic stability for time periods on the
order of one million years can be expected; i.e., the contribution of geologic and hydrologic
features to overall repository system performance can be assessed for time periods of this
duration. The NAS report concluded that there is no technical basis for selecting a shorter
compliance period, such as 10,000 years. However, the NAS also stated that EPA may select a
shorter period based on policy considerations.
The concept of geologic stability does not imply absence of geologic activity or absence of
change in geologic processes. Rather, the concept implies that processes and events such as
climate change, tectonic movement, and earthquakes will occur as in the past, and that variations
within these processes and events will be boundable. The NAS report does not explicitly justify
the assertion of million-year stability by providing a synopsis and interpretation of the
documented geologic record. Some of the references cited in the report contain information
about the geologic record, but none of the cited references interprets the record to indicate a
million-year stability of the geologic regime or the processes associated with it. Until recently,
DOE documents containing information about the geologic features of the Yucca Mountain site
anticipated that performance assessments for a disposal system at the site would be evaluated in
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terms of EPA's 40 CFR Part 191 regulations, which require evaluation of performance for a
period of 10,000 years. The 40 CFR Part 197 regulations specify the same time period.
The 10,000-year time frame for compliance with EPA's 40 CFR Part 191 regulations, which
applied to Yucca Mountain until Yucca Mountain was exempted by the WIPP Land Withdrawal
Act of 1992, was selected by the Agency because it was relatively brief compared to the time
frame for long-term factors, such as tectonic motion, that might affect the natural environment
and are not reasonably predictable over that period. On the other hand, the time period was long
enough to bring into consideration factors such as degradation of engineered barriers and
earthquakes that might affect disposal system performance and allow radionuclides to reach the
accessible environment.
Available information generally supports the NAS assertion that the fundamental geologic
regime at Yucca Mountain will remain stable over the next one million years. The overall
picture that emerges from available information is that the site and region had a highly dynamic
period of volcanism, seismicity, and tectonic action during the past, but that this very dynamic
' situation has matured into one where the magnitudes, frequencies, locations, and consequences of
such phenomena relevant to long-term future disposal system performance can be bounded and
projected with reasonable confidence.
Performance assessments define the expected behavior of the waste isolation system over time.
Within the framework of expected repository performance, it is convenient to characterize future
repository conditions over three time periods. A similar breakdown was presented by DOE in its
1998 Viability Assessment. In the first, short-term period, lasting about 100 to 1,000 years, the
repository is characterized by intact waste canisters, high temperatures, and temperature gradients
which serve as driving forces for transients such as chemical reactions, and the retention of short-
lived and long-lived radioactivity in the canisters. Percolation water may or may not contact the
canisters, depending on local conditions determined by the arrangement of waste packages in the
repository and the pattern of percolation into the repository.
In the intermediate period, with a duration between 1,000 and 10,000 years, temperature
gradients are diminished or gone and the engineered features of the repository start to degrade.
During this time, canisters begin to corrode and only long-lived radioactivity remains; some of
the radioactivity is released from a few canisters which are penetrated by water, but most is
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retained within the repository. Percolation water contacts and transports radioactive waste.
Releases are dominated by technetium-99 and iodine-129.
' In the long-term period, from 10,000 years to 1,000,000 years, the repository gradually evolves
into an assemblage of the oxides, hydroxides, or carbonates of waste-package and waste-form
materials at ambient conditions. Percolation water seeps through the repository level and
transports radionuclides that can be mobilized to the environment, where the radionuclide
concentrations are diluted and dispersed by ground water flow processes. Potential for radiation
doses is dominated by neptunium-237 released from the repository.
ES.5.2 Hvdrologic Features of the Yucca Mountain Site
The proposed repository depth in Yucca Mountain (about 300 meters or 1,000 feet) would locate
it in a geologic formation not fully saturated with water (the unsaturated zone). The unsaturated
zone depth to the water table beneath the repository horizon is variable but is on the order of 300
meters. Water that infiltrates into the mountain, percolates through the repository, and moves
through the matrix of the geologic formations in the unsaturated zone will travel slowly, thereby
delaying entry of radionuclides released from the repository into the saturated zone and ground
water system. Fractures in the rocks within the unsaturated zone can act as conduits for relatively
rapid movement of ground water through Yucca Mountain. Some radionuclides may be
chemically trapped in rock formations in the unsaturated zone.
ES.5.2.1
Characteristics of the Unsaturated Zone
Water flows slowly through the pore space in the matrix of partially saturated rock (the degree of
saturation in the Yucca Mountain formations is on the order of 80 to 90 percent) because there is
little areal recharge. If the pore space becomes saturated, the water will flow more quickly under
the existing hydrologic conditions. Also, water may flow quickly and preferentially through
fractures in the rock matrix. There is experimental evidence of "fast paths" for flow in some
rock fractures at Yucca Mountain. The fraction of total flow through these fast paths is
uncertain, may be episodic, and may be a small percentage of the total ground water moving
through the repository host rock. There is also evidence that some faults and fractures are
barriers to flow because of solids deposited along the fractures which block potential flow paths.
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The complexity of the geologic structures in the unsaturated zone and the complexity of flow in
partially saturated media make it difficult to develop accurate models to predict flow rates and
How paths in the unsaturated zone below the proposed repository location. Water flow and
storage in the unsaturated zone is three-dimensional and is controlled by structural, stratigraphic,
thermal, and climatological features of the system. The presence of features such as fractured
porous media, layered geologic units with widely varying hydrologic properties, tilted rock units,
and bounding faults can be expected to result in phenomena such as flow in both the fractures
and the matrix, diversion of flow by capillary barriers, lateral flow along discontinuities, perched
ground water zones, and vapor movement.
Water quantities that enter the mountain from precipitation and that percolate through the
geologic structures are spatially and temporally variable. The amount of water that percolates
along different paths is highly variable. Infiltration pathways depend on variations in the
properties of geologic units, the intersections of faults with the surface, and the presence of local
fracturing in individual rock units. Variations in the time it takes water to infiltrate are related to
the seasonally and relative infrequency of precipitation at Yucca Mountain. Over long time
frames, variations will occur because of climate changes. The interplay of all of these factors
may act to even out downward movement of ground water in the unsaturated zone with
increasing depth from the surface. There is evidence of rapid movement of infiltrating waters
along fracture zones in the rock.
Quantities of water that percolate through the mountain at the proposed repository depth cannot
be measured directly. Recent estimates, based on analysis of site characterization data, place the
percolation rate in the range of one to 10 mm/yr. Base-case performance assessments for the
TSPA-VA used a range of three to 23 mm/yr, with an expected value (60 percent probable) of
7.7 mm/yr. Values in this range are as much as two orders of magnitude higher than values
previously estimated using more limited data. The TSPA-VA also used a model of future
climate involving "long-term average" conditions, with an expected infiltration rate of 42 mm/yr,
and "superpluvial" conditions with an expected infiltration rate of 110 mm/yr.
Models of water flow in the unsaturated zone take into account potential for flow in both the
matrix and fractures, the relative distribution depending on the quantities of ground water
available. For example, at high percolation rates, a larger fraction maybe transported laterally
and/or transported in fractures, including fast-path fractures. The models also take into
consideration the possibility that radionuclides may be removed from water that is intercepted by
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geologic media having a high capacity to chemically absorb and retain some radionuclides, such
as the zeolite materials in the Calico Hills formation.
Using uncertainty distributions for the flow parameters, models are used to estimate values for
performance factors, such as the time necessary for water to move through the unsaturated zone.
Results of studies to date show that radionuclides, carried by water through fractures, cross the
unsaturated zone much more rapidly than those in water that travels through the rock matrix.
Similarly, radionuclides strongly sorbed on rocks, such as the Calico Hills zeolites, have transit
times through the unsaturated zone 50,000 times longer than for radionuclides that are soluble
and travel with the water. The conceptual models and transport parameters for water flow and
radionuclide transport in the unsaturated zone, and the results obtained from use of the models,
will be refined by DOE as additional data concerning the unsaturated zone are obtained from
future site characterization work.
ES.5.2.2 Characteristics of the Saturated Zone
Water that percolates through the repository and the unsaturated zone below .will enter the
saturated zone where ground water fills the pore spaces and fractures within these rocks. The
saturated zone at Yucca Mountain is located at depths on the order of 300 m below the repository
horizon. Radionuclides transported to this zone will move toward the environment away from
Yucca Mountain through ground water. Radionuclide concentrations in the saturated zone will
be reduced by dilution caused by dispersion as radionuclides are transported away from the
repository at rates and in directions according to the flow characteristics of the hydrologic
regime. The saturated zone is, like the unsaturated zone, composed of numerous layers of rocks
with widely different characteristics and complex structures resulting from the dynamic geologic
history of the region . Flow rates and directions are of interest for evaluating compliance with
EPA's standards, as are the locations at which radionuclides would be accessible to human use
and the radionuclide concentrations at those locations.
The sequence of volcanic rocks within and below Yucca Mountain has been described
hydrologically in terms of four hydrologic units characterized by their ability to transmit water.
Beneath the volcanic rocks, at depths on the order of 2,000 meters at some locations, are older
rocks which contain the Lower Carbonate Aquifer. The volcanic hydrologic units and the lower
carbonate aquifer may all have a role in transporting radionuclides from the repository to the
surrounding areas.
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The thicknesses of the rock formations and the depth to water in the saturated zone vary
significantly with distance and direction from the proposed repository location. For example, the
volcanic rock units are believed to thin out and disappear to the south of Yucca Mountain where
' they are covered by the alluvial deposits of the Amargosa Desert. In this region, as illustrated by
Figure ES.5-1, the formations containing the Lower Carbonate Aquifer are near the surface.
Depths to ground water currently used for human consumption and activities such as irrigation
are shallow in this area, i.e., on the order of a few tens of meters. Consequently, human
habitation and water supply wells are currently located in this area.
Available data indicate that much of the outflow from the volcanic aquifers moves laterally into
the alluvial aquifer as the volcanic rock formations thin out below it. The alluvial aquifer may
also be receiving water from the carbonate aquifer. The data are not sufficient to indicate where
and how these flow transitions occur. Comparison of recent water-level altitude maps with those
completed in the 1950s indicates that aquifer development may have had a significant impact on
water levels and flow directions. Pumping of the alluvial aquifer may have induced upward flow
from the lower carbonate aquifer into the alluvial system.
Discharge from the alluvial aquifer system can occur by interbasin flow, leakage to underlying
units evaporation, and extraction for human use. Available data indicate that the major
discharge area for the alluvial aquifer system is Alkali Flat, known also as the Franklin Lake
Playa The estimated discharge rate in this area is 10,000 acre-feet per year, primarily based on
bare-soil evaporation. Some of the alluvial aquifer flow may also move further to the southwest
and discharge in the Death Valley region, but the extent of this is unknown.
Estimates of rates and quantities of ground water flow in the saturated zone are based on
estimates of values for hydraulic conductivity, hydraulic gradient, and effective porosity of the
formations through which the water is flowing. Hydraulic gradient (i.e., the change in water
level between two locations) is generally the parameter best known and most easily measured.
In the Yucca Mountain region, three regions with distinct hydraulic gradients, designated as
small, moderate, and large, have been identified. Their extent and characteristics are governed
by the complexity and characteristics of the geologic formations.
Of particular interest to repository performance at Yucca Mountain is the high-gradient area,
located about two miles north of the mountain. The cause of the gradient, in which water levels
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decline by more than 900 feet over a distance of about a mile, is unknown. Possible causes
include a flow barrier, a fault, an intrusive dike from volcanic rock flow, or changes in the
detailed structure of the rocks. If the gradient is caused by a flow barrier of some type, a loss of
this barrier due to future geologic movements could cause a rise in the water table in the area of
the proposed repository. A rise in the water table would not be expected to intercept the
repository, but it would decrease the thickness of the unsaturated zone and decrease the
radionuclide travel time from the repository to the environment.
Ground water flow rates, like flow directions and quantities, are at present highly uncertain
because of limited data and the complexity of the geologic structures that create the hydrologic
regime. Flow rates in the alluvial aquifer, the volcanic rock aquifers, and the lower carbonate
aquifer will differ because of the different rock characteristics for these geologic regimes.
Ground water movement in the volcanic rocks of the saturated zone was estimated by DOE in
1993 to be in the range of 5.5 to 12.5 meters per year. A more recent estimate concluded that a
flow rate of five meters per year is in the middle of the range of reasonable estimates. However,
recent data from the Exploratory Studies Facility tunnel into Yucca Mountain suggest the
existence of "fast paths" through the unsaturated zone that can allow water to move from the
surface to depths as far as 300 m in 50 years. At present, DOE believes that only a small fraction
of percolating water is transported to the repository level through these pathways. Flow rates in
the lower carbonate aquifer have been estimated to be in the range of three to 3,000 meters per
year, depending on location. Pressure gradients are such that water flow from the volcanic
aquifers to the lower lying carbonate aquifer presumably does not occur. While reliable
estimates of flow rates in the alluvial aquifers are not available, flow rates in these strata are
believed to be lower than in the carbonate strata. This has the effect of preventing radionuclide
from moving into the higher flow rate paths in the carbonate aquifer. The areal extent of the
region where upward flow comes from the carbonate aquifer is highly uncertain.
If ground water containing radionuclides flows at a rate of five meters per year, it would take
1,000 years for the ground water to travel a distance of five kilometers and 4,000 years to travel a
distance of 20 kilometers. Concentrations of soluble radionuclides in the ground water at these
distances from the repository would depend on the initial concentration at the boundary of the
repository, the dilution that occurs as a result of mixing of water from various sources, and the
dispersion of radionuclides. Overall, the mixing, dilution, and dispersion processes have been
estimated potentially to reduce radionuclide concentrations at distances on the order of
20 kilometers from the repository by a factor of 10 in comparison with the initial concentration.
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Estimation of dispersion on a kilometer scale is difficult. The DOE used expert elicitation to fix
this parameter. The experts estimated that the parameter could vary from 1 to 100 with an
average assumed value of 10. The amount of concentration reduction that occurs may depend on
the direction of flow (i.e., as a result of dispersion being controlled by rock structures along the
flow path) as well as the distance over which flow has occurred.
Nye County, NV is currently drilling a series of 20 deep and shallow wells south of the Yucca
Mountain site and in Amargosa Valley to monitor the behavior of the saturated zone. Some of
the wells will measure hydraulic parameters of the alluvial and tuff aquifers. Other deep
monitoring wells will be installed to measure the properties of the carbonate aquifer and to define
how this aquifer connects with the shallower tuff and alluvial aquifers. These data will support
modeling of the saturated zone flow and transport on both site-scale and regional-scale. Results
to date indicate that the alluvium is complex and layered.
ES.5.3 Climate of the Yucca Mountain Region
The region surrounding Yucca Mountain currently has an arid climate, with total annual
precipitation on the order of 170 mm (six inches) of water. Precipitation rates vary throughout
the year, averaging about 18 mm/month during the fall and winter months and nine mm/month
during the spring and summer months. Current climate conditions have apparently prevailed
during the past 10,000 years, i.e., since the last ice age. Prior to the ice age, the climate cycled
between wet and dry; during the wet periods, many of the valleys that are now dry contained
lakes.
Future variations of precipitation and temperature are climate factors of considerable interest for
predicting the performance of a repository at Yucca Mountain. These factors influence the
percolation rate of water through the repository and the transport of radionuclides released from
the repository to the environment.
Current arid conditions are expected to persist well into the future. These conditions are
associated with the rain shadow caused by the Sierra Nevada Mountains to the west, which are
still rising. In addition, increases in greenhouse gases and global warming may affect general
atmospheric circulation and local climate conditions at Yucca Mountain. A panel of experts,
convened by the CNWRA, estimated that an enhanced greenhouse effect would probably
produce warmer conditions than have been experienced during the past few thousand years, with
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a likely increase in the upper limit of temperature in the Yucca Mountain region on the order of
two to three degrees Celsius. The time period during which these elevated temperatures would
persist would depend on assumptions about future human use of fossil fuels. In general,
increased temperatures would be accompanied by lower precipitation rates and, therefore, lower
rates of percolation through the repository. Opposite changes could occur, however, especially in
connection with any future glacial periods.
Performance assessments in the Department's TSPA-VA assumed that the climate alternates
between the present (dry) climate, a long-term average climate during which the precipitation
rate is twice the current rate, and a superpluvial climate during which the precipitation rate is
three times the current rate. The expected duration of the initial dry climate was 5,000 years.
Subsequent dry periods would have an average duration of 10,000 years. The expected duration
of the long-term average periods was 90,000 years. Two superpluvial periods of 10,000 years
each were assumed to occur over the 1,000,000-year model period. About 90 percent of the
1,000,000-year model period was characterized as having long-term average climate. Climatic
fluctuations were predicted to have virtually no impact on repository performance assessments
over a 10,000-year time frame. Over the longer term, climate assumptions affect the time at
which the peak dose rate occurs but not its magnitude.
ES.5.4 Repository Design Concepts Under Consideration for Yucca Mountain
Design concepts for a potential waste repository at Yucca Mountain have evolved significantly in
response to information from sources such as site characterization data, repository system
performance assessments, external technical reviews, and refinement of a waste isolation
strategy. The original design concept envisioned vertical emplacement of simple steel canisters
in individual boreholes; current plans call for end-to-end horizontal emplacement of large,
complex waste packages in parallel, excavated drifts. Design details are expected to continue to
evolve until a final design is selected for the License Application, if the site is approved for
disposal.
The repository design can be characterized as a multi-barrier system that functions to delay the
failure of the waste package, delay the release of radionuclides from the waste package, and
mitigate the effects of radionuclide release. Key design factors important to repository
performance and radionuclide release potential include the corrosion resistance of the waste
package wall material; the use of techniques to deflect or delay contact of percolating water with
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the waste packages; and the use of techniques to stop or delay migration of releases to the
environment.
One technique to delay waste package failure is to emplace the packages so that heat from the
wastes will keep temperatures in the repository high enough to vaporize percolation water, for as
long as possible. Corrosion by liquid water is thereby delayed until the heat emissions decrease
to levels such that water can enter the repository. Another technique for delaying waste package
failure is to use shields that deflect water dripping into the emplacement tunnels from contact
with the packages. In general, various technologies and concepts are available for each of the
basic functions for delaying waste package failure and decreasing radionuclide releases.
Key features of the design used by DOE in the recently completed TSPA for Site
Recommendation issued in December 2000 include the following:
Horizontal emplacement of 7,642 commercial SNF and 2,858 HLW canisters
positioned end-to-end in parallel, excavated, concrete-lined drifts, with an initial
thermal loading (to vaporize percolation water) on the surroundings corresponding
to emplacement of 85 MTHM/acre of reference spent commercial reactor fuel
Emplacement of waste canisters only between the Ghost Dance fault and the
Solitario Canyon fault
Disposal of 63,000 MTHM of spent commercial fuel and 7,000 MTHM
equivalent of defense HLW in 120 miles of tunnels and drifts, over 840 acres of
emplacement area, at depths on the order of one-eighth to one-quarter of a mile
Construction of 29 surface buildings encompassing 800,000 square feet of floor
space and serving the operational needs of 300 underground drift excavation
personnel and 600 surface and subsurface operational personnel
• Use of commercial spent fuel waste packages that are two meters in diameter, six
meters in length, with an Alloy 22 corrosion-resistant waste container.
.• Protection of the waste containers using titanium drip shields, which decrease
contact with water and extend the lifetime of the Alloy 22 containers.
Waste types to be disposed would include commercial SNF fuel in bare assemblies; canistered
commercial SNF; canisters of vitrified defense HLW; SNF from nuclear defense programs; and
other DOE-owned SNF such as unprocessed fuel from Hanford's production reactors.
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A plan view of the repository layout for the TSPA-VA is shown in Figure ES.5-2. In this
diagram, the dense array of parallel lines in the subsurface emplacement block represents the
drifts where the waste canisters would be emplaced. The waste package design features shown in
Figure ES.5-3 are representative of those used in the VA reference design.
In the TSPA-VA, DOE evaluated several design options as variants to the base case. The options
considered included use of backfill; use of drip shields to preclude water from impinging on the
waste packages; use of a ceramic coating on the waste packages to defer corrosion; and whether
or not to take credit for fuel rod cladding as a barrier to radionuclide mobilization and transport.
During early 1999, DOE evaluated alternative repository designs based on, and evolved from, the
Viability Assessment design. These alternative designs were intended to reduce uncertainties in
performance identified in the TSPA-VA analysis. DOE has selected, as the reference design for
the Site Recommendation, a design whose key features include an area mass loading of 60
MTU/acre, drift spacings of 81 m, waste packages with 2 cm of Alloy 22 over 5 cm of 316L
stainless steel, use of steel ground support, and use of drip shields. These design features
significantly reduced uncertainties and technical issues associated with the VA reference design.
The demonstration of this improved performance was conducted as the TSPA-SR. Engineered
design concepts may continue to evolve to the design selected for the LA.
ES.5.5 Repository System Performance Assessments
Assessments of future repository system performance are currently used by DOE to aid
repository design and, in association with the license application, will be used to demonstrate
compliance with regulatory standards. The assessments are done using analytical models of
factors that affect performance, such as the waste package corrosion rate, and computer codes
that combine the models of performance factors with performance parameter values and
modeling assumptions.
The NRC is developing independent capability to perform performance assessments in order to
be able to review DOE's license application. In addition, the Electric Power Research Institute
has developed performance assessment methods that are used in its oversight of the government
program, and EPA has developed methods that are used in support of its promulgation of the
Yucca Mountain standards.
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Figure ES.5-2. Repository Layout for the TSPA-VA Design
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Prior to the TSPA-VA, DOE issued reports on its total system performance assessments in 1991,
1993, and 1995. They also serve as precursors of the TSPA results that will be submitted with
the License Application to demonstrate compliance with regulatory requirements. The expected
performance of the repository will be strongly dependent upon the combination of engineered
barriers finally selected by DOE.
The performance assessment models and codes address engineered design features that affect
repository system performance. They also consider geologic and hydrologic features that can
affect performance, such as those discussed in Sections 4.1 and 4.2 of this BID, and uncertainties
in these factors that affect uncertainties in demonstrating compliance with regulatory
requirements.' Types of uncertainties that are considered include uncertainties in measured
values of performance parameters such as corrosion rates and hydrologic parameters; spatial
variability of parameters such as percolation rate and temperatures around the repository;
temporal variability of factors such as annual variation of precipitation and future climate
change; and uncertainties in the analytical models as a result of simplifications or imperfect
knowledge of the processes simulated by the models.
These uncertainties are taken into consideration by the codes used in TSPAs through use of
probabilistic techniques in selecting the model parameters for the calculations. The numerical
values for uncertain performance parameters used in the TSPA codes are characterized using a
range of possible values. Calculational techniques are used to sample values from the
distributions to produce a large number of individual TSPA results which collectively
characterize the uncertainty in the overall system performance as a result of uncertainties in the
individual factors. DOE uses the computer code GOLDSIM as the integrating shell to link the
various component codes. GOLDSIM includes parameter sampling capability to assess
uncertainty in dose rates as a function of uncertainty in the component models, and permits
flexible implementation of alternative conceptual models are part of its framework.
During the first 10,000 years, based on the TSPA-SR nominal scenario assumptions, there are no
releases from the repository. However, the TSPA-SR also accounts for the potential for
disruption of the repository by igneous activity. Two scenarios are evaluated in the TSPA-SR.
In the first igneous scenario, magma intrudes into the repository, completely destroying waste
packages it encounters, carrying waste to the surface, where a violent eruption occurs. Waste
materials, in this scenario, are then distributed along with the ash plume created by the eruption.
In the second igneous scenario, magma intrudes into the repository, destroying waste packages it
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encounters, but does not progress to the surface. In this scenario, the damage to the repository
permits water to contact the waste, leading to early releases from the repository. These two
igneous scenarios are the only mechanisms in the TSPA-SR leading to releases in the first 10,000
years. The mean dose rate from these scenarios reached a maximum of 0.1 mrem/yr in the first
10,000 years.
Inadvertent drilling intrusion as a result of searching for water was assumed to occur at 100 or
10,000 years. The former value is based on the proposed NRC guidance for 10 CFR 63, and the
latter reflects EPA's proposed position in the draft 40 CFR 197 that the waste package must
degrade for a period of time before it would be unrecognizable to a driller. The mean peak dose
calculated in the TSPA-SR for either intrusion time is 0.01 mrem/yr; there is virtually no
difference between the mean peak dose for the two assumptions.
The NRC is developing its own performance assessment models and codes for use in pre-
licensing technical exchanges with DOE, and, ultimately, for performing reviews of DOE's
license application for a repository at Yucca Mountain. The NRC models are similar in concept
and content to the DOE models in that they include models of the various factors relevant to
disposal system performance and have the capability to address uncertainties. NRC's recent
modeling has shown results similar to those developed by DOE in the TSPA-VA, even though
significant differences in underlying assumptions exist between the two approaches. NRC has
not yet updated their TSPA to reflect comparable conditions to the TSPA-SR.
The NRC has estimated that the 10,000-year dose rate is about 0.003 mrem/yr as compared to the
equivalent TSPA-VA dose rate of 0.04 mrem/yr. At 100,000 years the NRC calculated dose rate
is 0.2 mrem/yr while the equivalent value in the TSPA-VA is 5 mrem/yr. Reasons for the
differences are not readily apparent because the parameters and modeling approaches used by the
two agencies differed markedly. For example, the NRC did not assume credit for fuel rod
cladding as an engineered barrier while in the TSPA-VA DOE did take credit for Zircaloy
cladding. The NRC assumed that dilution of ground water radionuclide concentrations occurs
during pumping by the dose receptor; the DOE assumed that this dilution did not occur.
The Electric Power Research Institute has also conducted performance assessments using models
and codes which address the same general features and processes as those modeled by DOE.
However, the codes differ in approach and detail from those used in the TSPA-VA. Differing
parameters are also used by the two organizations. In spite of these differences, 10,000-year
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performance assessment results were in reasonable agreement. The 10,000-year dose calculated
by EPRI was 0.08 mrem/yr as compared to the DOE TSPA-VA estimate of 0.04 mrem/yr. After
100,000 years the EPRI dose rates were about two orders of magnitude lower than the TSPA-VA
results due, at least in part, to differing assumptions about the available inventories of iodine -
129 and technicium-99.
ES.6 BIOSPHERE PATHWAYS LEADING TO RADIATION EXPOSURE
In order to evaluate compliance of the repository system performance with regulatory
requirements, potential radiation doses to humans from repository releases must be calculated.
This evaluation requires estimating radionuclide releases; modeling their movement through the
environment; selecting and characterizing the person(s) for whom the potential radiation dose is
to be evaluated; and characterizing the pathways by which the person(s) receives the dose.
This estimation also requires assumptions concerning the location and exposure scenarios of an
individual or group of individuals likely to be at greatest risk from potential radionuclide releases
after repository closure and removal of institutional controls. Prior to closure, such assumptions
are unnecessary because possible contamination levels can be measured with considerable
accuracy both within and outside the repository.
Releases of radionuclides from the repository are not expected to occur sooner than several
thousands of years in the future; the start of release might be deferred much longer if certain
repository design features are used (i.e., those aimed at delaying the start of release, such as
corrosion-resistant drip shields and waste packages). After release from the repository, the
radioactivity would migrate through environmental pathways until it reaches the location of the
person(s) selected for the evaluation of potential doses. Thus, radiation doses might first be
incurred many thousands of years in the future, when locations and lifestyles of humans in the
vicinity of Yucca Mountain might differ from those of the present. Human locations and
lifestyles far in the future cannot, however, be reliably estimated. Therefore, evaluations of
future potential radiation doses are based on an understanding of current patterns of human
habitation, physiology, and activities as well as current technology. This approach to addressing
future states was affirmed by the NAS who concluded in their study mandated by the EnPA that:
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...based on our review of the literature we believe that no scientific basis
exists to make projections of the nature of future human societies to within
reasonable limits of uncertainty
• ... 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.
ES.6.1 Current Demographics and Land and Water Use
The boundaries of the unincorporated town of Amargosa Valley (the closest population center to
the repository site) encompass almost 500 square miles of the Amargosa Desert. The boundaries
of the town include all of the area in which the highest potential doses from a repository at Yucca
Mountain are anticipated. The remoteness and arid climate of the area are reflected by its
population of only about 1,000 residents. Only about 11 percent of the land is held privately; the
remainder is under Federal control.
Currently, agricultural activities in the Yucca Mountain region are restricted to the Ash Meadows
area and the southern portion of Amargosa Valley. Two commercial alfalfa farms, a dairy farm,
and one commercial sod farm operate full-time in the Valley; most other farms in the area
operate on a part-time basis. Despite some difficulties, a wide range of crops and livestock can
be raised. Alfalfa, hay and grass, wheat, fruits and melons, vegetable, cotton, nuts, poultry, beef
cattle, dairy cattle, and fish are being or have been grown on farms and ranches in Amargosa
Valley. However, because of local conditions, the population in the region does not currently
grow significant quantities of leafy vegetables, root vegetables, and fruit and grain crops for its
own use. Presently, no farming occurs closer than about 23 kilometers south of the repository
site.
Primary uses of water in the Amargosa Valley include domestic, industrial, agricultural, mining,
and recreational. Most residences are supplied by individual wells, though some trailer parks,
public facilities, and commercial establishments are served by small private water companies. A
number of springs also supply water, primarily to the resort area in Death Valley.
Water use data for Hydrographic Basin 230 (Amargosa Desert) in 1997 was 940 acre-feet for
domestic, quasi-municipal, and commercial uses exclusive of mining and irrigation. As such, the
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usage is typical of a small rural residential community. The average per person use rate was 0.8
acre-feet per year. Since no major demographic changes are expected, these values should be
representative of future communities in the regions.
At the present time nine farms varying in size from 65 to 800 acres are cultivating alfalfa in the
area. It is estimated that a total of 2,500 acres is being cultivated in 1999 and that water usage for
alfalfa irrigation is, as limited by current allocations, 5 acre-feet per acre. The nine alfalfa-
growing operations have an average size estimated to be 255 acres. This results in an average
annual water use for irrigation of 1,275 acre-feet per year. The domestic use of water by a small
farming community of 25 people is estimated to be 10 acre-feet per year, so the average volume
of water that would supply the annual water needs of a hypothetical future agricultural small
community would be 1,285 acre-feet.
ES.6.2 Radiation Protection of Individuals
According to current understanding, contaminated ground water is the principal pathway by
which a release of radionuclides from a repository at Yucca Mountain could cause radiation
exposures to humans. Figure ES.6-1 illustrates the ground water pathway leading to human
exposure from an undisturbed repository at Yucca Mountain. The major reservoirs (source
terms) containing radionuclides at various times following closure are depicted as rectangles.
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Critical group for
population dosa
Critical group?
Figure ES.6-1.
Schematic Illustration of the Major Pathways from a Repository at Yucca
Mountain to Humans (copied from the NAS report, 1995)
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Solid arrows between reservoirs represent the probable processes by which radionuclides are
transported from one reservoir to another in an undisturbed repository. Major processes and
events with the potential to modify normal behavior or drastically alter the physical integrity of
reservoirs are shown in the figure as diamonds. These modifiers are connected by dashed lines to
those reservoirs most likely to have the most significant impact.
Individuals in a human population may have greatly different responses to radiation exposure
reflecting differences in factors such as age, life style, and family history. In addition, their
potential exposure to radiation released from a repository at Yucca Mountain will depend on
factors such as where they live and what they eat and drink. A wide range of radiation exposures
and effects is therefore possible. Because of these variations, some specification of the exposure
conditions to be considered in measuring compliance must also be part of the regulations.
Specifying some variables in the compliance evaluations would provide a means to narrow and
characterize the range of conditions for which evaluations of compliance are to be made. More
than one approach is possible for assessing potential radiation doses to individuals down the
hydrologic gradient from the repository.
ES.6.3 Dose Estimation Approaches
To determine the risk to exposed individuals resulting from contaminated ground water requires
the development of a comprehensive exposure scenario that specifies discrete pathways and
quantifies the intake of individual radionuclides. Pathways for human exposure from
contaminated well water include internal exposure from the ingestion of drinking water,
vegetables, fruits, dairy products, and meats. For persons engaged in agricultural activities,
internal exposure may also result from the inhalation of airborne contaminants resuspended from
soil irrigated with contaminated water. Over time, the buildup of soil contaminants could reach
levels that also give significant external doses.
The implementation of an exposure scenario appropriate for a specified population requires a
complex array of pathway parameter values that define potential radionuclide concentrations in
various media to which individuals may be exposed. Exposure scenarios must also provide
quantitative descriptions that include where individuals live, what they eat and drink, and what
their sources of food and water are. Many key parameters needed to model human exposures at
Yucca Mountain are highly site-specific and reflect the desert conditions of the sparsely
populated Amargosa Valley. For example, the combined impacts of low rainfall, desert
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temperatures, and soil quality mandate extensive irrigation of farm crops and use of local ground
water for cattle. Under these conditions, contaminated well water has the potential for
developing unusually high radionuclide concentrations in all locally grown food products.
The Cricial Group Approach to Chatacterization of the Dose Receptor
The NAS report recommended the use of the critical group concept for the development of
environmental standards. The critical group concept was first introduced by the International
Commission on Radiation Protection (ICRP) in order to account for the variation in dose in a
given population which may occur due to differences in age, size, metabolism, habits, and
environment. This concept was adopted in total by the NAS panel, although the Academy differs
from ICRP in the implementation of the concept. The ICRP defines the critical group in dose
terms, while the NAS adapted the concept to individual risk. The critical group is defined by the
ICRP as a relatively homogeneous group of people whose location and lifestyle are such that they
represent those individuals expected to receive the highest doses (or be at highest risk) as a result
of radioactive releases. As part of the critical group definition, the ICRP specifies the following
additional criteria (also adopted by the NAS panel):
• Size - The critical group should be small in number and typically include a few to
a few tens of persons.
• Homogeneity among members of the critical group - There should be a relatively
small difference between those receiving the highest and the lowest doses. It is
recommended that the range between the low and high doses not differ by more
than a factor often or a factor of about three on either side of the critical group
average.
Magnitude of dose/risk - It is suggested that the regulatory limit defined by a
standard exceed the calculated average critical group dose by at least a factor of
ten.
• Modeling assumptions - In modeling exposure for the critical group, the ICRP
recommends that dose estimates be based on cautious, but reasonable
assumptions.
The ICRP does not, however, prescribe the lifestyle, habits, or conditions of exposure that may
define a critical group into the future. Its generic recommendations suggest use of current
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knowledge and cautious, but reasonable, assumptions for characterizing future exposure
scenarios.
To account explicitly for future uncertainties, the NAS report offered two probabilistic modeling
approaches. The first, described in Appendix C of the NAS report, A Probabilistic Critical
Group Approach, uses statistical methods and probability values to characterize members of the
critical group. The second, The Subsistence-Farmer Critical Group, described in Appendix D of
the report, also employs a probabilistic method, but identifies the subsistence farmer as the
principal representative of the critical group.
The NAS Subsistence Farmer Critical Group model is quite similar to the RMEI approach
(described below) that is used by EPA to characterize the dose receptor for purposes of
rulemaking. The model described in Appendix D of the NAS report specifies a priori one or
more subsistence farmers and makes assumptions designed to define a highly exposed farmer as
representative of the critical group. Subsistence farming does not exclude commercial farmers
who raise food for personal consumption, in addition to cash farm products. The NAS assumed
the subsistence farmer of the future would have nutritional needs consistent with those of a
present-day person. Like the subsistence farmer of today, most or all drinking water would be
obtained from an on-site well also used in the production of all consumed food. The subsistence
farmer is also assumed to live his/her entire life at the same location. Thus, the magnitude of the
dose to a subsistence farmer will largely be defined by the radionuclide concentrations in ground
water at the point of water withdrawal.
EPA 's Reasonably, Maximally Exposed Individual as the Dose Receptor
EPA has developed a method for estimating potential radiation doses based on the concept of the
reasonably, maximally exposed individual (RMEI). The RMEI concept, which involves
estimating the dose to a person assumed to be at high risk based on reasonable (i.e., not overly or
insufficiently conservative) assumptions, has been used in previous agency programs and
guidance.
The total population that might be exposed from ground water pathways is very small. There are
only about ten people living in community of Lathrop Wells in Amargosa Valley about 20 km
from the Yucca Mountain site. If this small population was defined as the critical group, the
exposure to the group would likely be on the same order as if the exposure was defined based on
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an RMEI living at that location. Thus, in the Yucca Mountain setting, there is no significant
difference between the critical group and the RMEI.
The basic approach for estimating doses to be incurred by the RMEI is to identify and
characterize the most important exposure pathway(s) and input parameters. By using maximum
or near-maximum (e.g., 95th percentile) values for one or a few of the most sensitive parameters,
while assuming average values for others, the resulting dose estimates should reasonably
correspond to the near-maximum exposures to any member of the exposed population. The
ultimate objective of the approach is to define an exposure well above average exposures, but
within the upper range of possible exposures. The RMEI is not intended to represent the most
extreme case.
ES.6.4 Exposure Scenarios
The EPA has considered four basic scenarios for estimating potential exposures of the RMEI in
the Yucca Mountain area. The scenarios involve characteristics of the region and represent
potential human habitation patterns and lifestyles in the Yucca Mountain region based on local
climatic, geologic, and hydrologic conditions.
(1) Subsistence (low technology) Farmer. In this scenario, the farmer is assumed to live in the
Yucca Mountain area and to be exposed chronically (both indoors and outdoors) to residual
concentrations of radionuclides in soil through all exposure pathways. Contaminated water from
the aquifer is the only source of water for these individuals. The location and habits of this
individual will be consistent with historical locations, and easily accessible water (approximately
30-40 km from the disposal system). All the individual's food and water would come from
contaminated sources.
(2) Commercial Farmer. Based upon economic factors and current technologies, certain areas
around Yucca Mountain are suitable for commercial crop production. These areas are either
currently being farmed (approximately 30 km from the Yucca Mountain disposal system) or
could be economically viable based upon reasonable assumptions, current technology, and
experience in other parts of the arid west. In addition, some parts of the region could possibly
support emerging technologies such as hydroponic applications and fish farming. Exposure
pathways in this scenario are the same as those described for the subsistence farmer.
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(3) Rural-Residential Person. In this scenario, individuals are assumed to live closer to Yucca
Mountain and to be exposed through the same pathways described for the subsistence farmer in
Scenario 1. However, in this case the residents are not assumed to be full-time agricultural
workers. Instead, these individuals work primarily out of the area and engage only in light
farming and recreational activities within it. Furthermore, it is assumed that all of the drinking
water ( 2 liters/day) and some of the food production will involve use of water contaminated with
radionuclides. This lifestyle is typical of most of the people currently living in the Amargosa
Valley.
(4) Domestic Use of an Underground Drinking Water Supply. Based upon current water usage in
the arid West, there could be an hypothetical water supply which could serve a community living
north of Interstate 95 closer to the repository site (inside the Nevada Test Site).
For each of these four scenarios, there are eight exposure pathways to be evaluated:
• External radiation from radionuclides in soil
Inhalation of resuspended soil and dust containing radionuclides
• Inhalation of radon and radon decay products from soil containing radium
• Incidental ingestion of soil containing radionuclides
• Ingestion of drinking water containing radionuclides transported from soil to
potable ground water sources
Ingestion of home-grown produce contaminated with radionuclides taken up from
soil
Ingestion of meat (beef) or milk containing radionuclides taken up by cows
grazing on contaminated plants (fodder)
Ingestion of locally-caught fish containing radionuclides
ES.6.5 Compliance Evaluation
The above discussion of receptor groups and exposure scenarios illustrates the factors involved
in assessing compliance with radiation protection standards. In practice, the critical group and
exposure scenarios to be used in assessing compliance with EPA's standards for Yucca Mountain
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will be implemented under regulations to be developed by the NRC in conformance with the
EPA standards. The NRC regulations will be the basis for review of DOE's License Application.
The License Application from DOE will include assessments of potential radionuclide releases
from the repository and assessment of compliance with regulatory standards under specified
exposure conditions. Because of the long time frames and uncertainties involved in predicting
repository performance, DOE will be required to demonstrate "reasonable expectation" of
compliance with the standards. The term "reasonable expectation" conveys the concept that
absolute numerical proof of compliance with the standards is neither necessary nor likely to be
obtainable.
One of the key factors in evaluating compliance with EPA's Yucca Mountain standards is the
radiation dose potential associated with each of the exposure pathways used by the receptor. The
dose potential is characterized in terms of dose conversion factors, which relate radionuclide
concentrations in the pathways for exposure, such as water and food consumed, to dose received.
The dose consequence of radionuclides in the environment therefore depends on the relative
importance of the various pathways for the exposed individual, which depends, in turn, on the
lifestyle of the exposed individual. It is to be expected, for example, that the pathways and dose
factors for a farmer residing in an arid environment, such as Yucca Mountain, will differ from
those for an urban resident.
Dose conversion factors for assessing compliance with regulatory standards have been evaluated
by DOE, EPA, and CNWRA for a wide variety of environmental conditions and receptor
lifestyles. The DOE is currently acquiring data to enable characterization of dose factors
specifically for environmental conditions and human activities in the Yucca Mountain region.
The DOE plans to use site-specific dose conversion factors in the License Application.
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CHAPTER 1
INTRODUCTION
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 EPA is promulgating in 40 CFR Part
197 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.3 These standards provide the basic framework to control the long-term storage and
disposal of three types of radioactive waste:
Spent nuclear fuel, if disposed of without reprocessing
High-level radioactive waste from the reprocessing of spent nuclear fuel
• Other radioactive materials that may be placed in the potential repository
The 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. However, the plans for
placement of these materials are very uncertain and therefore, for the purpose of the present
rulemaking, the information presented in this Background Information Document (BID) is
limited to spent nuclear fuel and high-level radioactive waste. More details about the current and
projected inventories of these wastes can be found in Chapter 5 of the BID.
1.1 PURPOSE AND SCOPE OF THE BACKGROUND INFORMATION DOCUMENT
This document presents the technical information used by EPA to understand the characteristics
of the Yucca Mountain site and to develop its rule, 40 CFR Part 197. The scope of the BED
encompasses the conceptual framework for assessing radiation exposures and associated health
risks. In general terms, this assessment discusses the radioactive source term characterization,
movement of radionuclides from the repository at Yucca Mountain through the appropriate
environmental exposure pathways, and calculations performed to date of doses received by
members of the general public.
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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The significant alternative models for site and engineered barrier performance are presented in
the BID to the extent necessary to portray the current understanding of the site and the major
uncertainties in that understanding. Most of the technical information discussed in the BID is
derived from investigations sponsored by the Department of Energy (DOE). DOE has conducted
years of research; most of what is known about Yucca Mountain and the performance of an
underground radioactive waste repository is the result of this research. However, where
appropriate, information from other sources is presented to supplement the DOE data base, to fill
data gaps, and to illustrate alternative conceptualizations of geologic processes and engineered
barrier performance.
The BID is not intended to be a technical critique of the investigations conducted by DOE and
other parties. Nor is it a regulatory compliance or criteria document. The BID is a summary of
the technical information considered by EPA in developing the rationale for, and specifics in,
40 CFR Part 197.
In addition, the BID discusses only those issues related to the disposal of radioactive wastes in a
geologic repository. Although additional disposal strategies have been examined by the U.S. and
other countries, a geologic repository continues to be the most promising. Technologies to
separate and transmute long-lived radionuclides in the waste to a stable form were examined
recently by the National Research Council. The Council concluded that such technologies do not
obviate the need for a geologic repository. The use of other disposal environments, such as the
seabed or natural or artificial islands, is fraught with political issues and therefore considered
infeasible. A final alternative of placing the waste into earth's orbit and accelerating it toward
the sun may be theoretically possible, but would require decades of technological development
and is likely to be much more costly than placing the waste in a geologic repository (NOR97).
Chapter 1 of the BID discusses EPA's regulatory authority for the current rulemaking and
summarizes the recommendations of the National Academy of Sciences report to Congress
entitled Technical Bases for Yucca Mountain Standards (NAS95). A summary of key events in
the history of EPA's rulemaking is also included. Chapter 2 provides a brief history of the
evolution of radiation protection activities in the United States as well as current U.S. regulatory
programs and strategies. A summary of key international programs for high-level waste disposal
is presented in Chapter 3. Chapter 4 describes U.S. programs for the management and disposal
of high-level radioactive waste and spent nuclear fuel. Current and projected inventories of spent
nuclear fuel and DOE defense high-level radioactive waste are presented in Chapter 5. Chapter 6
describes the methodology used by EPA for dose and risk estimation. Chapter 7 provides
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descriptions of the natural features of the Yucca Mountain site, the concepts under consideration
for the engineered features of a potential repository at the site, and analyses to date concerning
safety performance of a disposal system at the site. Chapter 8 describes the environment in the
. Yucca Mountain region, current conditions of human radiation exposure in the region, and
concepts that could be used to evaluate the consequences of radioactivity release from a
repository at Yucca Mountain. Chapter 9 discusses Yucca Mountain exposure scenarios and
compliance assessment issues, and finally, Chapter 10 provides a literature review of radiological
risks from alternatives to geologic disposal of high-level radioactive waste.
1.2 EPA'S REGULATORY AUTHORITY FOR THE RULEMAKING
The standards governing environmental releases from the Yucca Mountain repository have been
developed pursuant to the Agency's authorities under the Energy Policy Act (EnPA) of 1992
(Public Law 102-486). Section 801 of this Act directed EPA to promulgate standards to ensure
protection of public health from releases from radioactive material in a deep geologic repository
to be built at Yucca Mountain (EnPA92). EPA must set standards to ensure protection of the
health of individual members of the public. The EnPA also required EPA to contract with the
National Academy of Sciences (NAS) to advise the Agency on the technical bases for the Yucca
Mountain standards. These 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 (EnPA92).
1.3 THE NATIONAL ACADEMY OF SCIENCES RECOMMENDATIONS
In the EnPA, the Congress asked 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;
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• WJiether 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
(EnPA92).
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; 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 1914
contains an individual-dose standard, and it continues to rely on a
containment requirement that limits the releases of radionuclides 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.
* In 1985, EPA promulgated 40 CFR Part 191, "Environmental Standards for the Management and Disposal
of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes" (EPA85a). These are generally applicable
environmental standards promulgated under EPA's authority under the Atomic Energy Act of 1954 as amended. As
a result of court action, these standards were remanded back to EPA and were subsequently repromulgated in 1993.
(See Sections 1.3.1 and 1.3.4 for more detail.)
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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).
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 the following (NAS95, p. 9):
" ...physical and geologic processes are sufficiently quantifiable and the
related uncertainties sufficiently boundable that the performance can be
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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 oflO6 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):
...// 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.
...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.
1.4 HISTORY OF EPA'S RULEMAKING
Many significant events have occurred in the past 50 years concerning the management of high-
level radioactive waste and spent nuclear fuel. Table 1-1 provides a timeline of these events.
The following sections describe them in detail.
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Table 1-1.
Significant Events in the History of High-Level Radioactive Waste and Spent
Nuclear Fuel Disposal
1944
1949
1955
1957
1962
1965-1967
1968
1970
1970
1971
1974
1974
1976
1976
1976
1976
1978
1978
1979
1980
1981
Construction of first storage tanks for high-level radioactive waste (HLW)
The Atomic Energy Commission (AEC) initiates work to convert high-level liquid waste into a
stable form.
The National Academy of Sciences (NAS) Advisory Committee is established to consider disposal
of HLW in U.S.
The NAS suggests geologic disposal be investigated, particularly in naturally occurring salt
formations.
The AEC determines waste management to be technically feasible
Project Salt Vault demonstrates the safety and feasibility of handling and storing waste in salt
formations.
The AEC requests NAS to establish a Committee on Radioactive Waste Management (CWRM)
The CWRM concludes that the use of bedded salt is satisfactory for the disposal of radioactive
waste.
The AEC announces tentative selection of a site at Lyons, Kansas, for the establishment of a
national radioactive waste repository.
The AEC pursues alternative sites for repository.
The AEC publishes its first analysis of methods for long-term management of HLW
Congress passes the Energy Reorganization Act which abolishes AEC and creates a developmental
agency, the Energy Research and Development Agency (ERDA-now DOE) and an independent
regulatory commission, the Nuclear Regulatory Commission (NRC), which has authority to regulate
DOE facilities used for receipt and storage of HLW.
The Office of Management and Budget (OMB) establishes an interagency task force on commercial
•ILW.
The Federal Energy Regulatory Commission (FERC) publishes a status report on the management
of commercial radioactive waste.
'resident Ford issues a major policy statement on radioactive waste which includes a charge to the
IP A to issue general environmental standards governing releases of radioactive material to the
)iosphere.
The EPA announces its intent to develop environmental radiation protection criteria for radioactive
waste.
The EPA proposes criteria for management and disposal of radioactive wastes ||
President Carter establishes the Interagency Review Committee. j|
The DOE publishes a draft GEIS and decides to concentrate on mined geologic repositories as a
means for waste disposal. ||
resident Carter outlines a national radioactive waste management program. The President decides
o investigate four to five sites in a variety of environments before a license application is submitted
o NRC.
he EPA withdraws its proposed "Criteria for Radioactive Wastes."
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Table 1-1. Significant Events in the History of High-Level Radioactive Waste and Spent
Nuclear Fuel Disposal (continued)
ongress enacts the Nuclear Waste Policy Act which requires characterization of three sites and
construction of a geologic repository available to receive spent nuclear fuel and HLW by 1998.
982
The EPA proposes 40 CFR Part 191, "Environmental Standards for the Management and Disposal
of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes."
985
The EPA issues a final rule under 40 CFR Part 191.
987
Congress passes the Nuclear Waste Policy Amendments Act which identifies Yucca Mountain as
the single site for characterization.
987
The EPA's 40 CFR Part 191 is remanded by the Court.
1992
Congress enacts the Waste Isolation Pilot Plant Land Withdrawal Act which reinstated sections of
40 CFR Part 191 and exempted Yucca Mountain from the generic disposal standards set forth in
Subpart B of 40 CFR Part 191.
1992
Congress enacts the Energy Policy Act and directs EPA to develop regulations for Yucca Mountain
1993
The EPA issues amendments to 40 CFR Part 191.
1996
The DOE acknowledges it cannot proceed directly to License Application, but only to a
determination of site viability, by 1998.
1998
The DOE publishes a "viability assessment" concluding that Yucca Mountain is a promising site foi
a geologic repository and that work should proceed toward a site recommendation in 2001.
1999
The DOE published a "Draft Environmental Impact Assessment" for a geologic repository at Yucca
Mountain. .
2000
The DOE published the Total System Performance Assessment for Site Recommendation
1.4.1 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, and the EPA 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).
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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 established provisions
for the selection of at least two independent repository sites. Further, the NWPA limited the
quantity of spent nuclear fuel to be disposed of in the initial repository to 70,000 metric tons of
heavy metal (MTHM)5, or a quantity of solidified high-level radioactive waste resulting from the
reprocessing of such a quantity of spent nuclear fuel, until a second repository is inioperation
(NWP83). The NWPA also reiterated the existing responsibilities of the Federal agencies
involved in the national program and provided a timetable for several key milestones to be met
by the Federal agencies. As part of this national program, the 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).
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" (EPA85a).' These standards were 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 to the Agency for further
consideration.
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 the Yucca Mountain site as the location for the first high-level waste and spent
nuclear fuel repository (NWP87). Activities at all other potential sites were to be phased out. If
the Yucca Mountain site is found to be suitable, the President is required to submit a
recommendation to Congress to develop a repository at this location. In the event that site
characterization activities indicate that Yucca Mountain is an unsuitable site for the repository,
the Secretary of Energy is required to inform Congress and the State of Nevada of its findings.
The NWPAA prohibits DOE from conducting site-specific activities for a second repository
1 This is a measure of the uranium content of the spent nuclear fuel to be emplaced in the repository.
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unless authorized to do so by Congress. However, the NWPAA does require a report from the
Secretary of Energy on the need for a second repository no later than January 1, 2010.
Finally, the Act established a commission to study the need and feasibility of a monitored
retrievable storage facility to complement the nation's nuclear waste management program. The
commission submitted to Congress (required under the original Act, as amended by Public Law
100-507) a report outlining its recommendations on November 1, 1989 (NWP88, RMR89).
In October 1992, the Waste Isolation Pilot Plant Land Withdrawal Act (WIPP LWA) 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 (WDP92). However, the
EnPA directed the EPA to set site-specific radiation protection standards for the Yucca Mountain
disposal system (EnPA92).
As part of the Fiscal Year 1997 appropriation action, the Congress required EPA to perform a
comparative assessment of risks associated with management of commercial spent nuclear fuel
for three circumstances: permanent storage at the site where it is now stored; one or more
centralized storage sites; and deep geologic disposal at Yucca Mountain. This requirement was
established in Senate. Report 104-320 at page 98 and was retained by conference committee
action on the FY 1997 Energy and Water Appropriation Bill which stated that "The language and
allocations set forth in House Report 104-679 and Senate Report 104-320 should be complied
with unless specifically addressed to the contrary in the conference report and statement of the
managers" (Congressional Record, House, September 12, 1996, page HI0244).
The requirement was stated in Senate Report 104-320 as follows:
• Any radiation protection standard proposed by the Environmental
Protection Agency for the Yucca Mountain repository should consider
specific alternatives to deep geologic disposal at Yucca Mountain and
should include an analysis of the comparative risk to the public from each
alternative. The alternatives considered should include the permanent
storage of nuclear waste at the site where it is now stored and one or more
centralized storage sites recommended by the administration for the
above-ground, managed storage. The Agency shall evaluate each of these
alternatives against the standards proposed for deep geologic disposal at
Yucca Mountain.
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1.4.2 The Development of EPA's Role in the Federal Program
Since the inception of the nuclear age in the 1940s, the Federal government has assumed ultimate
responsibility for the disposal of spent nuclear fuel and high-level radioactive waste, regardless
of whether it is produced by commercial or national defense activities. In 1949, the Atomic
Energy Commission (AEC) initiated work aimed at developing systems for converting high-level
liquid waste into a stable form. Then, in 1955, at the request of the AEC, an NAS Advisory
Committee was established to consider the disposal of high-level radioactive waste within the
United States. Its report, issued in 1957, recommended that the AEC continue to develop
processes for the solidification of high-level radioactive liquid waste and that naturally occurring
salt formations be used as the medium for the long-term isolation of the solidified waste
(NAS57)
Project Salt Vault, conducted from 1965 to 1967 by the AEC in an abandoned salt mine near
Lyons, Kansas, demonstrated the safety and feasibility of handling and storing waste in salt
formations (McC70).
In 1968, the AEC again requested the NAS to establish a Committee on Radioactive Waste
Management (CRWM) to advise the AEC on its long-range radioactive waste management plans
and to evaluate the feasibility of disposing of solidified radioactive waste in bedded salt. The
CRWM convened a panel to discuss the disposal of radioactive waste in salt mines. Based on
the recommendations of the panel, the CRWM concluded that the use of bedded salt was
satisfactory for the disposal of radioactive waste (NAS70).
In 1970, the AEC announced the tentative selection of a site at Lyons, Kansas, for the
establishment of a national radioactive waste repository (AEC70). During the next two years,
however, in-depth site studies raised several questions concerning the safe plugging of old
exploratory wells and proposed expanded salt mining activities. These questions and growing
public opposition to the Lyons site prompted the AEC in late 1971 to pursue alternative sites
(DOU72).
In 1976, the Federal government intensified its program to develop and demonstrate a permanent
disposal method for high-level radioactive waste. The Office of Management and Budget
(OMB) established an interagency task force on commercial wastes in March 1976. The task
force defined the scope of the responsibility of each Federal agency's activities on high-level
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management, including the preparation of environmental standards for high-level waste by the
EPA (LYN76, ENG77a, ENG77b).
Shortly after the interagency task force was formed, the Federal Energy Regulatory Commission
(FERC) published a status report on the management of commercial radioactive waste. The
report, issued in May 1976, emphasized the need for coordination of administration policies and
programs relating to energy and called for an accelerated comprehensive government radioactive
waste program plan. The report also recommended that an interagency task force be formed to
coordinate activities among the responsible Federal agencies.
Subsequent to its findings, FERC established a nuclear subcommittee to coordinate Federal
nuclear policy and programs. The EPA was given the responsibility of establishing general
environmental standards governing waste disposal activities, including standards for high-level
radioactive waste to be delivered to Federal repositories for long-term management (FER76).
In October 1976, after the OMB interagency task force proposed its plan for spent nuclear fuel
and high-level waste management, President Ford issued a major policy statement on radioactive
waste. As part of his comprehensive statement, he announced new steps to assure that the United
States had the facilities for the long-term management of nuclear waste from commercial power
plants. He also reported that experts had concluded that the most practical method for disposing
of spent nuclear fuel and high-level radioactive waste would be in geologic repositories located
in stable formations deep underground. The EPA was charged with the responsibility of issuing
general environmental standards governing releases of radioactive material to the biosphere
above natural background radiation levels (FOR76). These standards were to place a numerical
limit on long-term radiation releases outside the boundary of the repository.
1.4.3 Early Federal Action
In December 1976, the EPA announced its intent to develop environmental radiation protection
criteria for radioactive waste to assure the protection of public health and the general
environment (EPA76). These efforts resulted in a series of radioactive waste disposal
workshops, held in 1977 and 1978 (EPA77a, EPA77b, EPA78a, EPA78b). Based on issues
raised during workshop deliberations, EPA published a Federal Register Notice on November 15,
1978 (43 FR 53262) (EPA78c) of intent to propose criteria for radioactive wastes and to solicit
public comments on possible recommendations for Federal Radiation Guidance. In this notice,
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EPA presented a set of criteria to address six key waste control decision issues: (1) the types of
materials to be categorized as radioactive wastes and subject to control; (2) the efficacy of
engineered controls and natural barriers to isolate wastes; (3) the usefulness of social institutions
in providing control, especially their viability over time; (4) the potential health risks of wastes
(over various time intervals and with differing levels of control); (5) the unacceptability of
various levels of risk; and (6) other considerations such as retrievability and communication of
waste disposal sites to succeeding generations to ensure continued isolation. As proposed, EPA
intended that the initial set of six criteria—each addressing one of the six key issues—would
serve collectively as the basis for developing environmental standards for different radioactive
waste sources.
During this time, President Carter established the Interagency Review Group (IRG) to develop
recommendations for an administrative policy to address the long-term management of nuclear
waste and supporting programs to implement the policy. The IRG report re-emphasized EPA's
role in developing generally applicable standards for the disposal of high-level waste, spent
nuclear fuel, and transuranic waste (DOE79). In a message to Congress in February 1980, the
President outlined the content of a comprehensive national radioactive waste management
program based on the IRG recommendations. The message called for an interim strategy for
disposal of spent nuclear fuel and high-level and transuranic wastes that would rely on mined
geologic repositories. The message reiterated that the EPA was responsible for creating general
criteria and numerical standards applicable to radioactive waste management activities (CAR80).
In March 1981, the EPA withdrew the proposed "Criteria for Radioactive Wastes" because it
considered the implementation of generic disposal guidance too complex given the many
different types of radioactive waste (EPA81).
In 1982, Congress enacted the NWPA, which established the current national program for the
disposal of spent nuclear fuel and high-level waste. The Act assigned DOE the responsibility of
siting, building, and operating an underground geologic repository for the disposal of these
wastes and directed the EPA to "promulgate generally applicable standards for the protection of
the general environment from off-site releases from radioactive material in repositories"
(NWP83). In that same year, under the authority of the AEA, the EPA proposed a set of
standards under 40 CFR Part 191, "Environmental Standards for the Management and Disposal
of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes" (EPA82).
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After the first comment period on the proposed rule ended in May 1983, the EPA held two public
hearings on the proposed standards—one in Washington, DC, and one in Denver, CO. During a
second public comment period, EPA requested post-hearing comments (EPA83a, EPA83b).
More than 200 comment letters were received during these two comment periods, and 13 oral
statements were made at the public hearings. Responses to comments received from the public
were subsequently published and released in August 1985 (EPA85b).
In parallel with its public review and comment effort, the EPA conducted an independent
scientific review of the technical bases for the proposed 40 CFR Part 191 standards through a
special subcommittee of the Agency's Science Advisory Board (SAB). The subcommittee held
nine public meetings from January to September 1983 and released a final report in February
1984 (SAB84). Although the SAB review found that the Agency's analyses in support of the
proposed standards were comprehensive and scientifically competent, the report contained
several findings and recommendations for improvement. The report was publicly released in
May 1984, and the public was encouraged to comment on the findings and recommendations
(EPA84). Responses to the SAB report were subsequently presented and released in August
. 1985 (EPA85c). • •
In February 1985, the Natural Resources Defense Council, the Environmental Defense Fund, the
Environmental Policy Institute, the Sierra Club, and the Snake River Alliance brought suit
against the Agency and the Administrator because they had failed to comply with the
January 1984 deadline mandated by the NWPA for promulgation of final standards. A consent
order was negotiated with the plaintiffs that required the standards to be promulgated on or
before August 15, 1985. The EPA issued the final rule under 40 CFR Part 191 on
August 15, 1985 (EPA85d, EPA85e).
1.4.4 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 (EPA85a).
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
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fuel cycle (40 CFR Part 190): 25 mrem to the whole body, 75 mrem to the thyroid, and 25 mrem
to any critical organ (EPA77c).
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 were:
Limits on cumulative releases of radioactive materials into the environment
during the 10,000 years following disposal
• Assurance requirements to compensate for uncertainties in achieving the desired
level of protection
Individual exposure limits based on the consumption of ground water and
any other potential exposure pathways for 1,000 years after disposal
• Ground water protection requirements in terms of allowable radionuclide
concentrations and associated doses for 1,000 years after disposal
(EPA85a)
Under sections 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. 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 5 pCi/L. Appendix A of the standards
provided acceptable radionuclide-specific 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
Arbitrary standards, not supported in the record, or not adequately explained
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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 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.
Outside the controlled area, the Court found that Section 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.
Section 191.15 permitted an annual dose of 25 mrem to the whole body and 75 mrem to any
critical organ from all pathways. 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 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
(Section 191.15). The Court indicated that the 1,000-year criterion was not inherently flawed,
but rather that the administrative record and the Agency's explanations did not adequately
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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 Section 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 issued on technical grounds about Section
191.16.
In August 1987, the Department of Justice petitioned the First Circuit Court to reinstate all of 40
CFRPart 191 except for Sections 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 Sections 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 section 191.16 of the original rule; altered the individual protection requirements; and
added Subpart C on ground water protection (EPA93). The amended standards represent the
Agency's response to the above legislation and to the issues raised by the court pertaining to
individual and ground water protection requirements. In so doing, EPA did not revisit any of the
regulations reinstated by the WIPP LWA.
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 standards for Yucca Mountain. The
EPA rule, 40 CFR Part 197, is responsive to this mandate.
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AEA54
AEC70
CAR80
DOE79
DOU72
ENG77a
ENG77b
EnPA92
EPA76
EPA77a
EPA77b
EPA77c
REFERENCES
Atomic Energy Act, Public Law 83-703, as amended, 42 USC 2011 et seq., 1954.
Atomic Energy Commission Press Release No. N-102, dated June 17, 1970.
The White House, President J. Carter, The President's Program on Radioactive
Waste Management, Fact Sheet, February 12, 1980.
U S Department of Energy, Report to the President by the Interagency Review
Group on Nuclear Waste Management, Report No. TID-29442, March 1979.
Doub, W.O., U.S. Atomic Energy Commission Commissioner, Statement before
the Science, 'Research and Development Subcommittee for the Committee on
Science and Astronautics, U.S. House of Representatives, U.S. Congress,
Washington, D.C., May 11 and 30,1972.
English, T.D. et al., An Analysis of the Back End of the Nuclear Fuel Cycle with
Emphasis on High-level Waste Management, JPL Publication 77-59, Volumes I
and II, Jet Propulsion Laboratory, Pasadena, California, August 12, 1977.
English, T.D. et al., An Analysis of the Technical Status of High-level Radioactive
Waste and Spent Fuel Management Systems, JPL Publication 77-69, Jet
Propulsion Laboratory, Pasadena, California, December 1, 1977.
Energy'Policy Act of 1992, Public Law 102-486, October 24, 1992.
U S. Environmental Protection Agency, Environmental Protection Standards for
High-level Wastes - Advance Notice of Proposed Rulemaking, Federal Register,
41 FR 53363, December 6, 1976.
U. S. Environmental Protection Agency, Proceedings: A Workshop on Issues
Pertinent to the Development of Environmental Protection Criteria for
Radioactive Wastes, Reston, Virginia, February 3-5, 1977, Office of Radiation
Programs, Report ORP/SCD-77-1, Washington, D.C., 1977.
U. S. Environmental Protection Agency, Proceedings: A Workshop on Policies
and Technical Issues Pertinent to the Development of Environmental Protection
Criteria for Radioactive Wastes, Albuquerque, New Mexico, April 12-17, 1977,
Office of Radiation Programs, Report ORP/SCD-77-2, Washington, D.C., 1977.
U S Environmental Protection Agency, Environmental Radiation Protection
Standards for Nuclear Power Operations, 40 CFR Part 190, Federal Register, 42
FR 2858-2861, January 13, 1977.
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EPA78a U. S. Environmental Protection Agency, Background Report - Consideration of
Environmental Protection Criteria for Radioactive Wastes, Office of Radiation
Programs, Washington, D.C., February 1978.
EPA78b U. S. Environmental Protection Agency, Proceedings of a Public Forum on
Environmental Protection Criteria for Radioactive Wastes, Denver, Colorado,
March 30 - April 1, 1978, Office of Radiation Programs, Report ORP/SCD-78-2,
Washington, D.C., May 1978.
EPA78c U. S. Environmental Protection Agency, Recommendations for Federal Guidance,
Criteria for Radioactive Wastes, Federal Register, 43 FR 53262-53268,
November 15, 1978.
EPA81 U. S. Environmental Protection Agency, Withdrawal of Proposed Regulations,
Federal Register, 46 FR 17567, March 19, 1981.
EPA82 U. S. Environmental Protection Agency, Proposed Rule, Environmental
Standards for the Management and Disposal of Spent Nuclear Fuel, High-Level
and Transuranic Radioactive Wastes, 40 CFR Part 191, Federal Register, 47 FR
58196-58206, December 29, 1982.
EPA83a U. S. Environmental Protection Agency, Environmental Standards for the
Management and Disposal of Spent Nuclear Fuel, High-level and Transuranic
Radioactive Wastes, Notice of Public Hearings, Federal Register, 48 FR 13444-
13446, March 31, 1983.
EPA83b U. S. Environmental Protection Agency, Environmental Standards for the
Management and Disposal of Spent Nuclear Fuel, High-level and Transuranic
Radioactive Wastes, Requests for Post-Hearings Comments, Federal Register, 48
FR 23666, May 26, 1983.
EPA84 U. S. Environmental Protection Agency, Environmental Standards for the
Management and Disposal of Spent Nuclear Fuel, High-level and Transuranic
Radioactive Wastes, Notice of Availability, Federal Register, 49 FR 19604-19606,
May8, 1984.
EPA85a 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.
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EPA85b U. S. Environmental Protection Agency, High-level and Transuranic Radioactive
Wastes - Response to Comments for Final Rule, Volume I, Office of Radiation
Programs, EPA 520/1-85-024-1, Washington, D.C., August 1985.
EPA85c U. S. Environmental Protection Agency, High-Level and Transuranic Radioactive
Wastes - Response to Comments for Final Rule, Volume II, Office of Radiation
Programs, EPA 520/1-85-024-2, Washington, D.C., August 1985.
EPA85d U. S. Environmental Protection Agency, High-Level and Transuranic Radioactive
Wastes - Background Information Document for Final Rule, Office of Radiation
Programs, EPA 520/1-85-023, Washington, D.C., August 1985.
EPA85e U. S. Environmental Protection Agency, Final Regulatory Impact Analysis - 40
CFR Part 191: Environmental Standards for the Management and Disposal of
Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, Office of
Radiation Programs, EPA 520/1-85-027, Washington, D.C., August 1985.
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.
FER76 Federal Energy Resources Council, Management of Commercial Radioactive
Nuclear Wastes - A Status Report, May 10, 1976.
FOR76 The White House, President G. Ford, The President's Nuclear Waste Management
Plan, Fact Sheet, October 28, 1976.
LYN76 ' Memorandum from J.T. Lynn, OMB to R. Train, EPA; R. Peterson, CEQ; R.
Seamans, ERDA, and W. Anders, NRC; Concerning the Establishment of an
Interagency Task Force on Commercial Nuclear Wastes, March 25, 1976.
McC70 McClain, W.C., and R.L. Bradshaw, Status of Investigations of Salt Formations
for Disposal of Highly Radioactive Power-Reactor Wastes, Nuclear Safety,
11 (2): 130-141, March-April 1970.
NAS57 National Academy of Sciences - National Research Council, Disposal of
Radioactive Wastes on Land, Publication 519, Washington, D.C., 1957.
NAS70 National Academy of Sciences - National Research Council, Committee on
Radioactive Waste Management, Disposal of Solid Radioactive Wastes in Bedded
Salt Deposits, Washington, D.C., November 1970.
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NAS95
NIX70
NWP83
NWP87
NWP88
RMR89
SAB84
USC87
WIP92
National Academy of Sciences - National Research Council, Committee on
Technical Bases for Yucca Mountain Standards, Technical Bases for Yucca
Mountain Standards, National Academy Press, Washington, D.C., 1995.
The White House, President R. Nixon, Reorganization Plan No. 3 of 1970,
Federal Register, 35 FR 15623-15626, October 6, 1970.
Nuclear Waste Policy Act of 1982, Public Law 97-425, January 7, 1983.
Nuclear Waste Policy Amendments Act of 1987, Public Laws 100-202 and 100-
203, December 22, 1987.
Nuclear Waste Policy Amendments Act of J988, Public Law 100-507 October 18
1988.
Nuclear Waste: Is There A Need For Federal Interim Storage?, Monitored
Retrievable Storage Review Commission, November 1, 1989.
Science Advisory Board, Report on the Review of Proposed Environmental
Standards for the Management and Disposal of Spent Nuclear Fuel, High-level
and Transuranic Radioactive Wastes (40 CFR Part 191), High-Level Radioactive
Waste Disposal Subcommittee, U.S. EPA, Washington, D.C., January 1984.
United States Court of Appeals for the First Circuit, Natural Resources Defense
Council, Inc., et al, v. United States Environmental Protection Agency,
Docket No.: 85-1915, 86-1097, 86-1098, Amended Decree, September 23, 1987.
Waste Isolation Pilot Plant Land Withdrawal Act, Public Law 102-579
October 20, 1992.
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CHAPTER 2
HISTORY OF RADIATION PROTECTION IN THE UNITED STATES
AND CURRENT REGULATIONS
2.1 INTRODUCTION
Radiation from cosmic rays and naturally occurring radioactivity contained in the earth make up
the natural radiation background environment in which all life forms have evolved. Society's
recognition of radiation began in 1895 with the discovery of X-rays; naturally occurring
radioactivity was observed in 1896. These discoveries marked the beginning of the study and use
of radioactive substances in science, medicine, and industry.
The discovery of radioactivity led rapidly to the development of medical radiology, industrial
radiography, nuclear physics, and nuclear medicine. By the 1920s, the use of X-rays in
diagnostic medicine and industrial applications was widespread. Radium was being routinely
used in luminescent dials and by doctors in therapeutic procedures. By the 1930s, biomedical
and genetic research scientists were studying the effects of radiation on living organisms, and
physicists were beginning to understand the mechanisms of spontaneous fission and radioactive
decay. In the 1940s, research in nuclear physics had advanced to the point where a self-
sustaining fission reaction was demonstrated under laboratory conditions. These events led to
the construction of the first nuclear reactors and the development of atomic weapons.
Today, the use of radiation, be it naturally occurring or man-made, is widespread and reaches
every segment of our society. Common examples include:
Nuclear reactors used: (1) to generate electricity, (2) to power ships and
submarines, (3) to produce radioisotopes used for research, medical,
industrial, space and national defense applications, and (4) as research
tools for nuclear engineering and physics
Particle accelerators used to produce radioisotopes and radiation and to
study the structure of matter, atoms, and common materials
Radioisotopes used in nuclear medicine, biomedical research, and medical
treatment
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X-rays and gamma rays used as diagnostic tools in medicine, as well as in
diverse industrial applications, such as industrial radiography, luggage X--
ray inspections, and nondestructive materials testing
Common consumer products, such as smoke detectors, luminous-dial wrist
watches, luminous markers and signs, cardiac pacemakers, lightning rods,
static eliminators, welding rods, lantern mantles, and optical glass
It was soon recognized that the use of radioactive materials would have to be controlled to
protect the public, workers, and the environment from radiation exposures. The following
sections present a brief history of the evolution of radiation protection activities, their principles
and concepts, and U.S. regulatory programs and strategies. Included in this discussion is the
influence that certain international advisory bodies, such as the International Commission on
Radiological Protection (ICRP), have had on the development of U.S. radiation protection
policies. Chapter 3 presents a summary of spent nuclear fuel and high-level waste disposal
programs in other countries.
2.2 THE INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION, THE
NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASUREMENTS,
AND THE INTERNATIONAL ATOMIC ENERGY AGENCY
Initially, the dangers and risks posed by X-rays and radioactivity were poorly understood. By
1896, however, "X-ray bums" were being reported in the medical literature, and by 1910, it was
understood that such "bums" could be caused by radioactive materials. By the 1920s, sufficient
direct evidence (from radium dial painters, medical radiologists, and miners) and indirect
evidence (from biomedical and genetic experiments with animals) had been accumulated to
persuade the scientific community that an official body should be established to make
recommendations concerning human protection against exposure to X-rays and radium.
In 1928, at the Second International Congress of Radiology meeting in Stockholm, Sweden, the
first radiation protection commission was created. Reflecting the uses of radiation and
radioactive materials at the time, the body was named the International X-Ray and Radium
Protection Commission. It was charged with developing recommendations concerning radiation
protection. In 1950, to better reflect its role in a changing world, the Commission was
reorganized and renamed the International Commission on Radiological Protection.
During the Second International Congress of Radiology, the newly created Commission
suggested to the nations represented at the Congress that they appoint national advisory
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committees to represent their viewpoints before the Commission and to act in concert with the
Commission in developing and disseminating recommendations on radiation protection. This
suggestion led to the formation of the U.S. Advisory Committee on X-Ray and Radium
Protection in 1929. In 1964, the Committee was congressionally chartered as the National
Council on Radiation Protection and Measurements (NCRP).
Throughout their existence, the ICRP and the NCRP have worked closely together to develop
radiation protection recommendations that reflect the current understanding of the risks
associated with exposure to ionizing radiation (ICR34, ICR38, ICR51, ICR60, ICR65). Neither
organization has official status, in that they do not have authority to issue or enforce regulations.
However, their recommendations often serve as the basis for the radiation protection regulations
adopted by the regulatory authorities in the United States and most other nations.
The International Atomic Energy Agency (IAEA) was chartered in July 1957 as an autonomous
intergovernmental organization under the aegis of the United Nations. The IAEA gives advice
and technical assistance to Member States on nuclear power development, health and safety
issues, radioactive waste management, and on a broad range of other areas related to the use of
radioactive material and atomic energy in industry and government. As is the case for ICRP and
NCRP, Member States do not have to follow IAEA recommendations. However, funding for
international programs dealing with the safe use of atomic energy and radioactive materials can
be withheld if Member States do not comply with IAEA recommendations. In addition, in
matters related to safeguarding special nuclear material, the full weight of the UN can be brought
to bear to "enforce" UN resolutions pertaining to the use of nuclear materials for peaceful
purposes. Many of the IAEA recommendations adopt ICRP recommendations with respect to the
Commission's radiation protection philosophy and numerical criteria.
In 1977, the ICRP released recommendations that are in use today. ICRP Publication No. 26
(ICR77) adopted the weighted, whole-body dose equivalent (defined as the effective dose
equivalent) concept for limiting occupational exposures. This approach reflected the increased
understanding of the differing radiosensitivities of various organs and tissues and was intended to
sum exposures from external sources and from internally deposited nuclides. (Note: The
concept of summing internal and external exposures to arrive at total dose had been mentioned as
early as ICRP Publication No. 1 [ICR60].)
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ICRP No. 26 defined the goal of radiation protection as the prevention or limitation of effects
from radiation exposure and the assurance that practices involving radiation exposure are
justified. The concept of collective dose equivalent for populations was also discussed. The
ICRP No. 26 recommendations represented the first explicit attempt to relate and justify
permissible radiation exposures with quantitative levels of acceptable risk. The ICRP concluded
that "...the mortality risk factor for radiation-induced cancers is about 10"4 per rem, as an average
for both sexes and all ages...." The risks of average occupational exposures (about 0.5 rem/year)
are roughly comparable to risks experienced in safe industries, 10~4 annually. At the permissible
limit of 5 rem/year, the risk is comparable with that experienced by some workers in occupations
having higher-than-average risk.
For members of the public, the ICRP considered that an annual risk in the range of 10~6 to 10"5
would likely be acceptable (ICR77). The ICRP recommended an annual individual dose limit of
100 mrem (1 mSv) from all radiation sources. However, the Commission also recognized that an
annual individual dose limit of 500 mrem (5 mSv) may be permissible, provided that the average
annual effective dose equivalent over a lifetime does not exceed the principallimit of 100 mrem
(1 mSv) (ICR85a). No dose limits for populations were proposed.
In 1979, the ICRP issued Publication No. 30 (ICR79) establishing the Annual Limit on Intake
(ALI) system for limiting the intake of radionuclides by workers. The ALI is the activity of a
given nuclide that would irradiate a person to the limit set in ICRP No. 26 for each year of
occupational exposure. It is a secondary limit, based on the primary limit of equivalent whole-
body irradiation, and applies to intake by either ingestion or inhalation. The recommendations of
ICRP No. 30 applied only to occupational exposures. In 1983, the ICRP issued a statement
(ICR84) clarifying the use of ALIs and Derived Air Concentrations (DACs) for members of the
public.
In 1985, the ICRP issued a statement (ICR85a) refining dose limits for members of the public.
ICRP No. 26 had endorsed an annual limit of 500 mrem, subject to certain conditions. In making
this endorsement, it was assumed that the conditions would, in practice, restrict the average
annual dose to about 100 mrem. In its 1985 statement, the Commission stated that the principal
limit was 100 mrem, while occasional and short-term exposures up to 500 mrem were thought to
be acceptable.
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The Commission has also published guidance for waste disposal (ICR85b) and for general
radiological protection (ICR91). The first of these, "Radiation Protection for the Disposal of
Solid Radioactive Waste," emphasizes an individual-risk approach that considers both the
probability of a breach of a disposal site and its consequence upon the critical group.
In 1987, the NCRP issued Report No. 91 (NCR87), which acknowledged the assumptions and
the basic thrust of the recommendations in ICRP Reports 26 and 30. In discussing risk estimates,
the NCRP noted in its report that new data were becoming available that might require changes '
in the current estimates. However, the value recommended in ICRP No. 26 of W4 per rem was
retained for a nominal lifetime somatic risk for adults.
The NCRP also noted that continuous annual exposure to 100 mrem gives a person a mortality
risk of about lO'5 annually, or approximately 10'3 in a lifetime (NCR87). Similar to the 1985
ICRP statement, annual limits of 500 mrem were recommended for infrequent exposures and 100
mrem for continuous (or frequent) exposures. These limits do not include natural background or
medical exposures.
In 1989, the IAEA issued reports 96 and 99 in its Safety Series (IAE89a, IAE89b). These
documents presented criteria and guidance for the underground disposal of nuclear waste. Safety
Series No. 99, "Safety Principles and Technical Criteria for the Underground Disposal of High-
Level Radioactive Wastes," set out basic design objectives to ensure that "humans and the human
environment will be protected after closure of the repository and for the long periods of time for
which the wastes remain hazardous." The report went on to state that for releases from a
repository due to gradual processes, the dose upper bound should be less than an annual average
dose value of 1 mSv (i.e., 100 mrem/yr)6 for prolonged exposures for individuals in the critical
group (defined as the members of the public whose exposure is relatively homogeneous and is
typical of individuals receiving the highest effective dose equivalent or dose equivalent from a
given radiation source). Finally, it suggested a risk upper bound of 10~5 per year for an individual
for disruptive events.
In 1990, the ICRP issued Publication 60, which broadened its recommendations to include a
wider range of exposure scenarios than had been previously addressed. Publication 60 also gave
6 The ICK-P has ad°Pted the international system of units (SI). Under this system, 1 Sv equals 100 rem As
such, 1 mSv equals 100 mrem.
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support to new concepts in the field of radiation exposure protection, most notably the ALARA
(as low as reasonably achievable) concept of worker protection optimization. The ALARA
principle suggests dose limits should be set at the lowest levels reasonably possible for a given
scenario. In recent years, several international organizations, including the Council of the
European Communities (CEC) and the Organization for Economic Cooperation and
Development/Nuclear Energy Agency's (OECD/NEA's) Committee on Radiation Protection and
Public Health (CRPPH), have worked to interpret this principle and develop guidelines for its
practical use (NEA94). The formality with which the ALARA principle has been adopted varies
widely internationally. In many cases, the ALARA principle is being applied only as part of a
nonqualified conceptual framework within which protection measures are implemented; in
other countries, the application of the ALARA approach to worker safety is becoming
increasingly formalized (OEC95a).
In recent years, the IAEA has been developing new international safety standards and guidance
documents. Foremost among these is "International Basic Safety Standards for Protection
Against Ionizing Radiation and for the Safety of Radiation Sources," known as BSS (Basic
Safety Standards, Safety Series 115-1). The BSS was approved by the IAEA Board of Governors
in 1994 and published as an interim document in December 1995. A joint effort of the Food and
Agricultural Organization of the United Nations, the International Labor Organization, the
OECD/NEA, the Pan-American Health Organization, and the World Health Organization, the
BSS-is notable primarily for its movement toward an integrated approach to managing exposure
risk in which potential but unlikely events (such as accidents) are evaluated along with
comparatively normal, likely scenarios for exposure. Previously, safety assessment had focused
only on comparatively normal, likely scenarios (OEC95a). IAEA has also been developing a
comprehensive set of safety standards for radioactive waste management called Radioactive
Waste Safety Standards (RADWASS). RADWASS includes a safety fundamentals document
entitled "The Principles of Radioactive Waste Management" and a safety standard document
entitled "Establishing a National Safety Standard for Radioactive Waste Management." Both of
these were approved by the IAEA Board of Governors and published in October 1995. Three
other safety standards (S-2, S-3, and S-6) addressing predisposal management of radioactive
waste, near-surface disposal of radioactive waste, and decommissioning are under review
(OEC95b). The entire RADWASS series is currently under review to ensure harmonization with
Safety Series Publications and BSS documents.
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Criteria development is also continuing through an IAEA Working Group on Principles and
Criteria for Radioactive Waste Disposal. The working group's focus includes post-closure
monitoring, optimization, retrievability, dose vs. risk, and safety indicators under different time
frames. The group's first report, entitled Safety Indicators in Different Time Frames for the
Safety Assessment of Underground Radioactive Waste Repositories, was published in 1994
(SNI95).
In recent years, the CEC has been developing directives on radiation safety standards for
radiation exposures established under European Atomic Energy Community (EURATOM)
agreements. In accordance with ICRP recommendations, the CEC suggested in 1993 that doses
to members of the public be limited to 100 mrem per year from all sources except medical and
that occupational doses be limited to 2,000 mrem annually. The CEC is also expected to propose
criteria for the shipment of radioactive waste among member countries and for the export of
radioactive waste to nonmember countries (OEC93).
Finally, in 1989, Radiation Protection and Nuclear Safety authorities in Denmark, Finland,
Iceland, Norway, and Sweden developed a set of safety criteria for the disposal of high-level
radioactive waste. Revised in 1993 after international review, the Nordic Principles are largely
consistent with other criteria developed on the international level. The Principles outline a
radiation protection approach employing the concept of optimization and an individual dose limit
of 0.1 millisievert (10 mrem) per year. Basic guiding objectives for HLW disposal programs
include reduction of burden for future generations, long-term environmental protection, and the
use of specific safety assurance measures. Finally, the Principles contain technical
recommendations for repository design, site geology, and closure (SNI95).
2.3 FEDERAL RADIATION COUNCIL GUIDANCE
The Federal Radiation Council (FRC) was established in 1959 by Executive Order 10831. The
Council arose as a result of new information that became available in the 1950s on the effects of
radiation. Before that time, only nongovernmental radiation advisory bodies (i.e., ICRP and
NCRP) existed, and their recommendations were not binding on users of radiation or radioactive
materials. The FRC was established as an official Government entity and included
representatives from all Federal agencies concerned with radiation protection. The Council
served as the primary coordinating body for all radiation activities conducted by the Federal
Government (FRC60a) and was responsible for:
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...advising the President with respect to radiation matters, directly or indirectly
affecting health, including providing guidance to all Federal agencies in the
formulation of radiation standards and in the establishment and execution of
programs of cooperation with States... .
The Council's first recommendations concerning radiation protection guidance for Federal
agencies were approved by President Eisenhower in 1960 (FRC60b). The guidance established
exposure limits for members of the general public. These included the yearly radiation exposure
of 0.5 rem per year for the whole body of individuals in the general population and an average
gonadal dose of 5 rem in 30 years for the general population (exclusive of natural background
and the purposeful medical exposure of patients).
The guidance also established occupational exposure limits, which differed only slightly from
those recommended by the NCRP and ICRP at the time (NCR54, NCR59). The guidance
included:
Whole body, head and trunk, active blood-forming organs, gonads or lens
of the eyes are not to exceed 3 rem in 13 consecutive weeks, and the total
accumulated dose is limited to 5 rems multiplied by the number of years
beyond age 18, expressed as 5(N-18), where N is the current age
Skin of the whole body and thyroid are not to exceed 10 rem in
13 consecutive weeks or 30 rem per year
• Hands, forearms, feet, and ankles are not to exceed 25 rem in
13 consecutive weeks or 75 rem per year
• Bone is not to exceed 0.1 microgram of radium-226 or its biological
equivalent
Any other organs are not to exceed 5 rem in 13 consecutive weeks or
15 rem per year
In addition to the formal exposure limits, the guidance also established as Federal policy that any
radiation exposure should be justified and that "...every effort should be made to encourage the
maintenance of radiation doses as far below this guide as practicable...." Both of these concepts
had previously been proposed by the ICRP. The inclusion of the requirements to consider
benefits and keep all exposures to a minimum was based on the possibility that there is no
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threshold for radiation. The linear, nonthreshold, dose-response relationship was assumed to
place an upper limit on the estimate of radiation risk. However, the FRC explicitly recognized
that it might also represent the actual level of risk.
Following the issuance of this initial guidance, the FRC continued to provide guidance on a
number of radiation protection matters. In 1970, the Council was dissolved, and its functions
were transferred to the Environmental Protection Agency under authority of Reorganization Plan
No. 3 (NIX70).
2.4 ENVIRONMENTAL PROTECTION AGENCY
Since its creation in 1970, the EPA has issued regulatory standards regarding radiation hazards
from a number of different sources, including underground mining (EPA71), the uranium fuel
cycle operations (EPA77), uranium and thorium mill tailings (EPA83), radionuclide air
emissions (EPA89a), and management and disposal of spent nuclear fuel and high-level and
transuranic radioactive wastes (EPA93). Recently, EPA issued compliance criteria for the WIPP
(EPA96). EPA is currently developing a standard for the disposal of contaminated soil at
decommissioned sites, including Federal facilities.
The Agency has also exercised its authority to issue Federal guidance to limit radiation exposures
to workers (EPA87), as well as to the general public. In December 1994, EPA issued proposed
Federal guidance to update the previous Federal Radiation Protection Guidance for Exposure to
the General Public which was originally adopted in 1960 and 1961 (EPA94). The Agency is now
finalizing these new recommendations.
EPA has also provided extensive technical information regarding the assessment of risk from
radiation hazards. Specific examples of such information include radionuclide intake limits,
occupational radiation doses, biological parameters, and dose conversion factors (EPA88). This
information has been used extensively in the development of EPA standards and guidance, as
well as specific site assessments.
In addition to its responsibility to provide Federal guidance on radiation protection, the EPA has
various statutory authorities and responsibilities for regulating exposure to radiation. The
standards and regulations that EPA has promulgated and proposed with respect to controlling
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radiation exposures are summarized in the following paragraphs. Their applicability to EPA's
proposed standards under 40 CFR Part 197 is also discussed.
2.4.1 Environmental Radiation Exposure
The Atomic Energy Act (AEA) of 1954, as amended, and Reorganization Plan No. 3 granted the
EPA the authority to establish generally applicable environmental standards for exposure to
radiation (AEA54, NIX70). The AEA is the cornerstone of current radiation protection activities
and regulations. In 1977, pursuant to this authority, the EPA issued standards limiting exposures
from operations associated with the light-water reactor fuel cycle (EPA77). These standards,
under 40 CFR Part 190, cover normal operations of the uranium fuel cycle. The standards limit
the annual dose equivalent to any member of the public from all phases of the uranium fuel cycle
(excluding radon and its daughters) to 25 mrem to the whole body, 75 mrem to the thyroid, and
25 mrem to any other organ. To protect against the buildup of long-lived radionuclides in the
environment, the standards also set normalized emission limits for krypton-85, iodine-129, and
plutonium-239 combined with other transuranics with a half-life exceeding one year. The dose
limits imposed by the standards cover all exposures resulting from radiation and radionuclide
releases to air and water from operations of fuel-cycle facilities. The development of these
standards took into account both the maximum risk to an individual and the overall effect of
releases from fuel-cycle operations on the population, and balanced these risks against the costs
of effluent control.
2.4.2 Environmental Impact Assessments
In 1969, Congress passed the National Environmental Policy Act (NEPA), which declared a
national policy that encouraged a productive and enjoyable harmony between the public and the
environment (NEP70). The Act recognized the profound impact of human activity on the
interrelations of all components of the natural environment and sought to promote efforts to
prevent or eliminate damage to the environment. To this end, the national policy is geared
towards increasing the understanding of the ecological systems and natural resources important
to the United States. In addition, the Act established a Council on Environmental Quality to
assist the President in determining the state of the environment and developing environmental
policy initiatives.
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The Act also directed all Federal agencies to use a systematic, interdisciplinary approach to
ensure the integrated use of natural, social, and environmental sciences in support of plans and
decisions that have a potential impact on the environment. Specifically, it mandated that a
detailed Environmental Impact Statement (EIS) be submitted for any major action proposed by a
Federal agency or for legislation that would significantly affect the quality of the environment.
The EIS must describe any adverse environmental effects that the proposal would cause,
alternatives to the proposed action, effects of the project on the long-term productivity of the
environment, and any irreversible and irretrievable commitment of resources involved in the
proposed action. The EIS must also be prepared through consultation with any Federal agency
having jurisdiction or special expertise regarding the project and its environmental impact.
The Final EIS prepared by the Department of Energy for the Yucca Mountain site must comply
with NEPA requirements.
2.4.3 Ground Water Protection
The Safe Drinking Water Act (SOWA) 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 picoCuries per
liter (5 pCi/L), and the MCL for gross alpha particle activity (including radium-226, but
excluding radon and uranium) is 15 pCi/L. 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 millirem to the total body or any internal organ. For tritium and
strontium-90, the MCLs are 20,000 pCi/L and 8 pCi/L, 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
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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
MCLGs, the EPA also committed itself to evaluating the feasibility, costs, and availability of
water treatment technologies, as well as other practical considerations. The proposed regulations
state the following MCLs: radium-226, 20 pCi/L; radium-228, 20 pCi/L; radon-222, 300 pCi/L;
uranium, 20 micro g/L; adjusted gross alpha, 15 pCi/L; and beta and photon emitters, 4 mrem
ede/yr.
Over the past 20 years, the EPA has used two different methods to calculate radioactivity
concentrations for beta particle and photon emitting radionuclides in drinking water
corresponding to the MCL of 4 mrem/yr. Each method incorporates successive improvements in
the risk models and dose conversion factors for ingested radioactivity recommended by national
and international advisory committees on radiation protection and adopted by the Agency.
The first method is a requirement (§141.6(b)) of EPA's 1976 Interim Regulations. It specifies
that, with the exception of tritium and strontium-90, the concentration of beta/photon emitters
causing 4 millirem (mrem) total body or organ dose equivalent shall be calculated on the basis of
a 2 liter per day drinking water intake using the 168 hour data listed in Handbook 69 of the
National Bureau of Standards (NBS63). The dose models used in preparing Handbook 69 are
based on earlier recommendations of the International Commission oh Radiological Protection
(ICR60). For tritium and strontium-90, the EPA provides derived activity concentrations in
' Table A of § 141 -6(b) based on specific dose models for these nuclides.
The second method is presented in EPA's 1991 proposed rule on final drinking water standards
for radionuclides (EPA91). This method is based primarily on the updated dosimetric data in
ICRP Publication 30 (ICR79) and uses the Agency's own risk assessment methodology
formalized in the RADRISK computer code (DUN80). Under this approach, concentration
levels are calculated for each radionuclide individually by limiting the dose to the total body
(i.e., the effective dose equivalent or ede) to 4 mrem/yr ede, rather than on a dose rate of
4 mrem/yr to the critical organ. Similar to the first method, the second method assumes
continuous intake of activity over a lifetime at a rate of 2 liters of drinking water per day.
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2.4.4 Radionuclide Air Emissions
In December 1979, the EPA designated radionuclides as hazardous air pollutants under Section
112 of the Clean Air Act (CAA) Amendments of 1977 (Public Law 95-95) (EPA79). In April
1983, the EPA proposed standards regulating radionuclide emissions from four source categories,
one of which included U.S. Department of Energy (DOE) facilities. The rule established annual
airborne emission limits for radioactive materials and specified that annual doses resulting from
such emissions should not exceed 25 mrem to the whole body and 75 mrem to any critical organ
for members of the general public. The EPA also proposed not to regulate several other
categories of facilities, including high-level radioactive waste disposal facilities. EPA based its
decision with respect to high-level waste disposal facilities on estimated releases from conceptual
repositories that indicated that the airborne exposure pathway would not cause doses high enough
to warrant regulation.
In October 1984, following a court order, the EPA withdrew the proposed emission standards
based on the findings that the control practices already in effect protected the public from
radionuclide releases with an ample margin of safety. The Agency also affirmed its position not
to regulate other categories of emission sources, including uranium fuel facilities and high-level
radioactive waste.
In December 1984, a U.S. District Court found the EPA in contempt of its order and directed the
EPA either to issue final radionuclide emission standards or make a finding that radionuclides
are not hazardous air pollutants. The EPA complied with the court order in 1985 by issuing
standards for selected sources (EPA85a, EPA85b). As a result of the decision in National
Resources Defense Council Inc. v. EPA, November 1987, the Agency submitted a motion to the
court requesting a voluntary remand of its national emission standards for the four original
categories of emission sources proposed in April 1983. In December 1987, the Court granted the
EPA's motion for voluntary remand and established a schedule to propose new regulatory
standards within one year. The Court decision also defined the analytical process under which
the EPA was to re-evaluate its standards. Two steps were identified: (1) determine what is safe,
based exclusively on health risk, and (2) adjust the level of safety downward to provide an ample
margin of safety.
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In March 1989, the EPA issued a proposed rule for regulating radionuclide emissions under the
CAA following the re-examination of the regulatory issues associated with the use of Section
112 (EPA89a). The rule proposed four policy alternatives to control emissions and
risks from 12 categories of sources. Each of the four approaches considered the acceptable risk
criterion differently. The four approaches were:
Case-by-Case Approach: Acceptable risk considers all health information,
risk measures, potential biases, assumptions, and quality of the
information. The maximum individual lifetime fatal cancer risk must not
exceed 1 x 10"4.
• Incidence-Based Approach: Based on the best estimate of the total
incidence of fatal cancer. The proposed acceptable level of incidence must
not exceed one fatal cancer per year per source category.
Maximum Individual Risk Approach (1 Odorless): Only risk indicator
considered is the best estimate of the maximum individual lifetime risk of
fatal cancer. The maximum individual lifetime risk must not exceed
1 x 10"4.
Maximum Individual Risk Approach (10~6 or less): This approach is
similar to the previous one. The maximum individual lifetime risk,
however, must not exceed 1 x 10"6.
Consistent with the two-step process established by the Court, the Agency determined an ample
margin of safety after ascertaining a safe level based solely on health risks. In reaching its final
decision, the EPA considered all health risk measures, as well as technological feasibility, costs,
uncertainties, economic impacts of control technologies, and any other relevant information.
In its radionuclide emission standards, EPA considered a lifetime risk to an individual of
approximately 1 in 10,000 as acceptable. The presumptive level provides a benchmark for
judging the acceptability of maximum individual risk, but does not constitute a rigid line for
making that determination.
In its final rule, EPA concluded that there was no need to establish air emission standards for
high-level waste disposal repositories since anticipated operations at the site would be governed
by 40 CFR Part 191. Radioactive materials received at such facilities are sealed in containers.
Normal operations do not require additional processing or handling because spent nuclear fuel or
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high-level waste is received and emplaced into the ground in its original containers. Operations
at the disposal site, which may require additional waste processing or repackaging before the site
is declared a disposal facility, are covered by 40 CFR Part 191 and must comply with Subpart I
of the National Emission Standards for radionuclides7 (EPA89b). Consequently, the Agency
believed there is an ample margin of safety since the likelihood of releases, and attendant risks, is
very low.
2.4.5 Disposal of High-Level Radioactive Waste and Spent Nuclear Fuel
Congress passed the Nuclear Waste Policy Act (NWPA) of 1982 to provide for the development
of repositories for the disposal of high-level radioactive waste and spent nuclear fuel, and to
establish a program of research, development, and demonstration regarding this disposal
(NWP83). The Act established a schedule for the siting, construction, and operation of
repositories that would provide a reasonable assurance that the public and environment would be
adequately protected from the hazards posed by high-level radioactive waste. The Secretary of
Energy was charged with nominating candidate sites for a repository and following a number of
steps through a process of Presidential and Congressional approval, site characterizations, public
participation, and hearings. The Act also required the Secretary to adhere to NEPA in
considering alternatives and to prepare an EIS for each candidate site.
Initially the Act called for the development of two mined geologic repositories. The first
repository was to be selected from nine candidate sites in western states; the second repository
was to be located in the eastern United States in crystalline rock. EPA was charged with the
responsibility of promulgating generally applicable standards for the protection of public health
and the environment from off-site releases from radioactive material in repositories. The NRC,
in turn, was responsible for promulgating technical requirements and criteria consistent with
EPA's standards to serve as the basis for approving or disapproving applications regarding the
use, closure, and post-closure of the repository. The Act also discussed interim waste storage
requirements, as well as the payment of benefits to affected States and tribal groups to allow
them sufficient resources to participate fully in the process.
7 Subpart I of the National Emission Standard can be found in 40 CFR Part 61.101 and is entitled "National
Emission Standard for Radionuclide Emissions from Facilities Licensed by the Nuclear Regulatory Commission
(NRC) and Federal Facilities Not Covered by Subpart H." Subpart H of the National Emission Standard addresses
radionuclide standards for DOE facilities.
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In 1987, the NWPA was amended to reflect a redirection of the nuclear waste program. The
generic nature of the original act was changed to reflect the selection of the Yucca Mountain site
in Nevada as the only candidate site for the repository (NWP87). The State of Nevada was also
identified as the affected community. All site-specific activities at other candidate sites were
phased out, and the Final EIS, necessary for compliance with the NEPA, was to be prepared
specifically for the Yucca Mountain site without further consideration of alternative sites. The
redirection charged DOE with reporting to Congress on the potential social, economic, and
environmental impacts of locating the repository at Yucca Mountain.
2.4.5.1 Generic Disposal Standards for High-Level and Transuranic Wastes
As discussed in Chapter 1, the First Circuit Court of Appeals remanded Subpart B of EPA's
standards for the management and disposal of spent nuclear fuel and high-level and transuranic
waste (40 CFR Part 191) in 1987. (See Section 1.3.4 for additional detail regarding the Court's
action on 40 CFR Part 191.) The Waste Isolation Pilot Plant Land Withdrawal Act (WIPP LWA)
of 1992 reinstated all of the disposal standards remanded by the First Circuit Court of Appeals in
1987 except the three aspects of the individual and ground water protection requirements that
were the subject of the court remand (WEP92). It then put the Agency on a schedule for issuing
the final disposal standards. They were published in December 1993. The law also provided an
extensive role for EPA in reviewing and approving various phases of DOE activities at the WIPP
and required EPA to certify whether the WIPP repository would meet the final 40 CFR Part 191
standards. Finally, and of greatest importance to the current rulemaking, the WIPP LWA
exempted radioactive waste disposal activities at Yucca Mountain from compliance with the
generic standards set forth under the 40 CFR Part 191 standards.
2.4.5.2 Site-Specific Disposal Standards for High-Level Radioactive Waste
The Energy Policy Act (EnPA) of 1992 addressed energy efficiency throughout the United States
in different situations and for various types of fuel. Title VIII of the Act dealt specifically with
high-level radioactive waste. Section 801 of the EnPA assigned EPA the responsibility of
promulgating 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. EPA is
to prescribe a 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 (EnPA92). The Act also requires that the standards developed be based upon and
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consistent with the findings and recommendations of the NAS. Specifically, the NAS was
charged with considering: the use of a dose-based standard, the reasonableness of post-closure
oversight in preventing breaches, and the predictability of human intrusion over a period of
10,000 years. NAS's findings and recommendations were published on August 1, 1995, in its
report Technical Bases for Yucca Mountain Standards (NAS95). These standards will apply
only to Yucca Mountain.
2.4.6 Evaluation of Radiation Dose
The radiation dose incurred by an exposed individual is evaluated using the "committed effective
dose equivalent" (CEDE) concept. The CEDE is the weighted sum of the "committed dose
equivalent" to specified organs and tissues. The committed effective dose equivalent is the total
effective dose equivalent, averaged over a given tissue or organ, that is deposited in the 50-year
period following the intake of a radionuclide.
The CEDE approach to dose evaluation therefore takes into account the differing dose effects of
various radionuclides in specific parts of the body over time, and the differing dose effects of
external exposure to ionizing radiations of different types and energy levels. It accounts, for
example, for the fact that some radionuclides that are taken into the body will be rapidly excreted
after ingestion or inhalation, so that the dose effect is small. Other radionuclides may be retained
indefinitely in specific organs so that if the decay rate is low and exposure continues over time,
the body burden of the dose source, and therefore the dose committed to the organ, will
continually increase with time. In general, the dose incurred will depend on the types and
concentrations of radionuclides present, the conditions and duration of exposure, the biological
half-life of the radionuclide in the body, and the effects of exposure on organs and tissues of the
body.
Ability to apply the CEDE approach to dose evaluation is the result of a decades-long
evolutionary process which has developed a data base for, and an understanding of, the
physiological effects of radiation exposure. A brief history of the evolution of information and
methodology for radiation dose evaluation, and a description of the CEDE methodology, are set
forth in EPA's Federal Guidance Report No. 11 (EPA88). This document also contains tables of
values for the committed dose equivalents per unit uptake for various radionuclides taken into the
body and for various body organs and tissues.
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In 1993, EPA issued a companion report, Federal Guidance Report No. 12 (EPA93a), which
tabulates dose coefficients for external exposure to photons and electrons emitted by
radionuclides distributed in air, water, and soil. The dose coefficient values provided in this
document are, like those in Federal Guidance Report No.l 1, intended to be used by government
agencies to calculate the dose equivalent to organs and tissues of the body for given exposure
conditions.
2.5 NUCLEAR REGULATORY COMMISSION
The NRC was created as an independent agency by the Energy Reorganization Act (ERA) of
1974 (ERA74); which abolished the AEC and moved the AEC's regulatory function to the NRC.
This Act, coupled with the AEA, as amended, provided the foundation for regulation of the
nation's commercial nuclear power industry. NRC regulations are issued under the U.S. Code of
Federal Regulations Title 10 Chapter 1.
The mission of the NRC is to ensure adequate protection of public health and safety, the national
defense and security, and the environment in the use of nuclear materials in the United States.
The NRC's,scope of responsibility includes regulation of commercial nuclear power reactors;
nonpower research, test, and training reactors; fuel cycle facilities; medical, academic, and
industrial uses of nuclear materials; and the transport, storage, and disposal of nuclear materials
and waste. In addition to licensing and regulating the use of byproduct, source, and special
nuclear material, the NRC is also responsible for assuring that all licensed activities are
conducted in a manner that protects public health and safety. The NRC assures that none of the
operations of its licensees expose an individual of the public to more than 100 mrem/yr from all
pathways (NRC91).
The dose limits imposed by the EPA's standards for uranium fuel-cycle facilities (40 CFR Part
190) apply to the fuel-cycle facilities licensed by the NRC. These facilities are prohibited from
releasing radioactive effluents in amounts that would result in doses greater than the 25 mrem/yr
limit imposed by that standard. Currently, NRC-licensed facilities are also required to operate in
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accordance with the requirements of the CAA (40 CFR Part 61), which limits radionuclide
emissions into the air (EPA89b).8
The NRC exercises its statutory authority over licensees by imposing a combination of design
criteria, operating parameters, and license conditions at the time of construction and licensing. It
assures that the license conditions are fulfilled through inspection and enforcement activities.
2.5.1 Fuel Cycle Licensees
The NRC licenses and inspects all commercial fuel cycle facilities involved in the processing and
fabrication of uranium ore into reactor fuel. NRC regulations require an analysis of probable
radioactive effluents and their effects on the population near fuel cycle facilities. The NRC also
assures that all exposures are maintained as low as reasonably achievable (ALARA) by imposing
design criteria for effluent control systems and equipment. After a license has been issued, fuel-
cycle licensees must monitor their emissions and set up an environmental monitoring program to
assure that the design criteria and license conditions have been met.
2-5.2 Radioactive Waste Disposal Licenses
The NWPA, as amended, specifies a detailed approach for high-level radioactive waste disposal..
DOE has operational responsibility and the NRC has licensing responsibility for the
transportation, storage, and geologic disposal of the waste. The disposal of high-level
radioactive waste requires a determination of acceptable health and environmental impacts that
may occur over a period of thousands of years. Current plans call for the ultimate disposal of
waste in solid form in a licensed, geologic disposal system. The NWPA, as amended, designates
Yucca Mountain, Nevada, as the candidate site for the high-level waste repository.
The EnPA provides additional direction to the NRC as to its role in the licensing of a specific
disposal site at Yucca Mountain. Section 801 of the EnPA requires the Commission to modify
its technical requirements and criteria under section 121(b) of the NWPA of 1982, as necessary,
to be consistent with EPA's standards for the Yucca Mountain site. The NRC's requirements
8 Pursuant to Section 112(d)(9) of the CAA Amendments of 1990, EPA is proposing to rescind Subpart I as
it applies to NRC-hcensed facilities. The NRC is proposing to adopt a constraint level rule which would limit
radionuclide airborne emissions to 10 mrem/yr.
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and criteria shall assume that engineered barriers and post-closure oversight provided by the
DOE will be sufficient to: (1) prevent any activity at the site that poses an unreasonable risk of
breaching the repository's engineered or geological barriers and (2) prevent any increase in-the
exposure of individual members of the public to radiation beyond allowable limits (EnPA92).
NRC's original generic regulations governing deep geologic disposal (which were largely
developed prior to the EnPA) are contained in 10 CFR Part 60 entitled Disposal of High-level
Radioactive Wastes in Geologic Repositories (NRC81, NRC83). However, since the EnPA
specifies that sites for consideration be limited to Yucca Mountain and since the legislation
specifies the types of standards the Commission is to implement, NRC decided to promulgate
site specific standards for Yucca Mountain at 10 CFR Part 63. The proposed rule is entitled
Disposal of High-level Radioactive Wastes in a Proposed Geologic Repository at Yucca
Mountain, Nevada (Federal Register, February 22, 1999). The proposed rule applies only to
Yucca Mountain. The generic rule at 10 CFR Part 60 will be modified to indicate that does not
apply to Yucca Mountain nor can it be used as a basis for litigation in NRC's Yucca Mountain
licensing procedures. The proposed 10 CFR Part 63 regulations are summarized below. In
addition, the NRC promulgates (under 10 CFR Part 71) packaging criteria for the transportation
of spent nuclear fuel and high-level and transuranic radioactive wastes. Under 10 CFR Part 72,
the NRC licenses independent spent nuclear fuel storage facilities (NRC88).
Under the proposed 10 CFR Part 63, DOE is required to conduct site characterization activities
prior to submitting a license application and to regularly report on these activities to ;NRC.
When DOE submits the license application it must contain certain prescribed general information
and a Safety Analysis Report. The license application must be accompanied by an environmental
impact statement. The prescribed general information includes:
A general description of the proposed geologic repository ;
Proposed schedules for construction, receipt of waste, and emplacement of wastes
A detailed plan to provide physical protection of the waste
A description of the material control and accounting program
A description of the site characterization work
The Safety Analysis Report is a comprehensive document with 22 prescribed elements including
such items as a description and discussion of the engineered barriers system, an assessment of the
expected performance after closure, an explanation of how expert elicitation was used, and a
description of the quality assurance program.
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After review of the license application and the environmental impact statement, NRC may
authorize construction of the geologic repository operations area. In deciding whether to provide
such authorization to DOE, NRC will examine safety, common defense and security and
environmental values in making its determination that construction can begin.
The NRC may subsequently issue a license to DOE to receive nuclear waste if it finds that
construction has been substantially completed, that the proposed activities in the operations area
are in conformity with the application, that the issuance of a license is not inimical to common
defense and security and will not constitute and unreasonable risk to public health and safety, and
that adequate protective measures can be taken in the event of a radiological emergency at any
time before permanent closure. The NRC license will contain a variety of conditions relating to:
Restrictions on the physical and chemical form and radioisotopic content of the
waste
Restrictions on the size, shape, and materials and methods of construction of the
waste packages
• Restrictions on the volumetric waste loading
• Testing and inspection requirements to assure that any restrictions are met
• Controls to limit access and prevent disturbance of the site
Administrative controls to assure that site activities are conducted in a safe
manner and in accordance with license requirements
Once the waste has been emplaced, DOE is required to file an application to amend the license
for permanent closure. The DOE submission shall include, inter alia, a updated performance
assessment of the geologic repository, and a detailed plan for post-closure monitoring of the site
including land use controls, construction of monuments and preservation of records. Upon
completion of permanent closure activities and D&D of surface facilities, DOE can then apply
for an amendment to terminate the license.
2.5.3 Repository Licensing Support Activities
The current NRC repository licensing program consists of both proactive and reactive activities.
Proactive activities include developing and reviewing regulatory requirements and guidance to
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identify and resolve regulatory and technical uncertainties. Regulatory uncertainties exist where
regulatory requirements are ambiguous and could be subject to various interpretations. Technical
uncertainties are related to demonstrating compliance with a particular regulation.
The NRC staff is currently developing and implementing performance assessment models using
Yucca Mountain site data. The models will assist the NRC in performing a technical assessment
of the site, as well as identifying areas of regulatory and technical uncertainty during the license
application review process. The uncertainties identified must be addressed in a timely fashion so
that the NRC can meet the three-year license review schedule mandated by the NWPA.
Additional details are provided in Chapter 7.
These activities have produced licensing review plans in anticipation of the DOE submissions.
They include review of the SCP, Study Plan, and Quality Assurance Plan (QAP).
The major focus of pre-licensing activities has been on 10 key technical issues (KTIs) that NRC
has identified as being most important to repository performance. NRC's objective is to seek
staff-level resolution of these issues during pre-licensing consultation with DOE although the
procedure does not preclude rasing the issues during the licensing process. These issues are:
• Total system performance assessment
• Unsaturated and saturated flow under isothermal conditions
• Evolution of the near-field environment
• Container life and source term
• ' Repository design and thermal-mechanical effects
• Thermal effects on flow
• Radionuclide transport
• Structural deformation and seismicity
• Igneous activity ;
• Activities related to NRC high-level radioactive regulations
NRC periodically publishes Issue Resolution Status Reports (IRSRs) which provide DOE with
feedback on KTI subissues. For example, NRC published IRSR Revision 1, on total system
performance assessment and integration, in November 1998 (NRC98). The report documents the
acceptance criteria NRC proposes to use for addressing each identified KTI subissue and the
review method NRC plans to use in determining whether or not the each acceptance criterion has
been met. As of the November date, 18 subissues relating to total system performance
assessment and integration had been resolved and 13 remained open. ;
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Reactive activities of the repository licensing program consist of pre-licensing reviews that
follow DOE's sequence and schedule of activities. To date, the NRC has reviewed a number of
the QAPs proposed by DOE and its contractors for Yucca Mountain. Any quality assurance
issues identified must be resolved before significant data collection activities are performed at the
Yucca Mountain site.
The NRC has also provided formal comments to DOE on the 1998 TSPA Viability Assessment
(NRC99).
As site characterization activities proceed, the NRC will review DOE's semiannual progress
reports on the site characterization program. The review will focus on the resolution of
previously identified concerns and will evaluate new information about the site and repository
design. In addition, the NRC will review selected DOE study reports and position papers that
document the results of work performed to date, and topical and issue resolution reports that
summarize the site characterization work for specific licensing topics. These reviews will be
used to evaluate compliance with NRC regulations.
All concerns identified by the NRC will be tracked by its staff. The tracking system now being
implemented will focus not only on the issues identified, but also on DOE's progress towards
their resolution. The system also provides a licensing record of all NRC and DOE actions related
to resolving specific issues.
2.6 DEPARTMENT OF ENERGY
DOE operates facilities for the production and testing of nuclear weapons; for the management
and disposal of radioactive waste generated in national defense activities; for research and
development; and for the storage of spent nuclear fuel. In addition, DOE is conducting several
remedial action programs, such as the program for the management of uranium mill tailings and
the cleanup of sites formerly used for nuclear activities. These facilities and activities are not
licensed by the NRC. However, to protect public health and the environment, DOE has
implemented orders and procedures that are consistent with NRC regulations under 10 CFR Part
20 (NRC60), standards promulgated by the EPA, and other applicable Federal regulations and
guidelines.
DOE is also responsible for the disposal of spent nuclear fuel from the generation of electricity
by commercial nuclear reactors and high-level radioactive waste from defense activities. The
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facilities developed by the DOE for the management and disposal of these wastes must be
licensed by the NRC. The Yucca Mountain site in Nevada is the candidate location for disposal
of these wastes.
DOE is responsible for operating its facilities in a manner that is environmentally safe and sound,
as stated in DOE Orders 5400.1 (DOE88) and 231.1 (DOE95a). In meeting this mandate, DOE
has issued a number of orders specifying environmental standards and procedures. Many of
these orders are currently under review to determine their conformance with NRC and EPA
regulations and standards and will be revised in accordance with the applicable NRC or EPA
guidance. Key DOE orders pertaining to the management of radioactive and hazardous
materials include:
• DOE Order 460.1 A (DOE96b), which establishes administrative procedures for
the certification and use of radioactive and other hazardous materials packaging
by the DOE.
DOE Order 460.2 (DOE95b), which specifies DOE's policies and responsibilities
for coordinating and planning base technology for radioactive material and
transportation packaging systems. (Cancels DOE Orders 1540.1 A, 1540.2, and
1540.3 A-Change 1.)
DOE Order 451.1A (DOE97), which establishes procedures for implementing the
requirements of NEPA (NEP70). The order requires new facilities and existing
facilities with proposed modifications to submit EISs with their proposed facility
design or design modification. In addition, the facilities are subject to extensive
design criteria reviews to determine compliance. (Cancels DOE Order 451.1.)
In addition to the above orders, in March 1993, DOE published a Notice of Proposed
Rulemaking for 10 CFR Part 834, entitled Radiation Protection of the Public and the
Environment (58 FR 16268) (DOE93). The proposed rule contains DOE's internal primary
standards for the protection of the public and environment against radiation. The requirements
would be applicable to control of radiation exposures from normal operations under the authority
of DOE and DOE contractor personnel. In December 1996, DOE proposed revisions to its siting
guidelines in 10 CFR Part 960 which were specific to the Yucca Mountain site (DOE96a). DOE
has not yet taken final action on its proposal.
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Order 231.1, September 30, 1995.
DOE95b U.S. Department of Energy, Departmental Materials Transportation and
Packaging Management, DOE Order 460.2, October 26, 1995.
DOE96a U.S. Department of Energy, General Guidelines for the Recommendation of Sites
for Nuclear Waste Repositories; Proposed Rule and Public Hearing, 10 CFR Part
960, Federal Register, 61 FR 66158, December 16, 1996.
DOE96b U.S.' Department of Energy, Packaging and Transportation Safety, DOE Order
460.1 A, October 2, 1996.
DOE97 U.S. Department of Energy, National Environmental Policy Act Compliance
Program, DOE Order 451.1A, June 5, 1997.
DOE98 U.S. Department of Energy, Viability Assessment of a Repository at Yucca
Mountain, DOE/RW-0508, December 1998.
DUN80 Dunning, D.E. Jr., R. W. Leggett, and M.G. Yalcintas, A Combined Methodology
for Estimating Dose Rates and Health Effects From Exposure to Radioactive
Pollutants, ORNL/TM-7105, 1980.
EnPA92 Energy Policy Act of 1992, Public Law 102-486, October 24, 1992.
EPA71 U.S. Environmental Protection Agency, Radiation Protection Guidance for
Federal Agencies: Underground Mining of Uranium Ore, Federal Register, 36
FR 12921, July 9, 1971.
EPA76 U.S. Environmental Protection Agency, National Interim Primary Drinking Water
Regulations, EPA 570/9-76-003, 1976.
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EPA77 U.S. Environmental Protection Agency, Environmental Radiation Protection
Standards for Nuclear Power Operations, 40 CFR Part 190, Federal Register, 42
FR 2858-2861, January 13,1977.
EPA79 U.S. Environmental Protection Agency, National Emission Standards for
Hazardous Air Pollutants, ANPRM, Federal Register, 44 FR 46738, December 27,
1979.
EPA83 U.S. Environmental Protection Agency, Health and Environmental Protection
Standards for Uranium and Thorium Mill Tailings, 40 CFR Part 192, Federal
Register, 48 FR 602, January 5, 1983.
EPA85a U.S. Environmental Protection Agency, National Emission Standards for
Hazardous Air Pollutants, Standards for Radionuclides, Federal Register, 50 FR
5190-5200, February 6, 1985. :
EPA85b U.S. Environmental Protection Agency, National Emission Standards for
Hazardous Air Pollutants, Standards for Radon-222 Emissions from
Underground Uranium Mines, Federal Register, 50 FR 15386-15394, April 17,
1985.
EPA87 U.S. Environmental Protection Agency, Radiation Protection Guidance to
Federal Agencies for Occupational Exposure, Federal Register, 52 FR 2822-2834,
January 27,1987.
EPA88 U.S. Environmental Protection Agency, Limiting Values of Radionuclide Intake
and Air Concentration and Dose Conversion Factors for Inhalation, Submersion,
and Ingestion, Federal Guidance Report No. 11, Office of Radiation Programs,
EPA 520/1-88-020, Washington, D.C., September 1988.
EPA89a U.S. Environmental Protection Agency, National Emission Standards for
Hazardous Air Pollutants: Regulation of Radionuclides, 40 CFR Part 61,
Proposed Rule and Notice of Public Hearing, Federal Register, 54 FR 9612-9668,
March 7,1989.
EPA89b U.S. Environmental Protection Agency, National Emission Standards for
Hazardous Air Pollutants: Regulation of Radionuclides, 40 CFR Part 61, Final
Rule and Notice of Reconsideration, Federal Register, 54 FR 51695, Debember
15, 1989.
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.
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EPA93 U.S. Environmental Protection Agency, 40 CFR Part 191, Proposed Rule,
Environmental Radiation Protection Standards for the Management and Disposal
of Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, 58 FR
7924 - 7936, February 10, 1993.
EPA93a U.S. Environmental Protection Agency, External Exposure to Radionuclides in
Air, Water and Soil, Federal Guidance Report No. 12, Office of Radiation and
Indoor, EPA 402-R-93-081, September 1993.
EPA94 U.S. Environmental Protection Agency, Federal Radiation Protection Guidance
for Exposure of the General Public; Notice, Federal Register, 59 FR 66414,
December 23, 1994.
EPA96 U.S. Environmental Protection Agency, Criteria for the Certification and
Recertification of the Waste Isolation Pilot Plant's Compliance with the 40 CFR
Part 191 Disposal Regulations; Final Rule, Federal Register, 61 FR 5224-5245,
February 9, 1996.
EPA98 U.S. Environmental Protection Agency, Health Risks from Low-Level
Environmental Exposure to Radionuclides, Federal Guidance Report No. 13, Part
I - Interim Version, Office of Radiation and Indoor Air, EPA 402-R-097-014,
January 1998.
ERA74 Energy Reorganization Act, as amended, 1974.
FRC60a Federal Radiation Council, Radiation Protection Guidance for Federal Agencies,
Federal Register, 25 FR 4402-4403, May 18, 1960.
FRC60b Federal Radiation Council, Staff Report No. 1, Background Material for the
Development of Radiation Standards, May 13, 1960.
IAE89a International Atomic Energy Agency, Guidance for Regulation of Underground
Repositories for Disposal of Radioactive Wastes, Safety Series No. 96, Vienna,
Austria, 1989.
IAE89b International Atomic Energy Agency, Safety Principles and Technical Criteria for
the Underground Disposal of High-Level Radioactive Wastes, Safety Series No.
99, Vienna, Austria, 1989.
ICR34 International X-Ray and Radium Protection Commission, International
Recommendations for X-Ray and Radium Protection, British Journal of Radiology
7, 695-699, 1934.
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ICR38 International X-Ray and Radium Protection Commission, International
Recommendations for X-Ray and Radium Protection, American Journal of
Roentgenology and Radium, 40, 134-138, 1938.
ICR51 International Commission on Radiological Protection, International
Recommendations of Radiological Protection 1950, British Journal of Radiology,
24,46-53,1951.
ICR60 International Commission on Radiological Protection, Recommendations of the
International Commission on Radiological Protection Report of Committee lion
Permissible Dose for Internal Radiation (1959), ICRP Publication 2, Pergamon
Press, Oxford, 1960.
ICR65 International Commission on Radiological Protection, Recommendations of the
ICRP 1965, ICRP Publication 9, Pergamon Press, Oxford, 1965.
ICR77 International Commission on Radiological Protection, Recommendations of the
International Commission on Radiological Protection, ICRP Publication 26,
Annals of the ICRP Vol. 1, No.3, Pergamon Press, New York, New York, 1977.
ICR79 International Commission on Radiological Protection, Limits for Intakes of
Radionuclides by Workers, ICRP Publication 30, Part 1, Ann. ICRP 2 (3/4),
Pergamon Press, Oxford, 1979.
ICR84 Annals of the ICRP, Vol 14, No. 1, 1984, Statement from the 1983 Washington
Meeting of the ICRP.
ICR85a Annals of the ICRP, Vol. 15, No. 3, 1985, Statement from the 1985 Paris Meeting
of the ICRP.
ICR85b International Commission on Radiological Protection, Radiation Protection
Principles for the Disposal of Solid Radioactive Waste, ICRP Publication 46,
Pergamon Press, Oxford, 1985.
ICR91 International Commission on Radiological Protection, 1 990 Recommendations of
the International Commission on Radiological Protection, ICRP Publication 60,
Pergamon Press, Oxford, 1991.
NAS95 National Academy of Sciences - National Research Council, Committee on
Technical Bases for Yucca Mountain Standards, Technical Bases for Yucca
Mountain Standards, National Academy Press, Washington, D.C., 1995.
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NBS63 National Bureau of Standards (1963), Maximum Permissible Body Burdens and
Maximum Permissible Concentrations of Radionuclides in Air and Water for
Occupational Exposure, NBS Handbook 69 as amended August 1963, U.S.
Department of Commerce. Revised and republished in 1963 as NCRP Report No.
22 by the National Committee on Radiation Protection and Measurements.
NCR54 National Council on Radiation Protection and Measurements, Permissible Dose
from External Sources of Ionizing Radiation, National Bureau of Standards
Handbook 59, 1954.
NCR59 National Council on Radiation Protection and Measurements, Maximum
Permissible Body Burdens and Maximum Permissible Concentrations of
Radionuclides in Air and in Water for Occupational Exposure, National Bureau of
Standards Handbook 69, 1959.
NCR87 National Council on Radiation Protection and Measurements, Recommendations
on Limits for Exposure to Ionizing Radiation, NCRP Report No. 91, June 12,
1987.
NEA94 Nuclear Energy Agency, NEA Annual Report: 1994 Activities, 1994.
NEP70 National Environmental Policy Act of 1970, Public Law 91-190, January 1, 1970.
NIX70 The White House, President R. Nixon, Reorganization Plan No. 3 of 1970,
Federal Register, 35 FR 15623-15626, October 6, 1970.
NRC60 U.S. Nuclear Regulatory Commission, Standards for Protection Against
Radiation, 10 CFR Part 20, Federal Register, 25 FR 10914, November 17, 1960,
and as subsequently amended.
NRC81 U.S. Nuclear Regulatory Commission, Disposal ofHigh-Level Radioactive
Wastes in Geologic Repositories: Licensing Procedures, 10 CFR Part 60, Federal
Register, 46 FR 13971-13988, February 25, 1981.
NRC83 U.S. Nuclear Regulatory Commission, 10 CFR Part 60, Disposal of High-Level
Radioactive Wastes in Geologic Repositories, Technical Criteria, Federal
Register, 48 FR 28194-28229, June 21, 1983.
NRC88 U.S. Regulatory Commission, 10 CFR Part 72, Licensing Requirements for the
Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste,
53 FR 31658, August 19, 1988.
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NRC91 U.S. Nuclear Regulatory Commission, 10 CFR Part 20 et al., Standards for
Protection Against Radiation, Final Rule, Federal Register, 56 FR 98, May 21,
1991.
NRC98 U.S. Nuclear Regulatory Commission, Issue Resolution Status Report, Key
Technical Issue: Total System Performance Assessment and Integration, Revision
1, Division of Waste Management, Office of Nuclear Material Safety arid
Safeguards, November 1998.
NRC99 U.S. Nuclear Regulatory Commission, Staff Review of the U.S. Department of
Energy Viability Assessment for a High-Level Waste Repository at Yucca
Mountain, Nevada, Letter to Lake H. Barrett, OCRWM/DOE from Carl J.
Paperiello, ONMSS/NRC, June 2, 1999.
NWP83 Nuclear Waste Policy Act of1982, Public Law 97-425, January 7, 1983.
NWP87 Nuclear Waste Policy Amendments Act of1987, Public Law 100-203, December
22,1987.
OEC93 Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Nuclear Waste Bulletin: Update on Waste Management Policies and.
Programs, No. 8, July 1993. ;
OEC95a Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Radiation Protection Today and Tomorrow, 1995.
OEC95b Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Nuclear Waste Bulletin: Update on Waste Management Policies and
Programs, No. 10, June 1995.
SNI95 Snihs, Dr. J.O., "Radioactive Waste Disposal: Radiological Principles and
Standards," IAEA Bulletin, 1995..
WIP92 Waste Isolation Pilot Plant Land Withdrawal Act, Public Law 102-579, October
20/1992.
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CHAPTER 3
SPENT NUCLEAR FUEL AND HIGH-LEVEL WASTE DISPOSAL PROGRAMS
IN OTHER COUNTRIES
As in the United States, other countries that use nuclear power are establishing long-term
programs for the safe management and disposal of spent nuclear fuel and high-level radioactive
waste. Such programs include adopting a national strategy, assigning the technical responsibility
for research and development activities to designated agencies, selecting disposal strategies and
development activities, and setting the appropriate regulatory standards to protect public health
and the environment. Management strategies may include spent nuclear fuel storage at and away
from reactor sites, spent nuclear fuel reprocessing, high-level waste vitrification and storage,
partitioning and transmutation of the waste into short-lived or stable forms, and disposal in deep
geologic media. Typically, the objective of such geologic disposal programs is to immobilize
and isolate radioactive waste from the environment for a sufficient period of time under
conditions such that any radionuclide releases from the repository will not result in unacceptable
radiation exposure of the public. This strategy takes advantage of the geology surrounding the
disposal site to act as a passive barrier to radionuclide releases and eliminates many surface
factors, such as sabotage, hurricanes, theft, and flooding, which could compromise an above
ground facility. ,
As discussed in Chapter 1 of the BID, deep geologic disposal is considered by many in the
scientific community to be the most promising method for disposing of long-lived nuclear waste.
Consequently, several nations have begun activities associated with disposal of spent nuclear fuel
and high-level waste by isolation in deep geologic formations. These countries envision
emplacing solidified high-level waste in a deep geologic formation located within their borders.
Only the United States and Germany have identified candidate locations for disposal of high-
level waste, i.e., the Yucca Mountain site in Nevada and the Gorleben site in Germany9. Other
countries are, to varying degrees, engaged in technical evaluations of the potential suitability of
indigenous geologic formations for disposal. Some nations, such as France, have several
geologic formations, such as clay and granite, that might be used for disposal, and each
9 As will be discussed in Section 3.5.2, the suitability of the Gorleben site has been questioned by the new
German government which was elected in 1998.
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alternative is being evaluated. Others, such as Canada, have focused on one type of geologic
formation. (Canada is evaluating a crystalline rock formation in a setting with low seismic
activity.) In addition, several countries, such as Canada and Sweden, have established
underground research laboratories (URL's) and extended their research programs to include
participation by other nations with similar candidate geologies. For example, the Swedish
research facility is in a crystalline rock formation and its research program has included
participation by Japan, Spain, Finland, Switzerland, the U.S.10, and Canada.
The disposal strategies for all nations assume that waste isolation will be maintained by reliance
on a combination of engineered and natural barriers between the emplaced waste and the
environment. •Currently, the United States is considering a potential repository site at Yucca
Mountain, Nevada where the disposal facility would be located in an arid environment and
wastes would be emplaced in an unsaturated geohydrologic regime.
The geohydrologic features of the Yucca Mountain site, and the fact that most of the spent fuel
has not been reprocessed and is hot, allow the use of a thermal loading strategy in which heat
emissions can deter water from contacting waste packages for an extended period of time. The
combination makes Yucca Mountain unique in comparison with the options available in other
parts of the world.
Other countries are generally contemplating colder repositories for reprocessed spent fuel in
strata that are saturated with moisture, and thus must contemplate longer-term corrosive contact
between water and waste packages. ;
All countries, including the United States, have evolved toward using more robust engineered
barrier systems to compensate for the uncertainties in predicting the performance of natural
barriers.
For example, in response to a mandate from the national government in the 1970s, the Swedish
commercial nuclear waste program developed an engineered barrier concept involving
emplacement of spent nuclear fuel in a copper matrix contained within a highly-robust copper
10 The United States no longer actively participates in cooperative R&D programs with other countries.
However, the U.S. continues to exchange information with other countries through its bilateral agreements and its
representation in international agencies.
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canister. The viability of this concept to maintain wastes in isolation for one million years in
Sweden's geologic formations must be demonstrated, as required by the governmental directive.
Sweden has not, however, committed to the use of this engineered barrier concept.
Various nations and international agencies, in addition to the United States, have begun to give
consideration to regulations and regulatory standards for high-level waste disposal. Some
nations have developed broad risk or dose criteria, and some have supplemented such criteria
with additional qualitative technical criteria concerning features of the disposal system.
International organizations, such as the Nuclear Energy Agency in Paris, France, provide
opportunities for discussion of regulatory criteria and also provide programs of common interest,
such as comparison of performance assessment computer codes. There are, however, no
international standards for high-level waste disposal accepted by all nations.
Although the performance standards and criteria for the various national regulations are similar,
each nation has established specific requirements to meet its needs. Current information
concerning the provisions of national and international criteria and objectives for the safety of
long-lived radioactive waste disposal is presented in Table 3-1. As will be clear from the ensuing
discussion, regulatory requirements are still evolving and the information Table 3-1 is subject to
change.
Characteristics of programs in ten nations with major commitments to nuclear power and
existing activities concerning disposal of high-level wastes are summarized below. The
descriptions address nuclear power utilization, waste disposal, and regulatory programs.
Discussions are provided for programs in Belgium, Canada, Finland, France, Germany, Japan,
Spain, Sweden, Switzerland, and the United Kingdom.
3.1 BELGIUM
3.1.1 Nuclear Power Utilization
In 1994, Belgium met about 56 percent of its electrical needs through nuclear power (EIA95).
The Belgian nuclear power program relies on seven pressurized light water reactors, all of which
are operated by Electrabel, a privately-owned company.
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From 1966 to 1974, Belgium reprocessed spent nuclear fuel at its Eurochemic facility. The
company Belgoprocess was created as a consortium with foreign firms to reactivate the
Eurochemic plant, but these efforts failed in the mid-1980s. Belgoprocess is now responsible for
decommissioning the plant. Belgium had previously shipped spent nuclear fuel to France for
reprocessing, but stores current inventories in reactor pools and in dry storage facilities pending
the results of a future parliamentary debate on whether to continue reprocessing. France will
Table 3-1. National and International Criteria and Objectives for the Disposal of
Long-Lived Radioactive Wastes (OEC95a)
Organization/
Country/
NEA(1984)
ICRP
(Pub. 46, 1985)
IAEA
(Safety Series 99,
1989)
CANADA
Reg. Document
R-104, 1987)
FINLAND
Decision of the
Council
of State, 1991)
FRANCE
[Baste Safety
Rule,
RFSIII.2.f. 1991)
Main Objective/
For HLW:
max. indiv. risk < 10"5/y
(all sources)
For HLW, for individuals
(all sources):
1 mSv/y (normal
evolution scenarios);
10'Vy (probabilistic scenarios)
Idem ICRP Publication 46
For HLW:
max. indiv. risk objective:
< 10-Vy
For LLW and ILW:
max. indiv. dose
< 0.1 mSv/y, with max.
indiv. dose < 5 mSv/y from
accident conditions caused by
possible natural events or
human actions
For ILW and HLW:
max. indiv. dose
< 0.25 mSv/y for normal
evolution scenarios;
for altered evolution scenarios.
risk may be considered
(probability
of scenario times effect
of exposure)
Other Main Features
Both probability and dose
should be taken into
account in ALARA
Period of time for
demonstrating compliance:
10Jy
No sudden and dramatic
increase for times > 1 0'y
Beyond 10a y, dose limit
is considered as a
"reference" level
Criteria for Judging
Human Intrusion
(HI) Scenarios
Indiv. risk/dose - best
criterion to judge long-
term acceptability
Future human activities
should be treated
probabilistically
Future human activities are>
randomly disruptive events
that usually are examined
probabilistically
Main criteria applicable to
no criteria specific to HI
scenarios
Max. indiv. dose < 5 mSv/y
from possible human
actions
Assumptions (French Basic
Safety Rule, Appendix 2):
Date of HI occurrence
> 500 y;
Existence of repository and
location forgotten;
Level of technology same as
present day
Comments
No consensus on
ALARA/
optimization
ALARA useful, notably
to compare alternatives,
but may not be the most
important siting factor
Also includes qualitative
technical criteria on
disposal system features
and role of safety
analysis and quality
assurance
Additional qualitative,
requirements and
guidelines in regulatory
documents
For spent nuclear tuel or
HLW, proposed criterion
for max. indiv. dose <
0.1 mSv/y
Technical criteria tor
siting established in 1987
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Table 3-1. National and International Criteria and Objectives for the Disposal of
Long-Lived Radioactive Wastes (OEC95a) (Continued)
Organization/
Country/
Reference
GERMANY
(Section 45, para.
1 of Radiation
Protection
Ordinance, 1989)
NORDIC
COUNTRIES
(Basic Criteria
Document, 1993)
SWITZERLAND
(Reg. Document
R-21, 1993)
UNITED
KINGDOM
(OECD/NEA
Doc. 66-94-
041,1995)
Main Objective/
Objective/Criteria
For all waste types.
max. indiv. dose
< 0.3 mSv/y for all reasonable
scenarios
For all waste types:
max. indiv. dose
< 0.1 mSv/y (normal scenarios):
max. indiv. risk
< 10'Vy (disruptive events)
For all waste types:
max. indiv. dose
< O.I mSv/y at any time for
reasonably probable scenarios;
max. indiv. risk
<10"6/y for unlikely scenarios
For L/ILW:
10"°/y target for indiv. risk from
a single facility
For HLW:
no specific criteria but likely
application of principles similar
to existing objectives for L/ILW
Other Main Features
Calculation of individual
doses limited to 104y
but isolation potential
beyond 1 0Jy may be
assessed
For HLW, additional
criterion on "total
activity inflow" limiting
releases to biosphere,
based on inflow of natural
alpha radionuclides
Repository must be
designed in such a way
that it can at any time be
sealed within a few years
without the need for
institutional control
No time frame for
quantitative assessments
specified
Criteria for Judging
Human Intrusion
(HI) Scenarios
No criteria for HI scenarios
except that for high
consequences, probabilities
can be
taken into account
Main criterion for HI
scenarios currently indiv.
risk
Comments
Additional qualitative
technical criteria in
guidelines and
regulatory document
Includes other
qualitative criteria
ALARA to be used to the
extent practical and
reasonable
soon begin returning to Belgium the high-level vitrified waste created from the reprocessing of
Belgian fuel. It is estimated that by the year 2000, Belgium will have produced about
2,500 MTHM of spent nuclear fuel.
In 1985, a vitrification plant, PAMELA, began processing high-level waste from the Eurochemic
plant. Vitrified high-level waste will be stored in an intermediate storage facility (recently
constructed by the National Agency for Radioactive Waste and Fissile Materials (ONDRAF)) for
50-70 years. Characterization of a potential site for a repository located in a clay formation at the
Mol-Dessel site in the northeast corner of the country is progressing.
3.1.2 Disposal Programs and Management Organizations
The Belgian program to establish a radioactive waste repository was initiated in 1974 with the
establishment of a government-sponsored research and development initiative. In 1982, the
National Agency for Radioactive Waste and Fissile Materials, ONDRAF, was established to
implement and manage a multi-year national program addressing the long-term management and
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disposal of radioactive wastes, including spent nuclear fuel, high-level waste, and other
reprocessed waste returned from French facilities. SYNATOM, an agency privatized in 1994, is
responsible for uranium procurement, reprocessing spent nuclear fuels, off-site waste
management, and disposal of packaged irradiated fuel assemblies. The Nuclear Research Center
(CEN), under the Ministry of Economic Affairs, provides technical assistance in basic and
applied R&D in nuclear energy and technology. ;
Belgium's waste disposal program takes a multi-barrier approach, relying significantly: upon the
low permeability of the Boom Clay formation and the assumption that there are no structural fast
paths through the clay. Engineered barriers are anticipated to last no more than a few thousand
years, after which the geologic barriers will provide primary containment. For this reason,
engineered barriers have been designed to minimize their impact on surrounding geology, and to
provide interim public health protection (NWT94).
ONDRAF intends to begin operation of a shallow land-burial facility for low-level waste by the
year 2000 and has established an underground laboratory in a clay formation at Mol-Dessel to
evaluate the site's suitability as a high-level waste repository. Twenty years of research at Mol-
Dessel has led to significant evolution of the design for the planned repository. The clay
formation in which the site is situated is the only suitable geological medium that has been
identified in Belgium. In 1980, a repository conceptual design was developed for a clay site, and
an underground research laboratory at Mol-Dessel (Project HADES) began operation by 1985.
In recent years, the repository's design has been altered. The original HADES design was
conceived according to the perceived thermal tolerance of the host rock and was intended to
allow for retrieval of containers of vitrified waste for a long period of time. The new design does
not permit easy retrieval and allows for more homogeneous dispersion of heat. The new design
is also believed to be simpler to construct and less damaging to the surrounding clay layer
(NWT95). ;
The underground research laboratory at Mol-Dessel extends to a depth of 224 meters. Research
conducted there includes experiments in corrosion properties of containers and engineered
barriers, geochemistry and radionuclide migration, backfilling and sealing technology, and near-
field effects of heat and radiation on clays. Over the next few years, a preliminary demonstration
test for clay disposal (PARCLAY) will be launched at the Mol-Dessel site. PARCLAY will be
used to investigate the thermal effects of final disposal on clay and will include the construction
of a new 1:1 scale gallery. Based on the outcome of these studies, a larger underground facility
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might be constructed for a full-scale demonstration project. Assuming that the results of
investigations at Mol-Dessel are favorable, repository construction could begin around 2025 and
operation around 2030 (OEC95b).
3.1.3 Regulatory Organizations and Their Regulations
Belgium does not currently have specific regulatory requirements or criteria governing the
disposal of spent nuclear fuel or high-level waste. In 1994, the Federal Nuclear Inspection
Agency (AFCN) was created to oversee inspection and surveillance of Belgium's nuclear
facilities under guidance of the Ministry of Employment and Labor and the Ministry of Public
Health and the Environment. The King of Belgium has the authority to grant, suspend, reject or
withdraw authorization for the construction and operation of nuclear facilities (OEC95c).
3.2 CANADA
3.2.1. Nuclear Power Utilization
In 1994, Canada produced about 19 percent of its electrical needs through nuclear power
(22 pressurized heavy-water cooled and moderated "CANDU" reactors (EIA95)). Canadian
utilities currently produce a surplus of electric power, and only one new nuclear facility is
planned before the year 2005 (OEC95a). Canada's nuclear power is produced by three provincial
utilities, Ontario Hydro11, Hydro Quebec, and New Brunswick Power. As of February 1, 1999
Ontario Hydro had 12 reactors in operation, seven under extended shutdown and one in a laid-up
decommissioning mode. The reactors under extended shutdown may be brought back on line in
the 2000-2009 timeframe depending on economic conditions, The Hydro Quebec reactor and the
New Brunswick Power reactor are both operational (KIN99).
Canada relies on the CANDU reactor design, which uses natural uranium in a once-through fuel
cycle. Currently, the program considers only direct disposal of spent nuclear fuel without
reprocessing, although the reprocessing option has not been completely ruled out. Until a
repository is available, spent nuclear fuel will initially be stored at each reactor site and, later,
possibly at a central facility. Estimates indicate that Canada will have produced about 34,000
11 Ontario Hydro was split into several companies on April 1, 1999 with nuclear power production assigned
to Ontario Power Generation Inc.
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MTHM of spent nuclear fuel by the year 2000. The country's five existing sites (the Ontario
Hydro utility has three sites) have adequate storage facilities for spent nuclear fuel, and there is
little urgency to dispose of waste. Current time lines suggest that a disposal facility could be
established by about 2025 (OEC95c).
Ontario Hydro fuel has been stored at the Pickering Waste Management Facility since 1995
(KIN99). Dry storage utilizes concrete-filled, steel-shelled vessels which contain 384 fuel
bundles each. Currently 600 metric tons of uranium (30,000 fuel bundles) are in storage at
Pickering. Ontario Hydro has filed an application to construct the Bruce Waste Management
Facility with planned operation in 2002, and a similar facility is planned for Darlington in 2005.
New Brunswick Power has operated a dry storage facility at its Pt. Lepreau reactor since 1992
where storage is done in concrete steel-lined vessels each containing 540 fuel bundles. Hydro
Quebec uses a concrete vault for spent fuel storage at its Gentilly-2 reactor site. ;
3.2.2 Disposal Programs and Management Organizations \
Responsibility for the management and disposal of Canada's nuclear waste was allocated in 1978
under the Canadian Nuclear Fuel Waste Management Program. Ontario Hydro (owner of 20 out
of the country's 22 total nuclear power units) assumed the responsibility for interim storage and
transportation of nuclear fuel waste12. The Federal corporation, Atomic Energy of Canada
Limited (AECL), took responsibility for research and development on deep repository disposal,
with support given by Ontario Hydro.
The Canadian disposal concept involves siting a repository at a depth of 500-1000 m in a
granitic formation located in the Canadian Shield, which is a large region of geologically old
rocks that is tectonically quiescent and centered around the Hudson Bay. It stretches east to
Labrador, south to Lake Ontario, and northwest to the Arctic Ocean. The repository would be
located at a depth between 500 and 1,000 meters. Spent nuclear fuel canisters would'be inserted
into floor cavities located in excavated disposal rooms and surrounded with a mix of bentonite
and silica sand. A mix of glacial rock clay and crushed granite aggregate, along with;engineered
barriers, would be used to seal most of the remainder of the vault (OEC93).
12 One additional reactor is owned by Hydro Quebec and one by New Brunswick Power.
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AECL submitted an Environmental Impact Statement (EIS) evaluating the planned disposal
program to the Federal Environmental Assessment Review Panel in 1994. AECL estimated that
siting, licensing, and construction of a disposal facility will take 25 to 30 years and that the
facility could, therefore, be in operation by 2025. The Environmental Assessment Panel reported
in early 1998 that the AECL concept for deep geologic disposal did not enjoy broad public
support and that a number of steps were required to achieve such support (EAP98). Until revised
management and public participation procedures were put in place and broad public support was
obtained, no work to search for a specific site should proceed. In the course of seeking public
support, the AECL repository concept may emerge as the most acceptable approach but that issue
must remain open until other policy issues have been resolved.
In response to the report of the Environmental Assessment Panel, the Canadian government set
forth its position in December 1998 (KIN99). In its response the government asserted that it
expects producers and owners of nuclear fuel to establish, as a separate legal entity, an '
organization to manage and coordinate the full range of activities relating to the management of
the fuel waste including disposal. The producers and owners of the fuel are also expected to
establish a fund to develop and compare waste management options, design and site the preferred
approach for long-term management, implement the preferred approach, and decommission the
waste management facilities. Ontario Hydro has assumed the lead in investigating how best to
establish and separately fund the waste management entity.
In 1986, AECL established the Whiteshell underground research laboratory in undisturbed
granitic rock at a depth of 240 meters at Lac du Bonnet in the Province of Manitoba. The AECL
has since deepened the facility to 440 meters. The purpose of the laboratory is to conduct large-
scale, in-situ experiments in the type of rock envisioned under the Canadian disposal concept,
demonstrating some of the components of the disposal concept (the facility is not a candidate
repository site). AECL is developing methodologies and analytical techniques to evaluate the
geomechanical and geohydrological properties of granitic rock. The underground research
laboratory was also recently used to study the possible effects of microbial activity in a disposal
system. Other studies conducted at the laboratory include large-block radionuclide migration
studies, container corrosion studies, and an alternate post-closure assessment case study.
As the result of a government-wide Federal Program Review begun by Ottawa in 1995, in
response to a large national budget deficit, AECL's responsibilities have been pared back
considerably. The AECL will now focus on a core mission of developing and vending CANDU
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reactor technology. The AECL's Whiteshell facilities, including the Underground Research
Laboratory, may be privatized. In 1997, the Canadian government was in final negotiations to
sell these facilities to Canadian Nuclear Projects Ltd., a consortium led by British Nuclear Fuels,
Ltd. and Wardrup, a large Canadian engineering firm. However, the deal fell through and the
government is looking for other economic opportunities for this site. In the meantime, AECL
continues to operate the Underground Research Laboratory.
3.2.3 Regulatory Organizations and Their Regulations
Regulation of nuclear matters in Canada is handled under the Atomic Energy Control Act. The
Atomic Energy Control Board (AECB) is currently the lead regulatory agency in Canada for
assessing the long-term performance of the disposal facility. The AECB also develops and issues
policy statements and regulatory guidance for the eventual licensing of the high-level waste
repository. Other provincial and Federal agencies operate under AECB in the regulation of some
activities in the nuclear fuel cycle.
In 1987, the AECB issued a policy statement containing objectives, requirements, and guidelines
on nuclear waste disposal and high-level waste repository siting (AECB87). The overall
objective expressed in these documents is to ensure that there is only a small probability that
radiation doses to the public associated with the repository will exceed a small fraction of natural
background radiation doses. Under this policy, predicted radiological risk of death to
individuals from a waste repository must not exceed 1 x 10'6 annually. For the purpose of
demonstrating compliance with the individual risk requirement, the time period need not exceed
10,000 years (AEC87). However, Canada does not have any nuclear-waste-specific regulation at
the present time (KIN99).
Canada is in the process of replacing the Atomic Energy Control Act with the Nuclear Safety and
Control Act (NSCA) and the new law is expected to be in place in mid-1999. Under the terms of
the NSCA, the AECB will be replaced with an expanded Canadian Nuclear Safety Commission
(CNSC). The NSCA will provide the framework under which licenses for site preparation,
construction, operation, decommissioning, and abandonment of nuclear waste facilities are
obtained. One of the tasks allocated to the CNSC is the preparation of regulatory guidance
documents. Any application to build a nuclear waste facility would initiate a requirement to
prepare an environmental assessment under the Canadian Environmental Assessment Act
(KIN99). '
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3.3 FINLAND
3.3.1 Nuclear Power Utilization
Currently, Finland has four nuclear power plants in operation. These facilities are owned and
operated by two separate utilities. Two 445 MWe PWR units, located at Loviisa on the southern
coast of Finland, are run by Imatran Voima Oy (FVO). Two 710 MWe BWR'units, located at
Olkiluoto on the western coast of Finland, are run by Teollisuuden Voima Oy (TVO). The
possibility of a fifth reactor has been under consideration for a number of years in Finland. The
Technical Research Centre of Finland (VTT) also operates a small (250 kW) Triga research
reactor at its Reactor Laboratory (VTT93).
In the past, the two utilities had different waste management strategies. Until recently, IVO has
shipped its spent fuel to Russia with no return of reprocessing wastes to Finland. Spent fuel
generated was allowed to cool for a period of four to five years at the IVO reactor site before
being shipped. However, this practice is no longer allowed. TVO's strategy involves disposing
Of its spent fuel in Finnish bedrock. No plans exist for reprocessing the spent fuel from TVO's
reactors, although this option remains open (VTT93).
In May 1995, TVO and IVO agreed to cooperate for the final disposal of spent fuel and a new
company, Posiva Oy, was established. This company, which is jointly owned by TVO and IVO,
took over TVO's program on spent fuel disposal. The total amount of spent fuel to be disposed
of in Finland is now estimated to be about 1,700 MTHM of BWR fuel from TVO's reactors at
Olkiluoto and 740 MTHM of PWR fuel from IVO's reactors at Loviisa (OEC96).
3.3.2 Disposal Programs and Management Organizations
In 1983, the Finnish government set general targets and schedules for research and development
of a nuclear waste management program. Based on these guidelines, TVO conducted
preliminary site investigations for a spent fuel repository at five sites: Olkilouto in Eurajoki;
Kivetty in Konginkangas (now a part of Aanekoski); Romuvarra in Kuhmo; Syyry in Sievii; and
Veitsivaara in Hyrynsalmi. Technical plans for managing and disposing spent fuel were also
developed. In addition to these activities, extensive research has been conducted on the
phenomena and processes affecting the safety of long term disposal of spent fuel. Several
Finnish research institutes, universities, and companies made R&D contributions, including the
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Technical Research Centre of Finland, the Geological Survey of Finland, and the Department of
Radiochemistry at the University of Helsinki. Finland has also drawn upon the expertise of other
countries, particularly the Swedish Nuclear Fuel and Waste Management Company (SKB)
(YJT92).
The decision, in principle, made by the Finnish government in 1983 required TVO to propose
suitable areas for more detailed site investigations by the end of 1992. As discussed above,
Posiva Oy was established in 1995 and began operation in early 1996. Since its creation, Posiva
Oy has continued investigating the three candidate sites originally selected in 1992: Olkiluoto,
Kivetty, and Romuvaara. hi 1994, TVO also conducted a preliminary feasibility study in
Kannonkoski, near Aanekoski. Another feasibility study has been initiated for the island of
Hastholmen in Loviisa (OEC96).
Finland plans to select the site for the repository in the year 2000 based on updated technical
information on encapsulation and facility design, as well as site-specific safety analyses
(OEC96). Detailed construction plans for the encapsulation facility and repository must be
presented in 2010 and the repository is to go into operation in 2020. Current plans call, for
sealing the repository in the year 2050 (YJT92).
3.3.3 Regulatory Organizations and Their Regulations
Finland's Ministry of Trade and Industry is responsible for nuclear power in the country. The
Finnish Centre of Radiation and Nuclear. Safety is responsible for nuclear safety, including
nuclear waste management. This latter organization reviews technical and safety-related license
applications (VTT93).
The principles of Finland's waste management policy were originally outlined in the
Government's policy decision of 1983. The nuclear energy law of Finland includes specific
directives concerning nuclear waste management. Each nuclear waste producer is responsible for
the safe handling and management of its waste, including final disposal. This responsibility
extends to the financing of such operations. No governmental organizations are envisioned for
waste management operations. The utilities contribute to future waste management activities
through the Nuclear Waste Management Fund established by the Finnish Ministry of Trade and
Industry (VTT93).
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In developing its radiological and safety criteria for nuclear waste disposal systems, Finland is
closely following the international efforts of the International Atomic Energy Agency (IAEA),
the International Commission on Radiological Protection (ICRP), and the Nuclear Energy
. Agency of the Organization for Economic Cooperation and Development. In addition, the
Nordic governments published joint recommendations in 1993 in a document entitled Disposal
of High Level Radioactive Waste, Consideration of Some Basic Criteria (OEC96).
3.4 FRANCE
3.4.1 Nuclear Power Utilization
In 1994, France met approximately 75 percent of its electrical needs through nuclear power,
having the highest per capita installed capacity in the world (OEC95c). The French nuclear
power program relies on 56 units, the vast majority of which are light water reactors. Older gas
cooled reactors are being phased out, while research and development activities and
demonstration projects focus on an alternate reactor designs (liquid metal fast breeder reactor) for
power production. France plans to continue building nuclear power plants, although some will
serve as replacements for old facilities. The overall contribution of nuclear power to the
country's electricity production is not expected to exceed 80 percent (GAO94).
The French radioactive waste disposal program is based on a closed fuel cycle involving spent
nuclear fuel reprocessing and recovery and re-use of plutonium in breeder and light water
reactors. From 1976 through 1990, France reprocessed over 20,000 MTHM of metallic and
oxide fuel. France has already begun to solidify high-level waste in glass and, ultimately, intends
to dispose of it—as well as alpha-emitting transuranic waste—in deep geological formations.
Vitrification plants for France's two reprocessing plants, UP2 and UP3, entered service in 1990
and 1992. France also provides reprocessing services to foreign customers; in 1993, the
international component comprised an estimated one-third of the country's reprocessing business.
3.4.2 Disposal Programs and Management Organizations
The French nuclear waste program has been entrusted to the National Radioactive Waste
Management Agency (ANDRA). ANDRA was formed in 1979 as an arm of the French Atomic
Energy Commission (CEA), but 1991 legislation made it an independent entity. ANDRA is
responsible for all radioactive waste disposal activities and long-term waste management. Other
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organizations with key roles in the management of the country's high-level waste include
Electricite de France (the national electric utility) and COGEMA (operator of spent nuclear fuel
reprocessing and high-level waste immobilization and storage facilities).
In 1987, ANDRA identified four geological media for potential high-level waste disposal— clay,
salt, granite, and schist—and began investigative work at a site in each medium. An
underground research laboratory was to be established at one or more of the candidate sites; if
found suitable, and one of these sites was to have been converted to an operating repository to
receive transuranic waste by 2000 and high-level waste by 2010. However, in light of the serious
public protests at three of the sites under investigation, former Prime Minister Michel Rocard
declared a one-year moratorium on siting activities to allow a reassessment of France's overall
waste management strategy. The moratorium began in February of 1990, and, in January 1991,
the Parliamentary Office for the Assessment of Technological Options published a report that
recommended major changes to the program.
On December 30, 1991, the Parliament enacted a new Law on Radioactive Wastes. The 1991
law requires the government to submit a report to Parliament after 15 years that assesses the
results of studies on partitioning and transmutation of actinides, the retrievable or permanent
storage of high-level waste in deep geologic formations including the use of underground
laboratories, and the technologies for waste conditioning and surface storage. (Work on the first
and third options is coordinated by CEA while work on the second option is the coordinated by
ANDRA.) The report must also propose a bill authorizing an underground waste repository. At
this time, no schedule has been set for developing such a repository. Instead, the Parliament will
reassess the program based on the results of the 15-year research phase. The law states that, once
the underground research laboratory is built, only research-level quantities of waste m^y be
emplaced into it until the Parliament votes to convert the laboratory into a repository. While no
direct disposal of spent nuclear fuel is envisioned, the law also requires that the government
perform research on direct spent nuclear fuel disposal options. Annual progress reports and the
final report on the three technological areas are to be prepared by a National Evaluation
Commission composed of eminent scientists (JOR99).
The new law allowed the government to resume site selection efforts for underground research
laboratories. A waste "negotiator" or "mediator" was appointed to discuss proposed
investigations with local and regional officials. About 30 localities subsequently expressed
interest in hosting a laboratory. In 1994, the government's Bureau of Geological and Mineral
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Research (BRGM) investigated these regions and eliminated those with adverse geology. In
early 1994, the negotiator announced the selection of four new regions as candidates to host a
repository: (1) the southern region of the Vienne "departement," in west-central France; (2) the
area surrounding Marcoule in the Gard departement; (3) the Meuse departement, bordering
south-eastern Belgium; and (4) the northern Haute Marne departement, north of Dijon. Two
other localities were selected as secondary choices because their local governments had not voted
on their candidacy; the four primary localities all voted in favor of their candidacy.
The Meuse and the Haute Mame sites were subsequently merged and designated as the East site.
Since two years of geophysical examination and drilling revealed no prohibitive factors at any of
the sites, ANDRA proposed in 1997 to proceed with underground laboratories at each location.
Whereas ANDRA had previously considered four types of potential host rock, the selected
regions represent only two: clay and granite. The National Evaluation Commission supported
selection of the East site and the Gard site which are in clay but noted confinement problems
with the granitic host rock at the Vienne site (JOR99). On December 9, 1998 the French
government authorized construction of underground laboratories at the East site in clay and at a
new site in a granitic formation to be located by ANDRA.
The 1991 law includes additional provisions designed to ease public concern about France's high-
level waste management program, including the creation of a policy of openness concerning the
country's high-level waste disposal program and a requirement that government grants and jobs
for the host municipality accompany the underground research laboratories.
3.4.3 Regulatory Organizations and Their Regulations
Agencies with regulatory responsibilities include the Directorate for the Safety of Nuclear
Installations (DSIN) within the Ministry of Industry; the CEA and its subsidiary, the Institute for
Nuclear Protection and Safety (IPSN); the BRGM; and SGN (architect and engineering services).
DSIN, France's principal nuclear regulatory authority, issued Fundamental Safety Rule III.2.f.
(DSI91) pertinent to high-level and alpha waste disposal, on June 10, 1991. The rule requires
that:
The impact of a deep geologic disposal facility on radiation exposures be as low
as reasonably achievable
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• Individual dose equivalent due to the facility be limited to 0.25 mSv (25 mrem)
per year for likely events
The stability of geologic barriers be demonstrated for at least 10,000 years
• High-level waste packages prevent the release of radioactive contents during the
period when short- and medium-lived radionuclides dominate total radioactivity
In preparation for the underground laboratory phase, the EPSN, within CEA, is independently
preparing facilities to evaluate the long-term safety requirements of a repository on behalf of the
regulatory authority DSIN. i
3.5 GERMANY
3.5.1 Nuclear Power Utilization
In 1994, Germany met about 30 percent of its electrical needs through nuclear power (EIA95).
The German nuclear power program relies primarily on pressurized light water reactors (14 units)
and boiling water reactors (7 units), although research and development activities and
demonstration projects are also evaluating alternate reactor designs (high temperature gas-cooled
reactors'and liquid metal fast breeder reactors) for power production. It is estimated that by the
year 2000, Germany will have generated about 9,000 MTHM of spent nuclear fuel. Germany has
historically planned to dispose of spent nuclear fuel in deep geological formations only after
reprocessing, as stipulated in a 1976 amendment to Germany's Atomic Energy Law. Plans for a
domestic reprocessing facility were abandoned in 1989 and German utilities chose instead to ship
their spent nuclear fuel to France and Britain for reprocessing. Resulting vitrified waste is
currently returned to Germany and stored in metal casks for planned subsequent disposal. A
1994 amendment to its Atomic Energy Law, however, legalized the direct disposal of spent
nuclear fuel elements as well (Atomic Energy Law, Article 4, amendment of section 9ja(l))
(GER94). Since then, German utilities have been considering both management options.
3.5.2 Disposal Programs and Management Organizations
The German government's Institute for Radiation Protection (BfS) is responsible for the design,
construction and operation of waste disposal facilities. Vitrified high-level waste returned from
foreign reprocessors was targeted for disposal at the Gorleben facility, a salt dome located in
Lower Saxony, if the site proves acceptable. Spent nuclear fuel would also be directly disposed
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at Gorleben. The newly legalized option of direct disposal has required modification of plans at
the facility. Until a repository is in operation, vitrified waste will be stored at Gorleben and the
Ahaus facility in Northrhine-Westphalia. Expansions at both the Gorleben and Ahaus storage
facilities are currently proposed. A former salt mine at Asse, which served until 1978 as a
repository for low-level (125,000 containers) and intermediate-level (1,300 drums) radioactive
wastes, now serves as an underground research laboratory for high-level waste disposal.
The Gorleben salt dome ranges in depth from 250 meters to 3000 meters. Construction of an
underground research laboratory was initiated in 1986, but all work was stopped for over a year
in 1987 because of a construction fatality. As of 1995, two shafts had been sunk to depths of 600
and 620 meters (emplacement at approximately 870 meters is anticipated (LOM95)). Current
areas of emphasis include hydrogeological investigation and seismic measurements (OEC95b).
Construction of the repository could start at the turn of the century, and the facility is scheduled
to remain operational for about 50 years. It is anticipated that the site will receive about
550 MTHM of vitrified waste, 200 metric tons of directly disposed spent nuclear fuel, and about
6,690 containers of low-level and intermediate-level waste per year (LOM95).
Repository design emphasizes the role of the surrounding geology as a barrier. It is anticipated
that the salt dome's formations will creep over time in response to radiogenic heating and
pressure gradients to encapsulate the waste. The use of steel and iron canisters is intended
primarily to contain waste in the short-term. The possibility of direct spent nuclear fuel disposal
has required additional research and design development.
Federal agencies have been generally positive about the suitability of Gorleben as a repository
site. However, while research appears to generally support the suitability of the Gorleben site,
the project has faced increasing opposition from the government of Lower Saxony. As a
precautionary measure, in 1995 the Federal Institute for Geosciences and Natural Resources
prepared two reports identifying other potential sites for a HLW repository in the event that the
Gorleben site is found unsuitable (OEC96).
Subsequently, Federal elections in 1998 dramatically altered the direction of German nuclear
program. A coalition agreement signed by the Greens and the Social Democrats on October
20,1998 stated that use of nuclear energy would be phased out (BRE99). With regard to waste
management, the agreement noted that a single repository in a deep geologic formation is
sufficient for all types of radioactive waste and the time-dependent target for HLW disposal is
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2030. The agreement further stated that the suitability of the Gorleben salt dome is questionable
and work there should be interrupted while potential sites in other host rocks are examined.
Emplacement of radioactive wastes at Morsleben (in former East Germany) is to be terminated
and the site decommissioned.
3.5.3 Regulatory Organizations and Their Regulations >
Key German agencies include the Federal Ministry for Environment, Protection of Nature and
Reactor Safety (BMU), the Federal Ministry for Research and Technology (BMFT), the BfS, the
Federal Institute for Geosciences and National Resources, and the host state's ministry for
environmental protection. As the primary federal supervisory authority, BMU receives advice
from two committees of independent experts, the Reactor Safety Commission (RSK) and the
Committee on Radiological Protection (SSK).
In Germany, the institutional and legal framework for the regulation of nuclear facilities is based
on the joint participation of Federal and state governments. State governments serve as licensing
authorities for all nuclear waste facilities, although the Federal government has the authority to
override these decisions. The Federal government retains primary responsibility for waste
disposal; the Atomic Energy Law and the Radiation Protection Ordinance (GER94) establish the
principles and requirements regarding the safe utilization and application of atomic energy and
radioactive materials, including the disposal of radioactive waste. Under the Radiation
Protection Ordinance, dosage limits are set at 0.3 mSv (30 mrem) per year for "all reasonable
scenarios" (OEC95a).
German regulators had been developing safety regulations that the Gorleben facility would be
required to meet through a site-specific safety assessment. It is expected that this safety
assessment will be required to demonstrate that potential exposure to radiation from disposed
waste will be kept within the range of natural radiation for a period of about 10,000 years and
that integrity of the repository system will be maintained over a longer period of time (GAO94).
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3.6 JAPAN
3.6.1 Nuclear Power Utilization
In 1994, Japan produced about 31 percent of its electrical needs through nuclear power provided
by 49 reactors (EIA95). It is anticipated that this figure will increase to approximately 33 percent
by the year 2000 and to about 42 percent by the year 2010 (AEC94). The Japanese nuclear
power program currently relies primarily upon light water reactors, although research and
development activities and demonstration projects are also evaluating alternate reactor designs
(gas cooled reactor, heavy-water moderated reactor, and liquid-metal fast-breeder reactor). By
1994, the country's first prototype fast breeder reactor, Monju, had reached criticality (OEC95b),
but a December 1995 coolant leak dealt a setback to the project.
Japan's spent nuclear fuel is currently reprocessed in France and England. However, both
countries have exercised their option to return vitrified residue to Japan; the first return delivery
from .France took place in February 1995. Domestically, the Power Reactor and Nuclear Fuel
Development Corporation (PNC) has operated a small reprocessing plant since 1977, where
roughly 720 tons of spent nuclear fuel had been reprocessed as of .1993. Furthermore, at the
Rokkasho site in Aomori prefecture, a private utility consortium, Japan Nuclear Fuel Services
Limited (JNFL), plans to begin operating a large commercial-scale plant shortly after the year
2000 (AEC94). It is estimated that by the year 2000, Japan will have discharged about
20,000 MTHM of spent nuclear fuel from its reactors. Vitrified high-level waste will be stored
30 to 50 years for cooling before disposal in a geologic repository.
3.6.2 Disposal Programs and Management Organizations
As noted above, Japan's current waste management strategy includes spent nuclear fuel
reprocessing using domestic and foreign facilities, on-site spent nuclear fuel storage, waste
solidification followed by long-term storage, and eventual disposal in a suitable deep geological
formation. Japanese nuclear utilities are responsible for storing high-level waste and funding its
disposal; JNFL is responsible for low-level waste disposal activities at Rokkasho (OEC95c).
Two government-sponsored organizations—PNC13 and the Japan Atomic Energy Research
Institute (JAERI)—are responsible for research and development addressing the fuel cycle, waste
13 On October 1, 1998 PNC was succeeded by the Japan Nuclear Cycle Development Institute (JNC).
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management, and disposal. In 1993, the Steering Committee on High-Level Radioactive Waste
(SHP) was created to spearhead planning for disposal of the country's high-level waste,
Radioactive waste is managed in accordance with Japan's Long Term Program for the
Development and Utilization of Nuclear Energy (AEC94). In 1994, the Atomic Energy
Commission (AEC) issued an update to the long-term disposal plan, placing particular emphasis
on the disposal of high-level waste and adding new details to the country's plans and timetables
for this effort. The 1994 update established a procedure for implementation of a deep geologic
repository and provided guidelines on storage, vitrification, and geologic disposal. The plan also
added clarity to the roles of Japan's nuclear-related organizations, which can be summarized as
follows:
• Government Research and Development Organizations (PNC and JAERI): PNC
is the lead organization implementing the research and development program in
various areas of the fuel cycle and geologic disposal, while JAERI performs
research in support of the government's safety evaluation of geological disposal,
as well as research on advanced waste management technologies.
Utilities and their Consortia: Utilities are responsible for funding high-level waste
disposal programs and for contributing to related research and development work.
• Government Agencies: Government agencies are responsible for oversight and
overall coordination of disposal. While it has not yet been decided what entity
will implement or license the disposal project, the AEC's 1994 update of the
country's long-term disposal plan suggests that this duty will be delegated by the
year 2000 and that SHP is responsible for studying the matter (SHP has not been
designated the implementing entity).
t
The 1994 plan also laid out a five-step process to develop a high-level waste repository. The first
phase, selection of effective formations, was completed in 1984. Subsequent steps as established
by the AEC include: (1) establishment of an implementing organization by the year 2000;
(2) selection of candidate disposal sites, subject to government cooperation and community
acceptance; (3) demonstration of disposal technology at the candidate site, followed by license
application; and (4) establishment of necessary laws and policies for the disposal implementation
and safety. The plan called for the repository to be operational by 2030 to 2045. Japanese
authorities have determined that high-level waste disposal should be possible in any geologic
formation excluding unconsolidated media (e.g., soil and sand). Because of geological
heterogeneities in Japan, geological characterization is expected to be difficult, causing
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uncertainties in predicting the performance of natural barriers. Thus, Japan is assigning a major
role to the engineered barrier system, while defining a small number of critical natural
characteristics for the site which are expected to be achievable in various geological settings.
In April 1997 the AEC issued Guidelines on Research and Development Relating to Geological
Disposal ofHigh-Level Radioactive Waste in Japan. The Guidelines describe the level of
technical reliability which must be demonstrated for a geologic repository. The Guidelines also
identify issues which must be addressed to establish the technical basis for selection of potential
disposal sites and for the formulation of safety standards. R&D activities to be conducted after
2000 are also identified (MAS99).
Considerable research and development has been underway in recent years. Most research is
conducted by PNC (now JNC) and regular plans are submitted to AEC; the most recent progress
report (designated H3) was submitted in 1992, and a subsequent report (tentatively designated
HI 2) is expected before the year 2000 (OEC95b)(MAS99). The H12 report (a draft of which
was issued in 1998) is expected to more rigorously address the feasibility of the specified
disposal concept than did the H3 Report. The H12 report is also expected to provide input for
regulatory and siting processes which will be set in motion after 2000 (MAS99).
PNC operates an underground test facility in both sedimentary and crystalline rock environments.
The test facility is located in the Tono Uranium Mine in central Japan. Major experiments in the
mine include ground water flow investigation; studies on the effects of excavation on the
mechanical and hydraulic behavior of the repository; natural analogue studies and evaluations of
the chemical durability of simulated waste glasses; and the corrosion rates of candidate-overpack
materials. Since 1988, PNC has also conducted major tests in the Kamaishi iron ore mine in
northern Honshu. Work at this mine is currently guided by a 5-year research plan, submitted by
PNC in 1993 and characterized by work in a deeper gallery. Major investigations at Kamaishi
have included detailed fracture mapping, cross-hole hydraulic and geophysical testing, drift
excavation-effect studies, in-situ stress measurements, single-fracture flow tests, and
observations of seismic activity. In December 1995, PNC signed an agreement with local
governments to build an underground research laboratory near the city of Mizunami. The
laboratory will be used to research ground water and rock mass characteristics for geologic
disposal (OEC96).
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Several new facilities were started in recent years, including: (1) the Tokai Vitrification Facility
(the first vitrification facility in the country, where operation began in 1995); (2) the Nuclear Fuel
Cycle Engineering Facility (NUCEF), where construction was completed in 1995; (3) the
Recycling Equipment Testing Facility (RETF), begun in 1995 to develop reprocessing techniques
for spent nuclear fuel from fast breeder reactors (OEC95b); and (4) the ENTRY facility ;at the
Tokai Research Lab to conduct full-scale engineering tests and non-radioactive simulations of the
performance of natural and engineered barriers.
3.6.3 Regulatory Organizations and Their Regulations ',
The Atomic Energy Basic Law of 1955 established the AEC and the principles and requirements
for the safe utilization and application of atomic energy and radioactive materials, including the
disposal of radioactive waste. In addition to the AEC, other key agencies or organizations
include the Nuclear Safety Commission (NSC), the Ministry of International Trade and Industry
(MITI), and the Science and Technology Agency (STA). Regulatory requirements for the high-
level waste repository have not yet been established, nor have formal individual dose limits been
issued.
3.7 SPAIN ;
3.7.1 Nuclear Power Utilization
As of December 31, 1990, Spain had a total of nine light water nuclear power plants in operation.
A tenth reactor, the graphite-gas Vandellos plant, was expected to begin decommissioning in
1996. Spent fuel from the light water reactors is stored on site in pools specifically designed for
this purpose. Spent fuel from the Vandellos 1 plant was sent to France for reprocessing: Spain's
radioactive waste is currently managed by the Empresa Nacional de Residues Radioactivos, S.A.
(The Spanish National Radioactive Waste Company - ENRESA) which was established by Royal
Decree 1522 on July 4, 1984. Eighty percent of the company is owned by the Spanish Centre for
Energy, Environmental and Technological Research (CIEMAT) and 20 percent is owned by the
National Institute for Industry (INI) (SMI91). ;
!
Two types of high-level wastes will have to be managed in Spain: spent fuel from light water
nuclear power plants and the vitrified wastes from reprocessed fuel from the Vandellos 1 plant.
At the end of 1990, 974 MTHM of spent fuel were stored at the sites of Spain's nine nuclear
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power plants. It is projected that approximately 5,200 MTHM of spent fuel will ultimately be
managed in Spain. In addition, 180 m3 of vitrified wastes from the Vandellos 1 plant will require
final disposal (SMI91).
3.7.2 Disposal Programs and Management Organizations
ENRJESA manages all radioactive wastes in Spain, including low-level, intermediate level, and
high-level wastes. The high-level waste program in Spain considers both intermediate storage
and final disposal of these wastes. Intermediate storage, which allows the radioactive elements
of the waste to cool down and decay, includes both dry storage in casks and vaults and liquid
storage in pools. Waste has been stored on-site, although Spain has contemplated a Temporary
Centralized Storage option (SMI91).
The goals and objectives of Spain's program to permanently dispose of high-level waste and
spent fuel were outlined in its Third General Radioactive Waste Plan. This strategy, which was
defined and initiated in 1987, included three areas of work:
Search for a site for facility construction—Spain considered granites, salts, and
clay media
Acquisition of technology and training of teams required for characterization of
the chosen site and construction of the disposal facility
Development of the basic design for a deep geological disposal facility (SMI91)
Progress is being made in terms of defining the conceptual design of a geological repository in
Spain. The development of a preliminary conceptual design for granite and clay was completed
in 1992 and for clay in 1994. A non-site specific conceptual design for salt has also been
completed. More recently, a probabilistic performance assessment in a generic granite formation
was done in 1997 and a similar study in a generic clay formation was done in 1998 (SAN99).
Spain has also adopted its Third Research and Development Plan which covers the period from
1995 through 1999. This plan, which includes all types of radioactive wastes, has as its main
objective the support required for the performance of the high-level waste program. The Plan
primarily emphasizes verification of site characterization methodologies and preliminary
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repository designs, application of numerical models, and acquisition of specific data for the
performance of long-term safety assessments (OEC96). ;
Public pressure caused a cessation of field work in 1996 and, in response, the Spanish Senate
created an Inquiry Committee to provide recommendations to the government on how to develop
a radioactive waste management policy (SAN99). The Committee noted that, while deep
geological disposal is the basis for most international programs, there is growing interest in
partitioning and transmutation of long-lived radionuclides. They stated that decisions needed to
have a high and broad level of socio-political consensus which could be obtained only through
involvement of a wide range of institutions and administrations. Subsequently, the Spanish
government provided further guidance on waste management policy in early 1998 which
included the following:
• No decision on disposal of high-level waste will be made before 2010
No further siting activities will be undertaken before 2010 (After that date, siting
activities must be voluntary)
• Deep disposal will continue to be studied but other technologies such as
partitioning and transmutation should also be analyzed
ENRESA is modifying its strategy to be congruent with the government guidance and the
changes will be reflected a new R&D Plan (1999-2003).
3.7.3 Regulatory Organizations and Their Regulations
The Spanish Nuclear Safety Council (Consejo de Seguridad Nuclear) has officially adopted a
level of individual risk below 10'6 per year for the long term disposal of radioactive wastes. This
risk value equates to a dose to individuals in the critical group of less than 0.1 mSv/year (10
mrem/year)(SMI9.1).
3.8 SWEDEN I
3.8.1 Nuclear Power Utilization |
Following a 1980 national referendum, the Swedish Parliament decided to phase out nuclear
power plants by the year 2010. Although the Swedish government maintains this commitment,
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the country remains dependent on nuclear fuel for approximately 51 percent of its electrical
power needs (as of 1994). Sweden's nuclear power is produced with nine boiling water reactors
and three pressurized water reactors (EIA95).
By 2010, Sweden will have produced nearly 8,000 MTHM of spent nuclear fuel. In the 1970s,
Swedish utilities had entered into agreements with other countries to reprocess foreign sources of
spent nuclear fuel; however, this approach was abandoned following the 1980 referendum and
the utilities have since sold their reprocessing contracts or traded high-level waste from
reprocessing for other spent nuclear fuel. A joint utility consortium, the Swedish Nuclear Fuel
and Waste Management Company (SKB), manages the disposal of radioactive waste. In 1985,
SKB began operating a centralized spent nuclear fuel storage facility (CLAB) that will eventually
hold all of Sweden's spent nuclear fuel for about 40 years. As of 1994, this facility was filled to
about 45 percent capacity (SKB94). The facility is situated in an underground granite cavern at a
depth of 30 meters, near an existing nuclear power plant (Oskarshamn). In 1998 the Swedish
government authorized expansion of CLAB and the work is expected to be completed in 2004.
3.8.2 Disposal Programs and Management Organizations
Nuclear waste management activities in Sweden are guided by the Act Concerning Nuclear
Activities and the Act Concerning the Management-of Natural Resources. Every three years,
SKB is required to provide Swedish regulators with a research and development plan for
activities related to the management and disposal of the country s radioactive waste. The 1992
plan was approved contingent upon additional details regarding deposition and canister design
(SKB94).
As outlined in the 1993 plan, Sweden's reference disposal concept for spent nuclear fuel is to
encapsulate it in high-integrity copper canisters and emplace the canisters in a repository built in
crystalline rock at a depth of about 500 meters, backfilling the deposition holes with highly-
compacted bentonite and the tunnels and shafts with a mixture of sand and bentonite. SKB is
evaluating alternative concepts such as deep boreholes and tunnel emplacement, as well as
alternative canister designs. Canisters are expected to consist of a steel insert (for mechanical
protection) inside of a copper sleeve (for corrosion protection).
SKB's 1995 R&D Programme (SKB95) set forth the following schedule for establishing a deep
geologic repository:
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8KB is currently conducting feasibility studies, planned at a total of five to ten
municipalities
Feasibility studies are expected to be completed by 1997, after which time two
municipalities, and locations within both, will be selected for site investigation
One site will be selected around 2001, and deposition of encapsulated fuel is
planned for 2008, when a small portion (approximately 800 tons) of Sweden's
nuclear fuel will be deposited
Feasibility studies were completed in Storumann in May, 1995, but a local referendum rejected
further research there. A feasibility study at a second municipality, Malaa, was completed in
March of 1996'. The municipality of Malaa is conducting an independent review of the feasibility
study involving local stakeholders and independent experts, and the decision on whether to
organize a local referendum on further research will be based on the results of this review
(SKB96). Other sites where feasibility studies are underway or being considered include four of
the five municipalities with existing nuclear facilities (Varberg, Oskarshamn, Nykoping, and
Osthammer). The fifth site with an existing nuclear facility, Kavlinge, is currently not
considered a candidate. <
However, site selection has not progressed as rapidly as envisioned in the 1995 R&D :
Programme. The 1998 RD&D Programme calls for feasibility studies at five sites to be
completed and two sites to be selected by 8KB for site investigations by 2001 (HED99).
The OECD/NEA conducted an international research project in an underground research
laboratory at Sweden's Stripa mine from 1980 to 1991 (OEC95b). 8KB has recently completed
construction of a second laboratory under the island of Aspo, 2 km north of Oskarshamn, at a
depth of 450 meters. The Aspo site will be used to test methods of site selection and •
characterization, and to research disposal technologies for later use in Sweden's deep geologic
repository.
3.8.3 Regulatory Organizations and Their Regulations
Key government entities with direct responsibilities in waste management include the Swedish
Nuclear Power Inspectorate (SKI), the National Institute for Radiation Protection (SSI), and the
Swedish Consultative Committee for Nuclear Waste Management (KAS AM). All operate under
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the supervision of the Ministry of the Environment and Natural Resources. The National Board
for Spent Nuclear Fuel (SKN), a former public entity with regulatory responsibilities, was
absorbed into SKI in the early 1990s.
As of 1995, safety requirements for management and disposal of high-level waste were in
development. These requirements are the responsibility of SSI, in cooperation with SKI. Both
SSI and SKI favor a total systems approach, without specifying detailed sub-system quantitative
criteria in early phases of repository development. Proposed guidelines for the deep repository
would require that:
Radiation doses to individuals be limited to 0.1 mSv/yr for a reasonably
predictable period of time (one million years), after which radionuclide fluxes are
to be limited to a level corresponding to naturally occurring fluxes of
radionuclides
A passive multi-barrier approach be used
Future safety of the facility requires no further controls after the facility is sealed
The repository be designed to not restrict future attempts to change the repository
or retrieve the waste (SKB95)
Established general principles for the management of nuclear waste state that:
Radiation protection take into consideration issues of biodiversity and natural
resource use in addition to human health
Radiation protection be independent of whether doses arise today or in the future,
or whether they originate within or outside the country
The disposal of nuclear waste pose a risk no greater than that of other portions of
the nuclear fuel cycle
All activities must be justified, protection must be optimized, and the individual
must be protected by dose limits (SKB95)
3.9 SWITZERLAND
3.9.1 Nuclear Power Utilization
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In 1994, Switzerland's five nuclear power plants supplied about 37 percent of the country's
electrical power needs (EIA95). The Swiss nuclear power program relies on a mix of pressurized
and boiling light water reactors (three PWRs and two BWRs). Although there is currently a
.moratorium on construction of new nuclear plants, capacity increases at existing plants have kept
supply high, and a 10 percent increase by the year 2000 is planned (OEC95c). ;
The Swiss estimate that, by the year 2000, they will have produced about 1,800 MTHM of spent
nuclear fuel. Switzerland currently ships its spent nuclear fuel to France and Britain for
reprocessing but maintains the options of spent nuclear fuel management both with and without
reprocessing in the future.
3.9.2 Disposal Programs and Management Organizations
The responsibility for establishing radioactive waste disposal facilities in Switzerland lies with
the National Cooperative for the Storage of Radioactive Waste (NAGRA), a joint government
and utility cooperative agency. NAGRA was established in 1972 to manage the disposal of
radioactive wastes, including spent nuclear fuel, high-level waste and other reprocessed waste
returned from the French and British reprocessing facilities. Waste conditioning and interim
storage of reprocessed waste, high-level waste and spent nuclear fuel is the responsibility of
ZWILAG, a cooperative comprised of nuclear utility operators. ;
Overall nuclear policy is governed by the Swiss Atomic Law, to which two major changes were
proposed in 1994. These proposed revisions were subsequently dropped by the parliament in
favor of drafting an entirely new national nuclear energy law. A draft of this law is in
development (OEC96). The overall goal of the Swiss program is to establish the viability of a
repository in Switzerland by the year 2000, although commissioning of a repository will not
occur before 2020 to allow a 40-year spent nuclear fuel/high-level waste cooling period.
Participation in any international repository projects that may develop is also under '
consideration.
The Swiss have historically considered two rock types, crystalline rock and sedimentary rock, as
potential host media for a high-level waste repository. In 1984, NAGRA launched studies in
crystalline rock by drilling seven deep boreholes into the crystalline basement of northern
Switzerland and, subsequently, conducted geological and safety assessment studies. In 1994,
NAGRA released a synthesis of this research (Kristallin I), expressing optimism about the use of
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crystalline rock as a host rock; specifically, NAGRA is considering crystalline rock formations in
northern Switzerland as viable sites for a repository and is planning additional field work there
(OEC95b). In support of crystalline rock studies, a new three-year Phase IV study was launched
at the Grimsel Rock Laboratory in 1994. The Swiss have also considered two sedimentary rock
types, Opalinus clay and freshwater molasse, and have conducted field research on both
formations. In January 1994, the Safety Inspectorate, NAGRA, and representatives of the
relevant government agencies identified Opalinus clay as the preferred sedimentary host rock
option (OEC95b). A site in Benten, just north of Zurich, has been identified for seismic survey
and the construction of an 800-meter borehole to further examine the feasibility of a repository in
clay.
By the year 2000, NAGRA must submit a program—the Siting Feasibility Project—for
government approval that demonstrates the feasibility of siting a repository in one or more of the
crystalline or sedimentary media under consideration.
3.9.3 Regulatory Organizations and Their Regulations
Key organizations or agencies with direct regulatory responsibilities in waste management
include: the Nuclear Safety Division (HSK) of the Federal Energy Office (BEW) within the
Federal Department of Transport, Communications, and Energy (EVED); the Federal
Commission for the Safety of Nuclear Installations (KSA); the Federal Department of Interior
(EDI); and the Institute for Reactor Research (EIR). An interagency working group (AGNEB)
was also established to coordinate activities in support of government decisions on the licensing
of nuclear waste facilities.
In November 1993, HSK released the current guidelines for management of nuclear waste in the
country, entitled Radiation Protection for the Disposal of Nuclear Waste (HSK93). Dosage is
limited to 0.10 mSv (10 mrem) per year for reasonably probable scenarios, and annual risk is
limited to 10"6 for unlikely scenarios. Candidate repositories must produce a system capable of
meeting these requirements in order for their application to be considered (GAO94); furthermore,
all repositories must be designed to be sealed at any time within a few years, after which it must
be possible to dispense with institutional controls.
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3.10 UNITED KINGDOM
3.10.1 Nuclear Power Utilization
In 1994, the United Kingdom (UK) met about 26 percent of its electrical needs through nuclear
power (ELA95). During the 1960s and 1970s, the UK depended primarily on a series of Magnox
(magnesium-clad, uranium metal-fueled) reactors (20 units in operation as of 1995), but began to
use advanced gas-cooled reactors (AGRs) during the 1970s and 1980s (14 units in operation as of
1995.). One pressurized water reactor (PWR) was commissioned in 1995. Use of a fast reactor
was explored as well, but a prototype facility in Dounreay was closed in 1994. ;
F
British Nuclear Fuels, pic. (BNFL), a government-owned corporation, reprocesses spent nuclear
fuel at its Sellafield facility on behalf of both domestic and foreign utilities. Spent metallic fuel
from the country's Magnox reactors is reprocessed at the Sellafield facility at a rate of
approximately 400 cubic meters annually (NIR94). In March 1994, BNFL began operating the
Thermal Oxide Reprocessing Plant (THORP) at Sellafield to reprocess spent nuclear fuel
produced by the country's AGR and PWR reactors and by international customers. THORP is
the country's first commercial-scale reprocessing plant for oxide fuels.
In 1994, the Board of Tirade and the Secretary of State for the Environment placed portions of the
British nuclear program under review. In May 1995, at the conclusion of the review, it was
announced that the country's comparatively modern facilities (7 AGR stations and the Sizewell
PWR) and aH future facilities were to be privatized (nuclear power had been excluded from the
1990 privatization of the electric utility industry). Under privatization, all AGR and PWR
stations have been grouped under a new company called British Energy pic. Magnox stations
have been transferred to a new company called Magnox Electric pic (Magco), a government-
owned company responsible for operation of and liabilities resulting from these stations. It is
expected that Magco will ultimately become a subsidiary of BNFL. ;
Since 1952, over 30,000 MTHM of metal fuel from the Magnox reactors have been reprocessed
in the UK. It is estimated that by the year 2000, Britain will have about 4,000 cubic meters of
high-level waste destined for storage or disposal due to the reprocessing of some 60,000 metric
I
tons of spent nuclear fuel. High-level waste is currently stored in an air-cooled facility at
Sellafield. \
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3.10.2 Disposal Programs and Management Organizations
The responsibility for managing the storage and disposal of radioactive waste lies with its
producers. BNFL has the lead responsibility for management of high-level waste from
reprocessing, hi 1990, BNFL began operating a vitrification plant at Sellafield. In 1982, the
government established the Nuclear Industry Radioactive Waste Executive (NIREX) to develop
and operate intermediate- and low-level radioactive waste disposal facilities. NIREX was
originally established as a partnership consisting of private firms and governmental agencies. In
1985, NIREX was restructured as an independent legal entity, UK NIREX Ltd.
Historically, the UK's radioactive waste disposal strategy has postponed the development of a
high-level waste disposal facility, considering deep disposal of low- and intermediate-level
wastes a higher priority. NIREX had been researching a potential disposal site near Sellafield
and plans suggested that a repository for low- and intermediate-level wastes could be operational
there by the year 2010. A recent governmental review, however, has comprehensively rejected a
proposed environmental impact study by NIREX that was necessary to proceed with construction
of an underground research laboratory at Sellafield. This rejection has caused NIREX to refocus
its program onto more generic issues while continuing to condition and package intermediate-
level wastes for eventual disposal (HOL99).
Eventual deep disposal is planned for high-level waste.. Current plans call for continued
reprocessing of spent nuclear fuel, solidification of high-level waste, and surface storage for
about 50 years. Under this schedule, the need for a high-level waste repository is not expected
before the year 2040. Vitrified high-level waste would then be disposed in deep geologic media.
The UK Department of Environment, Transport and the Regions instituted, in 1997, a two-year
review of options for management and eventual disposal of spent fuel. The study, which is
nearing completion, is designed to formulate a program for development of a deep repository for
HLW and to define the key elements of the required R&D work.
The UK has also adopted a policy of monitoring the results of research activities being conducted
by other countries. Depending on the outcome of research being conducted abroad, Britain
would then develop a high-level waste disposal and repository strategy using concepts that best
fit British needs.
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3.10.3 Regulatory Organizations and Their Regulations
The Atomic Energy Act of 1946 establishes the authority and responsibility to control and
regulate the development of nuclear power in Britain. The Act has been amended several times
to establish new requirements, including those addressing the management and disposal of
radioactive waste.
The regulatory functions are performed by the Nuclear Installations Inspectorate, which is part of
the Health and Safety Executive; the Radiochemical Inspectorate of the Department of the
Environment; the Ministry of Agriculture, Fisheries, and Food; the UK Atomic Energy i
Authority; and the Secretaries of State of Scotland and Wales. The government also takes advice
from several independent experts and advisory committees, including the Radioactive Waste
Management Advisory Committee and the National Radiological Review Board.
!
Exposure limits are based on recommendations of the National Radiological Protection Board.
While no current limits pertaining to exposure from spent nuclear fuel and high-level waste have
been set, indications are that they will be similar to those set for low- and intermediate-level
wastes (10-6 per year for individual risk from a single facility (OEC95a)). The British waste
management philosophy favors the use of broad safety goals over prescriptive regulatory
approaches, placing the burden of compliance upon the operator. . ;
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REFERENCES
AEC94 Atomic Energy Commission (Japan), Long-Term Program for Research,
Development, and Utilization of Nuclear Energy, 1994.
AEC87 Atomic Energy Control Board (Canada), AECB Regulatory Document R-104,
June 1987.
BRE99 Brennecke, P.W. et al., Realization of the German Repository Concept - Current
Status and Future Prospects, WM '99 Conference, Tucson AZ, February 28-
March4, 1999.
DSI91 Directorate for the Safety of Nuclear Installations (France), Rule No. HI.2.f., June
10, 1991.
EAP98 Environmental Assessment Panel, Report of the Nuclear Fuel Waste Management
and Disposal Concept Environmental Assessment Panel, Canadian Environmental
Assessment Agency, February 1998.
EIA95 Energy Information Administration, World Nuclear Outlook, 1995, DOE/EIA-
0436(95), October 1995.
GAO94 General Accounting Office, Nuclear Waste: Foreign Countries' Approaches to
High-level Waste Storage and Disposal, GAO/RCED-94-172, August 1994.
GER94 Article 4, Section 9a(l) of German Atomic Energy Law; amended to the law on
June 19, 1994.
HED99 Hedman, T., Management of Spent Nuclear Fuel in Sweden, WM '99 Conference,
Tucson AZ, February 28-March 4, 1999.
HSK93 Nuclear Safety Division (Switzerland), Radiation Protection for the Disposal of
Nuclear Waste, Regulatory Document R-21, 1993.
JOR99 Jorda, M. and I. Forest, The Radioactive Waste Management Program in France,
WM '99 Conference, Tucson AZ, February 28 - March 4, 1999.
HOL99 Holmes, John and John Mathieson, Recent Developments in the United Kingdom
Programme for the Deep Disposal of Radioactive Wastes, WM '99'Conference,
Tucson AZ, February 28-March 4, 1999.
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KIN99 King, Frank, Recent Developments in Nuclear Waste Management in Canada,
WM '99 Conference, Tucson AZ, February 28-March 4, 1999.
LOM95 Lommerzheim, A., Repository Planning for the Gorleben Working Model,
Proceedings of the Fifth Annual Conference on Radioactive Waste Management
and Environmental Remediation, Volume 2, New York: American Society of
Mechanical Engineers, 1995.
MAS99 Masuda, S. et al., Key Aspects of the Second Progress in the Japanese R&D
Programme for HLW Disposal, WM '99 Conference. February 28-March 4, 1999.
NIR94 United Kingdom Nirex Limited, Going Underground: An International
Perspective on Radioactive Waste Management and Disposal, 2nd Edition,
September 1994.
NWT94 United States Nuclear Waste Technical Review Board, Report to the U.S.
Congress and the Secretary of Energy: January to December 1993, May 1994.
NWT95 United States Nuclear Waste Technical Review Board, Report to the US Congress
and Secretary of Energy: 1994 Findings and Recommendations, March 1995.
OEC93 Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Nuclear Waste Bulletin: Update on Waste Management Policies and
Programs, No. 8, July 1993.
OEC95a Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Future Human Action at Disposal Sites, OECD/NEA Document 66-94-
041, 1995.
OEC95b Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Nuclear Waste Bulletin: Update on Waste Management Policies and
Programs, No. 10, June 1995.
OEC95c Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Nuclear Energy Programs of OECD/NEA Member Countries, 1995.
OEC96 Organization for Economic Cooperation and Development/Nuclear Energy
Agency, Nuclear Waste Bulletin: Update on Waste Management Programs,
No. 11, June 1996. ;
SAN99 Santiago, J. and J. Astudillo, Overview of the Spanish R&D Program for High-
level Waste Disposal, WM '99 Conference, Tucson AZ, February 28-March 4,
1999. i
SKB94 Swedish Nuclear Fuel and Waste Management Company, Activities 199$, 1994.
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SKB95
SKB96
SMI91
VTT93
YJT92
Swedish Nuclear Fuel and Waste Management Company, RD&D Programme 95:
Treatment and Final Disposal of Nuclear Waste, September 1995.
Swedish Nuclear Fuel and Waste Management Company, SKB Annual Report
1995, May 1996.
Spanish Minesterio de Industria, Comercio Y Turismo, Third General
Radioactive Waste Plan, July 1991.
Technical Research Centre of Finland, Final Report of the Project, Performance
Assessment and Economic Evaluation of Nuclear Waste Management, 1993.
Nuclear Waste Commission of Finnish Power Companies, TVO-92. Safety
Analysis of Spent Fuel Disposal, December 1992.
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CHAPTER 4
U.S. PROGRAMS FOR THE MANAGEMENT AND DISPOSAL OF
SPENT NUCLEAR FUEL AND HIGH-LEVEL RADIOACTIVE
WASTE AND THE EVALUATION OF YUCCA MOUNTAIN
4.1 INTRODUCTION
The Department of Energy (DOE), the U.S. Nuclear Regulatory Commission (NRC), and the
U.S. Environmental Protection Agency (EPA) each have legislatively defined roles in
management and disposal of spent nuclear fuel and high-level radioactive wastes (HLW) at the
proposed Yucca Mountain disposal site. As stated in the Nuclear Waste Policy Act of 1982
(NWP83), DOE is responsible for developing, constructing, and operating repositories for
disposal of these wastes. The NRC has responsibility to license the repository and related
facilities, and the EPA is to promulgate radiation protection standards which the NRC is to adopt
as basis for their licensing actions. The Nuclear Waste Policy Amendments Act of 1987
(NWP87) designated the Yucca Mountain site in Nevada as the only site to be evaluated by DOE
as a potential location for disposal of spent fuel and HLW. The Energy Policy Act of 1992
(EnPA92) directed EPA to promulgate site-specific radiation protection standards for the Yucca
Mountain site.
The legislative framework also prescribes roles for state governments, local governments, and
Indian tribes in the waste management and disposal program, and establishes .the Nuclear Waste
Technical Review Board which provides oversight of the DOE program. This chapter presents
an overview of the responsibilities and program activities of the DOE, NRC, and these groups.
Responsibilities and activities of the EPA are described in Chapter 1 of this BID.
4.2 THE DEPARTMENT OF ENERGY
As noted above, DOE is responsible for the management and disposal of high-level radioactive
waste, which includes spent nuclear fuel and other waste generated by nuclear reactors and
reprocessing plants.14 Disposal of these wastes would occur at the Yucca Mountain site if it is
found suitable and approved for this function. Other radioactive waste categories defined by
14
DOE typically separates spent nuclear fuel from other high-level waste by definition, although NRC
includes spent fuel as part of its high-level radioactive waste category.
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DOE are transuranic (TRU) and low-level waste (LLW). TRU, consisting of material with
atomic numbers greater than 92, is generated as a result of defense production operations. DGd
began disposal of TRU wastes at the Waste Isolation Pilot Plant (WIPP), in New Mexico, in
1999. LLW is buried at DOE sites where it is generated or, if commercially generated, at sites
operated by private firms in several locations.
Fulfillment of its responsibility for radioactive waste management and disposal involves four
principal program activities in DOE: (1) receipt, transport, interim storage, and disposal of sp^ > r
nuclear fuel from commercial nuclear power operations, (2) management and disposal of DOE
spent nuclear fuel, which originates from DOE production and research operations and from
naval propulsion reactors, (3) solidification and disposal of high-level waste generated by
reprocessing operations for spent nuclear fuel from DOE's production reactors at Hanford and
Savannah River, (4) storage and disposition of fissile materials from dismantled nuclear
weapons, and (5) disposal of high-level waste from a former commercial waste processing
facilities at West Valley, NY now managed by DOE. Materials generated by the dismantling of
weapons may be treated and disposed of like high-level radioactive waste or'they may: be used as
reactor fuel. In either case, such materials will eventually become part of the disposal inventory.
I
In addition to commercial and DOE spent nuclear fuel, and high-level waste from DOE and
commercial processing operations, other radioactive wastes that have been considered for
disposal in a repository at Yucca Mountain include fissile materials from dismantled nuclear
weapons, and low-level radioactive wastes known as Greater-Than-Class-C (GTCC). ;The
radioactivity levels of wastes in this latter category exceed the NRC's limits for Class C wastes
as established in the 10 CFR Part 61 regulations. Decisions concerning disposition of these
radioactive materials have not been made. The NWPA limits the contents of a repository at
Yucca Mountain to "...70,000 metric tons of heavy metal or a quantity of solidified high-level
radioactive waste resulting from the reprocessing of such a quantity of spent nuclear fuel until
such time as a second repository is in operation" (NWPA, Section 114(d)). As detailed in
Chapter 7 of this BID, DOE currently plans that a repository at Yucca Mountain would contain
approximately 63,000 metric tons of spent commercial reactor fuel. Defense high-level wastes,
DOE spent nuclear fuel, and Navy spent fuel would contribute the equivalent of 7,000 metric
tons of heavy metal.
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4.2.1 DOE'S Office of Civilian Radioactive Waste Management (OCRWM)
The DOE's Office of Civilian Radioactive Waste Management (OCRWM) was established by
Congress specifically to provide management and disposal of spent nuclear fuel from commercial
nuclear power reactors. Under a 1985 Presidential Executive Order, the repository established by
OCRWM is also to be used for disposal of high-level waste from DOE operations. The
OCRWM charter includes responsibility for receipt of spent nuclear fuel from commercial
reactors at the reactor sites and from storage at DOE sites, interim storage of spent nuclear fuel as
necessary prior to disposal, transport of spent nuclear fuel to the site for interim storage and
disposal. The Navy program, which manages a small portion of DOE spent nuclear fuel, will
transport their own spent fuel to the repository. DOE has developed alternative designs for a
central interim storage facility (known historically as a Monitored Retrievable Storage (MRS)
facility), but, as of March 2000, the Department has not established a site for such a facility.
Since passage of the Nuclear Waste Policy Amendments Act in 1987, OCRWM activities have
been focused on evaluating Yucca Mountain as the disposal site for spent nuclear fuel and high-
level waste. In accordance with the Site Characterization Plan (DOE88a), characterization of the
Yucca Mountain site is proceeding with surface-based and sub-surface activities. Recently, DOE
has focused on the "Viability Assessment" (VA), which is intended to allow a greatly improved
appraisal of the prospects for geologic disposal at the Yucca Mountain site. The VA consists of:
A reference engineered design for the repository and the waste package
• A total system performance assessment describing the probable behavior of the
repository based on available data and the reference engineered design
A plan and cost estimate for completing a License Application (LA)
Cost estimates for constructing and operating the repository
The VA was published in December 1998. It was the basis for continued evolution of the
engineered design for the repository and for future data acquisition activities. DOE has issued a
Draft Environmental Impact Statement (EIS) and is in the public comment phase to issue a final
EIS. The Total System Performance Assessment for Site Recommendation (TSPA-SR) was
published in late 2000. The Site Recommendation is planned to be submitted to the President in
2001 if the site is found suitable, and the License Application (LA) is planned to be submitted to
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the NRC in about 2002 (depending resources) if the site is approved for disposal. To date,
principal program accomplishments include:
I
Completion of excavation of the north-south Exploratory Studies Facility (ESF)
tunnel at Yucca Mountain and the Enhanced Characterization Repository Block
Cross-Drift; both excavations have been mapped and will be used as sources of
in-situ data at the repository horizon
• Initiation of various types of testing in alcoves and niches in the ESF and the
Cross-Drift
• Development of a market-driven plan for to storage and transportation of
commercial spent nuclear fuel i
Completion of the TSPA-SR, which included an analysis of enhanced design
alternatives aimed at resolving some of the issues identified in the VA.
• Publication of the Draft Environmental Impact Statement
The OCRWM program has produced thousands of technical documents concerning its mission
l
and activities. Future technical documents are expected to support the Environmental Impact
Statement, the Site Recommendation, and the License Application if the site is approved for
disposal.
4.2.2 DOE Management and Disposal of Defense Wastes
The DOE's defense programs have produced significant amounts of high-level waste that may
eventually be disposed of in a repository at Yucca Mountain (see Chapter 5). Other wastes
produced by these defense programs (e.g., TRU waste) will be managed and disposed of
separately.
During the last 40 years, DOE and its predecessor agencies generated, transported, received,
stored, and reprocessed spent nuclear fuel at facilities throughout its nationwide complex. Spent
nuclear fuel was generated by nuclear weapons production reactors; U.S. Navy nuclear!
propulsion program power reactors; government, university, and test reactors; special-case
commercial reactors; and research reactors. The DOE operated production reactors at the
Hanford and Savannah River Sites to provide special nuclear materials and other isotopes. These
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production reactors are no longer operating. However, the Naval Nuclear Propulsion Program
and some test and research reactors are still in operation.
The DOE has reprocessed more than 100,000 metric tons of heavy metal (MTHM) of spent
nuclear fuel.at the Idaho National Engineering and Environmental Laboratory (INEEL), the
Hanford Site, and the Savannah River Site to recover fissile material (uranium-235 and
plutonium-239) and other nuclides needed for national defense or research and development
programs. These reprocessing operations generated large quantities of high-level radioactive
waste. This waste exists as liquid, sludge, solids, and calcine and is stored primarily at its
reprocessing sites.
In April 1992, the DOE began to phase out defense spent nuclear fuel reprocessing. As a result,
approximately 2,500 MTHM of unreprocessed spent nuclear fuel exist today in the DOE
inventory. This spent nuclear fuel is in a wide range of enrichments and physical conditions and
is stored at several locations throughout the United States. The majority of this spent fuel and
high-level waste is stored at three major sites in Idaho, South Carolina, and Washington. In
addition to this inventory, the DOE estimates that over the next 40 years it will generate another
100 MTHM from defueling DOE and naval reactors.
4.3 THE NUCLEAR REGULATORY COMMISSION
The NRC is responsible for licensing and regulating the receipt and possession of high-level
waste, including spent fuel, at privately owned facilities and at certain facilities managed by
DOE. This responsibility will extend to a repository at Yucca Mountain. The NRC currently
licenses temporary storage facilities at reactor sites, as well as commercial spent nuclear fuel
storage facilities at West Valley, New York, and Morris, Illinois.
4.3.1 Legislative Requirements and Regulatory Framework
The NWPA specifies that licensing of a geologic repository will occur in three phases. In the
first phase, which follows site characterization and approval of the site for disposal, DOE will
submit a License Application (LA) for the repository to NRC. After the LA is submitted, NRC
will have three years to perform its review, conduct a public hearing, and reach a construction
authorization decision by an independent licensing board. To comply with this schedule, NRC is
already reviewing DOE's site characterization, repository design, and performance assessment
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activities to identify and resolve potential licensing issues. However, during the licensing
proceeding itself, all issues, including those previously resolved, can potentially be re-opened by
the licensing board. ;
In the second phase, as construction of the repository nears completion, DOE will request a
license to receive high-level waste and spent nuclear fuel. Only after that license is granted will
DOE begin placing waste into the repository. In the third phase, when all waste is in place, DOE
will apply for a license amendment to decommission and permanently close the disposal facility.
The.NWPA directed both EPA and NRC to publish standards and criteria for the storage and
disposal of high-level waste. In response to the NWPA, NRC developed a generic regulation for
geologic disposal at 10 CFR Part 60. Although the regulation has been amended several times,
the technical criteria date to 1983. As previously noted, the Energy Policy Act of 1992 directed
EPA to develop new individual dose standards for the Yucca Mountain site and for NP>C to
conform its standards to the new EPA standards. In light of the requirements of the EnPA, NRC
has elected to develop additional regulations specific to Yucca Mountain. To that end Ithe
Commission has proposed a new rule at 10 CFR Part 63 entitled "Disposal of High-Level
Radioactive Waste in a Proposed Geologic Repository at Yucca Mountain Nevada". Additional
discussion of the proposed rule is included in Chapter 2 of this BID.
The proposed NRC 10 CFR Part 63 regulation does not contain prescriptive criteria, but does
require DOE to demonstrate defense in depth. Under its own 10 CFR Part 960 regulations, if
DOE identifies potentially adverse conditions, the Department must demonstrate that the
conditions can be compensated for by the repository design or favorable site conditions. DOE
has proposed revision of the 10 CFR Part 960 siting guidelines to 10 CFR Part 963, which would
base site-suitability evaluation on total system performance assessment.
4.3.2 Status of NRC's Program ;
The NRC's Prelicensing High-Level Waste Repository Program is currently part of the NRC's
Office of Nuclear Material Safety and Safeguards (ONMSS). This program was refocused in
FY 1996 based on three events: (1) a reduction in congressional funding, (2) a reorganization of
DOE's high-level waste program, and (3) the publication of the National Academy of Sciences'
report, Technical Bases for Yucca Mountain Standards (NRC97). ;
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The NRC program is now focused on the following ten issues which the Commission believes
are most important to repository performance:
Igneous activity
« Structural deformation and seismicity
« Evolution of the near-field environment
« Container life and source term
« Thermal effects on flow
" Repository design and thermal-mechanical effects
« Total system performance assessment and integration
• Activities related to the development of the NRC high-level waste regulations
« Unsaturated and saturated flow under isothermal conditions
« Radionuclide transport
The status of resolution of these Key Technical Issues (KTIs) will be periodically re-evaluated
based on new information, performance assessments, and technical interactions with DOE.
4.4 NUCLEAR WASTE TECHNICAL REVIEW BOARD
The NWPAA established the Nuclear Waste Technical Review Board comprising 11 members
recommended by the National Academy of Sciences and appointed by the President. These
individuals are experts in the fields of science, engineering, or environmental sciences and
represent a broad range of scientific and engineering disciplines, including hydrology,
underground construction, hydrogeology, and physical metallurgy. No member of the Board may
be employed by DOE, its contractors, or the National Laboratories. The current Board is
composed of individuals with academic and public and private sector experience.
As defined in Section 503 of the NWPAA,
The Board shall evaluate the technical and scientific validity of activities
undertaken by the Secretary [of Energy]..., including-
(1) site characterization activities, and
(2) activities related to the packaging or transportation of high-level radioactive
waste or spent nuclear fuel.
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The NWTRB meets four times a year in open public meetings. Two of these meetings jare held
in Nevada. In addition, the Board reports to Congress and to the Secretary of Energy at least
twice a year on scientific issues associated with the high-level waste and spent fuel disposal
program. The Board also publishes a periodic newsletter and other information about its views
and activities. Information concerning the Board's membership, activities, and links to NRC and
DOE activities can be found at the Board's website, www.nwtrb.gov.
4.5 STATE AND LOCAL AGENCIES •
Congress provided for active State participation in both the NWPA and the NWPAA. the
NWPAA provides for financial assistance to the State of Nevada and any affected unit of local
government to allow for participation in activities related to the establishment of a repository at
Yucca Mountain. Specific activities include: '
• Reviewing all work done at the Yucca Mountain site to determine any pbtential
economic, social, public health and safety, and environmental impacts of a
repository on a State or local government and its residents ;
• Developing an impact assistance request ;
• Monitoring, testing, or evaluating site characterization programs
• Providing information to State residents ;
• Requesting information from and making comments or recommendations to the
Secretary of Energy
The State of Nevada and any affected unit of local government may also request assistance to
mitigate any economic, social, public health and safety, and environmental impacts that are likely
to result from site characterization activities at Yucca Mountain. The NWPAA specifies that this
financial assistance shall continue until "such time as all such activities, development, and
operation are terminated at such site." •
The Nevada legislature created the State's Nuclear Projects/Nuclear Waste Project Office
(NWPO) in 1985 to oversee Federal high-level nuclear waste activities in the State. Since then,
the NWPO has dealt primarily with the technical and institutional issues associated with DOE's
efforts to characterize the Yucca Mountain site.
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Yucca Mountain lies in Nye County, Nevada. This county and nine others that are contiguous
have been designated "affected" and are therefore eligible to receive financial assistance under
the NWPAA. Nye County sponsors a year-round on-site representative. The nine other counties
include: Churchill County, Clark County, Esmeralda County, Eureka County, Lander County,
Lincoln County, Mineral County, and White Pine County, all in Nevada; and Inyo County,
California.
4.6 NATIVE AMERICAN TRIBES
Native American tribes have a unique sovereign status in U.S. law which was recognized by the
NWPA and the NWPAA. This govemment-to-government relationship between the Federal
Government and the tribes obligates the Federal Government to interact directly and specifically
with tribes in areas where repository or MRS siting activities will occur. The NWPA, as
amended, under Section 2(2), defines an affected tribe as any tribe:
• (A) within whose reservation boundaries monitored retrievable storage
facility, test and evaluation facility, or a repository for high-level waste or
spent nuclear fuel is proposed to be located; or (B) whose federally
defined possessory or usage rights to other lands outside of the
reservation's boundaries arising out of congress tonally ratified treaties
may be substantially and adversely affected by the locating of such a
facility. Provided, That the Secretary of the Interior finds, upon the
petition of the appropriate governmental officials of the tribe, that such
effects are both substantial and adverse to the tribe... (NWP83)
As noted above, specific provisions of the NWPA, as amended, that delineate the participation
activities and rights of affected States in repository and MRS siting decisions also apply to
affected tribes. The means for an affected tribe to disapprove of the site selection and
designation process is given in Section 118(a). An affected tribe is also eligible to receive the
same grants, financial and technical assistance, and payments equal to taxes for which a State is
eligible under Section 116(c). Since the passage of the NWPAA, no tribes have been designated
as affected tribes. However, to ensure compliance with the American Indian Religious Freedom
Act (AIRFA), the National Historic Preservation Act (NHPA) and related statutes, the Native
American Graves Protection and Repatriation Act (NAG90) and the National Environmental
Policy Act (NEPA), the DOE is cooperating with Indian tribes that have current or traditional
religious or cultural ties to the Yucca Mountain site or that may be located near the transportation
routes to or around the site (DOE88b).
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In 1985 and in keeping with the NHPA, the Advisory Council on Historic Preservation I(ACHP)
issued guidelines for discussing which tribes should be involved in the Yucca Mountain cultural
resource study (STO90). The guidelines contributed to the Yucca Mountain Project's >
Programmatic Agreement (PPA), which was jointly produced by DOE and the ACHP. The PPA
requires that DOE consult with tribal groups having traditional cultural ties to the Yucck
Mountain area prior to land-disturbing activities to assure that cultural or religious values are
preserved to the extent practicable. The PPA further stipulates that when such activities are
thought to have a negative effect that cannot be avoided, the DOE will consult further with the
tribal groups and others to identify ways to mitigate those effects. [
DOE has established the Yucca Mountain Site Characterization Project, which led to the Cultural
Resources Program to meet resource preservation requirements set forth in the PPA. The
preliminary site characterization (DOE87) identified the ethnic and tribal affiliations of the tribal
groups most likely to have traditional ties to cultural resources located in the Yucca Mc-untain
region. These groups consist of Southern Paiute, Western Shoshone, and Owens Valley
Paiute/Shoshone people from Nevada, Utah, Arizona and California. Extensive ethnographic
research led to the identification of 15 tribes and one Native American organization. In the mid-
1990s, an additional tribe was included. The following 17 tribal entities are commonly involved
in the Yucca Mountain Cultural Resources Program: .
i
1. Benton Paiute Indian Tribe, California .
2. Timbisha Shoshone Tribe, California :
3. Bishop Paiute Indian Tribe, California '-
4. Big Pine Indian Tribe, California
5. Fort Independence Indian Tribe, California '
6. Lone Pine Indian Tribe, California
7. Yomba Shoshone Tribe, Nevada
8. Duckwater Shoshone Tribe, Nevada
9. Pahrump Paiute Indian Tribe, Nevada
10. Las Vegas Paiute Indian Tribe, Nevada ;
11. Las Vegas Indian Center, Nevada
12. Chemehuevi Tribe, California
13. Colorado River Indian Tribes, Arizona j
14. Moapa Paiute Tribe, Nevada ;
15. Paiute Indian Tribes of Utah
16. Kaibab Paiute Tribe, Arizona
17. Ely Shoshone Tribe, Nevada !
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All groups requested that they be included in the project. The DOE informs tribes of the status of
the project through a cooperative agreement with the National Congress of American Indians.
Through this group, DOE and the tribal governments have established a consulting relationship
through which the concerns of the tribal peoples can be expressed.
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REFERENCES :
DOE87 U. S. Department of Energy, Native Americans and Nuclear Waste Storage at
Yucca Mountain, Nevada: Potential Impacts of Site Characterization Activities,
Ann Arbor: Institute for Social Research, University of Michigan, 1987.
DOE88a U. S. Department of Energy, Site Characterization Plan, Yucca Mountain Site,
Nevada Research and Development Area, DOE/RW-0199, December 1988.
DOE88b U. S. Department of Energy, Draft 1988 Mission Plan Amendment, DOE/RW-
0187, June 1988.
EnPA92 Energy Policy Act of 1992, Public Law 102-486, October 24, 1992. |
NAG90 Native American Graves Protection and Repatriation Act, Public Law 101 -601,
November 1990. '..
NRC97 U.S: Nuclear Regulatory Commission, NRC High-level Radioactive Waste
Program Annual Progress Report: Fiscal Year 1996, NUREG/CR-6513, No. 1,
January 1997. ' \
NWP83 Nuclear Waste Policy Act of 1982, Public Law 97-425, January 7, 1983.
NWP87 Nuclear Waste Policy Amendments Act of 1987, Public Law 100-203, December
22, 1987.
STO90 Stoffle, Richard W., David B. Halmo, John E. Olmsted, and Michael J. Evans,
Native American Cultural Resource Studies at Yucca Mountain, Nevada, Ann
Arbor: Institute for Social Research, University of Michigan, ISBN 0-877944-
328-6, 1990. I
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CHAPTER 5
QUANTITIES, SOURCES, AND CHARACTERISTICS OF SPENT NUCLEAR
FUEL AND HIGH-LEVEL WASTE IN THE UNITED STATES
5.1 INTRODUCTION
This chapter presents current and projected inventories of spent nuclear fuel and DOE defense
high-level radioactive waste. Current plans call for both of these waste forms to be disposed of
in the Yucca Mountain repository. The waste inventories cited are from sources of Federal
Government information publicly available (DOE94a, DOE95a, DOE95b, DOE95c, DOE95d,
DOE95e, DOE95f). The waste forms are inventoried by mass or volume and radioactivity
content.
Since this BID was prepared, DOE has provided updated information on spent nuclear fuel in the
Draft Environmental Impact Statement (DEIS)(DOE99). While some of the detailed waste
inventory values reported in the DEIS may differ from those reported here, they do not
substantively affect the technical discussion in this chapter nor EPA's regulatory considerations.
Information in this chapter describes fuel inventories either in terms of metric tons of initial
heavy metal (MTEHM) or in terms of metric tons of heavy metal (MTHM) depending on the
metric used in the source document. The former term (MTfflM) is useful since it is a metric that
is independent of fuel bumup, while the latter term (MTHM) is useful since it is a metric that is
consistent with the repository regulatory limit (i.e., 70,000 MTHM). Heavy metal refers to the
mass of actinide elements (elements with atomic numbers greater than 89) in the fuel. Generally,
the initial heavy metal is mostly uranium. Differences between MTIHM and MTHM are small.
5.2 SPENT NUCLEAR FUEL
Spent nuclear fuel is defined as fuel that has been withdrawn from a nuclear reactor following
irradiation and whose constituent elements have not been separated by reprocessing (EPA85).
Generators of spent nuclear fuel include: (1) commercial Light Water Reactors (LWR), which
consist of Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR);
(2) government-sponsored research and demonstration programs, DOE test and research reactors,
universities, and industry; (3) experimental reactors, e.g., liquid-metal, fast-breeder reactors
5-1
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(LMFBR) and high-temperature gas-cooled reactors (HTGR); (4) U.S. Government nuclear
weapons production reactors; and (5) Department of Defense (DOD) reactors. ;
Approximately 98 percent of the spent nuclear fuel from commercial power reactors is stored at
the reactor sites where it was generated; the rest is stored at central commercial storage facilities.
The majority of DOE spent nuclear fuel is stored at three sites—the Hanford Site in Washington,
the Idaho National Engineering and Environmental Laboratory (INEEL), and the Savaiinah River
Site in South Carolina. Some of the Fort St. Vrain spent nuclear fuel is being stored at INEEL,
but the remainder is being stored in Colorado at the Fort St. Vrain facility (DOE95a). i The fuels
at these DOE facilities are Government-owned and are not scheduled for reprocessing; in support
of DOE defense activities. i
!
The fuel for LWRs consists of uranium dioxide pellets encased in zirconium alloy (Zircaloy) or
stainless steel tubes. During reactor operation, fission of the uranium-235 produces etiergy,
neutrons, and radioactive isotopes known as fission products. The neutrons produce further
fission reactions and thus sustain the chain reaction. The neutrons also convert a portion of the
uranium-238 into plutonium-239, which can also undergo fission. In time, the fissile Suranium-
235, which originally constituted some 3 to 4 percent of the enriched fuel, is depleted; to such a
level that power production becomes inefficient. Once this occurs, the fuel bundles are deemed
"spent" and are removed from the reactor. In the United States, reprocessing of commercial
spent nuclear fuel to recover the unfissioned uranium-235 and the plutonium for reuse as a fuel
resource is currently not taking place, nor is it expected to occur in the future. j
i
The radioactive materials associated with spent nuclear fuel fall into three categories: (1) fission
products; (2) actinide elements (atomic numbers of 89 and greater); and (3) activation products.
Typically, fresh spent nuclear fuel contains more than 100 radionuclides as fission products.
Fission products are of particular importance because of the quantities produced, their high
radiological decay rates, their decay-heat production, and their potential biological hazard. Such
fission products include: strontium-90; technetium-99; iodine-129 and -131; cesium isotopes,
such as cesium-134, -135, and -137; tin-126; and krypton-85 and other noble gases.
Activation products include tritium (hydrogen-3), carbon-14, cobalt-60, and other radioactive
isotopes created by neutron activation of fuel assembly materials and impurities in cooling water
or in the spent nuclear fuel. The actinides include uranium isotopes and transuranic elements,
such as plutonium, curium, americium, and neptunium. The exact radionuclide composition of a
5-2 \
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particular spent nuclear fuel sample depends on the reactor type, the initial fuel composition, the
length of time the fuel was irradiated (also known as "burnup"), and the elapsed time since its
removal from the reactor core.
5.2.1 Commercial Spent Nuclear Fuel Inventory and Projection
By the end of 1999, there were 40,000 MTIHM15 of spent nuclear fuel in inventory from
commercial reactor operations. Approximately 37,000 MTIHM is stored at reactor sites. The
remainder is stored at the West Valley Demonstration Project (WVDP) (27 MTIHM) in West
Valley, New York; the Idaho National Engineering and Environmental Laboratory (43 MTIHM)
in Idaho Falls, Idaho; and the Midwest Fuel Recovery Plant (MFRP) (744 MTIHM) in Morris,
Illinois (DOE95e). The historical (1970-1994) and projected (1995-2030) spent nuclear fuel
inventories and accumulated radioactivities are given in Table 5-1.
Table 5-1. Historical and Projected Mass and Radioactivity of Commercial Spent Nuclear Fuel
(DOE94a, DOE95e, DOE95f, DOE96a)
End of Calendar Year
1970
1975
1980
1985
1990
1994
1995C
2000C
2005C
2010C
2015C
2020C
2025C
2030C
Mass Accumulated
(MTIHM)1
55
1,567
6,558
12,684
21,547
29,811
32,022
43,100
53,500
63,600
73,900
80,000
85,500
87,900
Radioactivity Accumulated
(106Ci)b
215
3,315
. 10,137
14,228
22,910
26,661
25,600
32,600
36,900
39,800
36,700
34,700
32,100
24,700
a Metric tons initial heavy metal refers to the original mass of the actinide elements of the fuel.
b A curie of radioactivity corresponds to 3.7 x 10'° disintegrations per second.
c Projections beyond 1994 are based on the DOE/EIA Low Case.
15 Commercial spent nuclear fuel reported in DOE95e is in units of metric ton (tonne) of initial heavy metal
(MTIHM) to avoid difficulties arising from the need to estimate ranges of varied heavy-metal content that result
from different levels of enrichment and reactor fuel bumup. A metric ton is 1,000 kilograms, corresponding to about
2,200 pounds.
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Projections of nuclear capacity are based on the DOE/EIA Low Case assumptions, whiph
forecast an increase in the installed nuclear capacity from 99.1 gigawatts-electric (GW(e)) in
1994 to a peak of 100.3 GW(e) in 2000, then a decrease to 2.3 GW(e) by 2030, as shown in
. Table 5-2 (DOE95f). The Low Case scenario is based on these assumptions: (1) reprocessing of
spent nuclear fuel will not occur; (2) currently licensed reactors will be retired when their initial
license terms expire; and (3) new advanced LWRs will not be available before 2015. The
DOE/EIA projections also assume that bumup levels of spent nuclear fuel will increase from
their current average of 33,065 and 39,989 Megawatt days (MWd) per MTfflM to 42,QOO and
54,000 MWd/MTIHM for BWRs and PWRs, respectively. This increase is predicted over the
period 1994 to 2020. Based on currently-mandated limits, only 63,000 MTHM of commercial
spent nuclear fuel can be accommodated at the Yucca Mountain site. |
Table 5-2. Historical and Projected* Installed Nuclear Electric Power Capacity (D0E95f)
End of Calendar Year
1960
1965
1970 -
1975
1980
1985
1990
1994
Total GW(e)
0.3
0.4
5.8
38.3
51.9
78.5
99.6
99.1
End of Calendar Year
1995*
2000*
2005*
2010*
2015*
2020*
2025*
2030*
100.3 '
100.3 i
91.1 i
61.4 !
46.7 ;
22.0 I
2.3
* Lower Reference Case projected capacity includes all existing reactors, completed <
construction, plus additional new reactors beyond the year 2005.
5.2.2 DOE Spent Nuclear Fuel
The DOE reprocessed most of its spent nuclear fuel in the facilities at INEEL, the Hanford Site,
and the Savannah River Site. However, some spent nuclear fuel remains because of U.S.
Government decisions to stop reprocessing. Most of this fuel came from the Hanford Site N-
Reactor, a dual-purpose reactor designed to produce plutonium for use in nuclear weapons a.nd to
generate electricity for commercial use. Smaller amounts of spent nuclear fuel associated with
nuclear weapons production are stored at the Savannah River Site. Spent nuclear fuel|from the
Naval Nuclear Propulsion Program is stored at INEEL and, for short time, at some naval nuclear
shipyards. The DOE will also assume responsibility for fuel from some special-case commercial
nuclear reactors, foreign research reactors, and certain domestic research and test reactors. The
following sections discuss the nature and quantity of this spent nuclear fuel and DOE's plans to
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manage it. Most of the discussion that follows is derived from the Spent Nuclear Fuel FEIS
(DOE95c). Additional details are provided in the DEIS (DOE99).
Hanford Site
The Hanford Site produced plutonium for use in nuclear weapons from the start of the Manhattan
Project until DOE halted production in 1989. Hanford's production reactors generated 2,100
MTHM of the existing DOE inventory of spent nuclear fuel. There is a total of 2,096 metric tons
of spent N-reactor fuel at Hanford, which comprises all but about 1 percent by heavy metal mass
of the spent nuclear fuel inventory at the site. This fuel is stored in three facilities; DOE's
interim plans for management of this fuel include possible relocation to a single storage facility.
Sources of the other spent nuclear fuel at the site included single-pass Hanford production
reactors, the Fast Flux Test Facility, Shippingport Core H, and miscellaneous test facilities.
Idaho National Engineering and Environmental Laboratory
Six major facility areas at the INEEL store spent nuclear fuel: Argonne National Laboratory-
West; Idaho Nuclear Technology and Engineering Center; Naval Reactors Facility; Power Burst
Facility; Test Area North; and the Test Reactor Area. Spent nuclear fuel is kept in a variety of
dry and wet configurations. The INEEL stores about 10 percent of DOE's current inventory of
spent nuclear fuel, i.e., about 300 MTHM.
Savannah River Site
The DOE has 200 MTHM, or about 8 percent of its system-wide spent nuclear fuel inventory, in
storage at the Savannah River Site. This fuel is stored in the Receiving Basin for Off-site Fuels
(RBOF), in three reactor disassembly basins, and in basins in the F- and H-Area Canyons.
The F- and H-Area Canyons are among the only remaining operable chemical separation
facilities of their kind in the DOE complex. Each canyon has an associated storage basin that
serves as an interim staging area where spent nuclear fuel awaits chemical separation.
The DOE has stored most aluminum-clad spent nuclear fuel from Savannah River Site reactors in
water-filled concrete basins. These basins contain spent nuclear fuel and target material. The
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basin structures were built in the 1950s and were not intended for prolonged storage of J
radioactive materials.
The RBOF has been receiving fuels of U.S.-origin since 1964, including fuel manufactured in the
United States but irradiated in foreign reactors. About 50 percent of the fuels in the SR|S basins
consist of uranium clad in stainless steel or zircaloy. :
Other Generator/Storage Locations !
The DOE has in its possession, or has title to, a small amount of spent nuclear fuel in many other
locations throughout the United States. These locations include both DOE and non-DQE
facilities. For example, the Oak Ridge National Laboratory (ORNL) stores less than 1 MTHM of
spent nuclear fuel. This fuel is left over from research on fuel elements removed from \
commercial or demonstration reactors, as well as fuel removed from reactors that operated at
ORNL. This fuel will be transferred to either INEEL or the Savannah River Site. |
Besides ORNL, DOE is responsible for spent nuclear fuel from research and test reactors at the
Brookhaven, Los Alamos, Sandia, and Argonne-East Laboratories. These facilities ha\je a total
of about two MTHM in storage. Other E>OE sources include: ;
Non-DOE Research Reactors - The DOE has title to the spent nuclear fuel that is
stored at or is generated by 57 small research reactors. These reactors operate at
universities, commercial establishments and other government agencies; such as
' the Department of Defense. These reactors have a current inventory of jess than 5
MTHM and will generate very little additional spent nuclear fuel by 2035.
i
Commercial Power Reactors - The DOE has possession of 125 spent nuclear fuel
assemblies and 20 complete or sectioned spent nuclear fuel rods from several
commercial power reactors that supported DOE-sponsored research and
development programs. This fuel is stored at the West Valley Demonstration
Project and at the Babcock and Wilcox Lynchburg Technology Center in
Campbell County, Virginia. Other commercial spent nuclear fuel is already stored
at the INEEL, the Hanford Site, and the Savannah River Site. '
Foreign Research Reactors - The DOE has accepted limited amounts ofspent
nuclear fuel from foreign reactors (WCM95a). In some cases, this fuel was
manufactured by the DOE. The DOE will, under the Spent Nuclear Fuel FEIS,
continue to receive and store spent nuclear fuel from foreign sources (DOE95c).
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All the spent nuclear fuel listed above, if not already transferred, will be shipped to INEEL or the
Savannah River Site, with the exception of Fort St. Vrain spent nuclear fuel, which will remain
in Colorado.
Spent Nuclear Fuel Management Options
The Spent Nuclear Fuel FEIS says that most spent nuclear fuel in the possession of the DOE will
be stored until a geologic repository is available. Most DOE spent nuclear fuel will be stored dry
to reduce identified vulnerabilities. The DOE is also considering options to stabilize some of its
corroding spent nuclear fuel (WCM95b). One of these options is "melt and dilute" at the
Savannah River Site.
A summary inventory of DOE spent nuclear fuel is given in Table 5-3. Spent nuclear fuel in this
listing includes material from fuels other than those discharged from production reactors.
Table 5-3 includes nuclear fuel that has been withdrawn from or resides in storage at a reactor
following irradiation but that has not been reprocessed. Also included are some defective fuel
elements and special nuclear forms, as well as some commercially generated nuclear fuels and
fuels from foreign reactors and university research reactors.
The estimates in Table 5-3 have been recently updated .by DOE (DOE98, vol. 3, Table 3-13).
The more recent estimate is 2,496 MTHM. Based on legislative limits, only 2,333 MTHM is
scheduled for disposal at Yucca Mountain.
5.3 DEFENSE HIGH-LEVEL RADIO ACTIVE WASTE
High-level radioactive wastes are the highly radioactive materials resulting from the reprocessing
of spent nuclear fuel, including liquid waste produced directly in reprocessing, and any solid
material derived from such liquid waste (EPA85, NWP83). Commercial high-level radioactive
waste currently stored at the West Valley Demonstration Project will be converted to a solid form
(glass) prior to disposal (NRC88). Substantial quantities of this waste have been vitrified.
5-7
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Table 5-3. DOE Spent Nuclear Fuel Inventory (DOE95a)
Generator or Storage Site
DOE Sites
Hanford Site
Idaho National Engineering and
Environmental Laboratory
Savannah River Site
Oak Ridge Reservation
Other DOE Sites
Naval Nuclear Propulsion
Reactors
Foreign Research Reactors
Non-DOE Domestic
Domestic Research and Test
Reactors'
Special-Case Commercial SNF at
non-DOE locations'1
Total'
Existing (1995)
MTHM1 Percent
2132.44 80.6
261.23 9.9
206.27 7.8
0.65 <0.1
0.78 <0.1
0.00b 0.0
0.00 0.0
2.22 <0.1
42.69 1.6
2646.27
96.5
Future Increases
(through 2035)
MTHM Percent
0.00 0.0
12.92 13.5
0.00 0.0
1.13 1.2
1.50 1.6
55.00 57.6
21.70 22.7
3.28 3.4
0 0
95.53
3.5
Total (2,035)
MTHM ; Percent
2132.44 ; 77.8
274.14 ; 10.0
206.27 i 7.5
1.78 ' <0.1
2.28 ! 0.1
55.00 2.0
21.70 : 0.8
!•'
5.50 : 0.2
42.69 : 1.6
2741.80 '
[
100.0 !
- MTHM = metric tons of heavy metal.
b The existing inventory of Naval Nuclear Propulsion Program spent nuclear fuel (10.23 MTHM) stored at the
INEEL is included in the INEEL total. <
c Includes research reactors at commercial, university, and government facilities. . .„„.. ' ._ ,n
* The total inventory of spent nuclear fuel from special case commercial reactors is 186.41 MTHM. The 42.69
MTHM listed here is that stored at the Babcock & Wilcox Research Center, Fort St. Vram Reactor, and West Valley
Demonstration Project. The remaining special-case commercial spent nuclear fuel is stored at the INEEL, the Oak
Ridge Reservation, and the Savannah River Site, and is included in the totals for those locations.
e Numbers may not sum due to rounding.
High-level waste is generated by the chemical reprocessing of spent research and production
reactor fuel, irradiated targets, and naval propulsion fuel. The fission products, actinides, and
neutron-activated products of particular importance are the same for high-level waste as they are
for spent nuclear fuel assemblies (DOE88, DOE95e). ;
Weapons program reactors were operated mainly to produce plutonium. Reprocessing; to recover
the plutonium was an integral part of the weapons program. Naval propulsion and DO|E
research/test reactor fuel elements were also reprocessed to recover the highly enriched, uranium
that remained after use. DOE decided in 1992 to phase out the domestic reprocessing of
irradiated nuclear fuel of defense program origin, so minimal amounts of high-level waste will be
added to the current inventory. I
5-S
-------
High-level radioactive waste that is generated by the reprocessing of spent reactor fuel and
targets contains more than 99 percent of the nonvolatile fission products produced in the fuel or
targets during reactor operation. It generally contains about 0.5 percent of the uranium and
plutonium originally present in the fuel. Most of the current high-level waste inventory, which is
the result of DOE national defense activities, is stored at the Savannah River Site, INEEL, and
the Hanford Site. A limited quantity of high-level waste is stored at the West Valley
Demonstration Project. These high-level wastes have to date been through one or more treatment
steps (e.g., neutralization, precipitation, decantation, evaporation). It is currently planned that
this HLW will be solidified, using a vitrification process, for disposal. Vitrification is well
underway at the Savannah River Site and the West Valley Demonstration Project.
The DOE defense high-level waste at INEEL results from reprocessing nuclear fuels from naval
propulsion reactors and special research and test reactors. The bulk of this waste, which is
acidic, has been converted to a stable, granular solid (calcine). At the Savannah River and
Hanford Sites, the acidic liquid waste from reprocessing defense reactor fuel is or has been made
alkaline by the addition of caustic soda and stored in tanks. During storage, this alkaline waste
separates into three phases: liquid, sludge, and salt cake. The relative proportions of liquid and
salt cake depend on how much water is removed by waste treatment evaporators during waste
management operations.
Both alkaline and acidic high-level wastes were generated at West Valley. The alkaline waste
was generated by reprocessing commercial power reactor fuels and some Hanford N-Reactor
fuels. Acidic waste was generated by reprocessing a small amount of commercial fuel containing
thorium.
Projected volumes and total radioactivity for high-level waste stored at the Hanford Site, INEEL,
the Savannah River Site, and WVDP are given in Table 5-4. Projected inventories for each site
are based on specific assumptions and are subject to change. New treatment methods and waste
forms are possible and may affect the future projections. Since all sites are progressing toward
closure, there should be minimal amounts of waste added to the current inventory. Interim
storage of DOE high-level waste will be required and will most likely be at the site where the
waste is produced.
DOE currently estimates that 10,110 MTHM of HLW will be available for disposal (DOE98,
Vol. 3, Section 3.5.1.5). Of this total, only 4,667 MTHM are actually scheduled for disposal
based on a regulatory limit for the repository of 70,000 MTHM of spent nuclear fuel and HLW.
5-9
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5.3.1 High-Level Waste Inventories at the Hanford Site
The alkaline high-level waste (239,000 m3) located at Hanford is stored in underground carbon-
steel tanks. Currently 155,800 m3 is solid (salt cake and sludge) and 83,200 m3 is liquid; waste
volumes change with time because of on-going waste management activities. There are
approximately 350 million curies of total radioactivity contained in the waste, which has been
accumulating since 1944. The high-level waste was generated by reprocessing production
reactor fuel for the recovery of plutonium, uranium, and neptunium for defense and other Federal
programs.
All of the fuel reprocessing methods generated acidic waste streams. Sodium hydroxide or
calcium carbonate was added to the waste before it was transferred to the tanks to neutralize the
acid and minimize tank corrosion. The tanks currently contain moderate to strong alkaline
solutions. Additional post-processing of the waste to recover plutonium and uranium, or-to
reduce the volume of high-level waste, has resulted in the addition of ferrocyanide and some
organic compounds listed as hazardous. Fuel reprocessing was suspended from 1972 until
November 1983. Most of the high-heat-emitting isotopes (strontium-9Q, cesium-13 7, and their
decay products) have been removed from the old waste, converted to solids as strontium fluoride
and cesium chloride, placed in double-walled capsules, and stored in water basins. A total of
2,217 capsules were manufactured and 1,933 remain. (A portion of these capsules have been
used outside the facility or have been dismantled.)
Double-shell tanks continue to receive waste generated by decommissioning and cleanup of
Hanford Site facilities. This includes: effluents associated with the deactivation program for the
PUREX Plant; waste from B-Plant maintenance activities; laboratory waste; and miscellaneous
waste streams from ion-exchanger resin regeneration.
The tanks now contain a mixture of salt cake, liquid, and sludges with both radioactive and
hazardous components. Sludge consists primarily of solids (hydrous metal oxides) precipitated
from the neutralization of acid waste. Salt cake consists of the various salts formed from the
evaporation of water from the waste. Liquids exist as supernatant (liquid above solids) and
interstitial liquid (liquid filling the void between solids) in the tanks.
The tank waste is mostly inorganic, containing sodium hydroxide; salts of nitrate, nitrite,
carbonate, aluminate, and phosphate; and hydrous oxides of aluminum, iron, and manganese.
5-11
-------
The radioactive components consist primarily of long-lived fission products and shorter-lived
radionuclides, such as strontium-90 and cesium-137, and isotopes of uranium, plutoniu'm, and
americium. Some tanks contain the chelating agents EDTA and HEDTA. Some contain
halogenated and nonhalogenated organic contamination, while others contain mixed waste with
detectable levels of lead, chromium, and cadmium. ;
DOE has in place a program to treat and remediate some of this tank waste. In August 1998
DOE awarded a contract to BNFL Inc. to undertake tank waste remediation. Under the contract
BNFL will spent the initial two years in facility design. Assuming that DOE provides approval, a
facility will then be constructed and remediation will begin. Facility operation is expected to
[
begin in 2005 or 2006 and, during the initial ten-year operational period, DOE expects to process
waste from 11 storage tanks. The material treated during this initial phase is estimated to
constitute about 10 percent of the total waste mass and 20 to 25 percent of the total radioactivity.
DOE plans to separate tank contents into high-level and low-level components, thereby reducing
the amount of high-level radioactive waste. All remaining high-level liquid waste would then be
vitrified and placed in stainless steel canisters for storage on site until a geologic repository is
available for disposal. Vitrification is also planned for the low-level (low activity) waste. A
privatized high-level waste vitrification plant is currently scheduled to begin operation on HLW
in 2007 (90 percent confidence date) (DOE98a). !
5.3.2 High-Level Waste Inventories at INEEL . ;
About 11,000 m3 of high-level waste, containing approximately 50 million curies of total
radioactivity, is currently stored at INEEL; this volume consists of 7,200 m3 of acidic liquid
waste (1,306 m3 is high-level waste; the remainder is high-level waste that contains sodium) and
3,800 m3 of solid materials. Liquid high-level waste was generated at INEEL primarily by the
reprocessing of spent nuclear fuel from the national defense (naval propulsion nuclear reactors)
and reactor testing programs; a small amount was also generated by reprocessing fuel from non-
defense research reactors. This acidic waste is stored underground in large, high-integrity,
stainless steel tanks and these tanks are inside concrete vaults. Waste that has been converted to
a calcine is stored in retrievable stainless steel bins housed in reinforced concrete vaults. Greater
than 90 percent of the total radioactivity is contained in the calcine. '
5-12
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5.3.3 High-Level Waste Inventories at the Savannah River Site
Approximately 126,300 m3 of alkaline high-level waste that has accumulated at the Savannah
River Site over the past three decades is currently stored underground in carbon-steel tanks. The
current inventories consist of alkaline liquid, sludge, and salt cake that were generated primarily
by the reprocessing of nuclear fuels and targets from plutonium production reactors. The sludge
is formed after treatment with caustic agents. Salt cake results when the supernatant liquor is
concentrated in waste treatment evaporators. The high-level waste consists of 58,100 m3 of
liquid and 68,200 m3 of solid material having a total radioactivity of approximately 500 million
curies.
Tank farms at the Savannah River Site contain 24 single-shell and 27 double-shell tanks for
storing high-level waste. The DOE plans to remove the liquid waste from these tanks by 2035
(DOE95d). The removal process includes these process steps involved in vitrifying the waste:
• The salt solution is removed from the tanks and treated in the -salt processing
facility.
At the Defense Waste Processing Facility, which began operation in 1996, the
sludge is combined with glass frit and vitrified. The vitrified waste is contained in
stainless steel canisters.
5.3.4 High-Level Waste Inventories at the West Valley Demonstration Project
About 2,180 m3 of high-level waste is stored at the WVDP facility and consists of 2,040 m3 of
liquid alkaline waste and 140 m3 of solid waste (consisting of alkaline sludge and inorganic
zeolite ion-exchange medium). The alkaline waste is stored in an underground carbon-steel tank,
and the zeolite waste is stored in an underground carbon-steel tank covered by an aqueous
alkaline solution. Reprocessing was discontinued at the WVDP in 1972. No additional high-
level waste has been generated since.
hi June 1996, the vitrification of HLW into glass logs was initiated at the WVDP. The glass logs
are two feet in diameter by 10 feet long. As of mid-1999, more than 680 glass logs have been
made.
5-13
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5.4 SIGNIFICANT RADIONUCLIDES CONTAINED IN SPENT NUCLEAR FUEL AND
HIGH-LEVEL WASTE j
Of the 70,000-tonne capacity limit for Yucca Mountain, about 40,785 MTHM and |
22,210 MTHM represent spent PWR and spent BWR fuel, respectively (DOE95g). About
4,667 MTHM of vitrified high-level waste and 2,333 MTHM of DOE spent nuclear fuel
represent the balance of the total repository inventory. For the Yucca Mountain Site, |
radionuclide-specific activity levels are estimated by assuming that all spent fuel had been
removed from the reactors 30 years before emplacement with bumups of 39,651 MWd/MTHM
for PWR fuel and 31,186 MWd/MTHM for BWR fuel16. Although the burnup of spent: fuel
producing HLW is generally unknown, this uncertainty is thought to affect the adjustment for
decay only marginally.
Table 5-5 lists some radionuclide inventories for PWR and BWR reactor fuels, based on the
DOE assumptions concerning bumup and cooling time as cited above. These values are
generated from the ORIGEN2 computer code (ROD86), which calculates depletion, buildup, and
decay of isotopes for given fuel initial conditions and utilization histories. Also shown' in
Table 5-5 are estimated nuclide inventories for the defense high-level waste, based on ;
assumptions comparing bumup and fissile material contents for fuel from defense production
reactors and commercial power reactors. The values shown in Table 5-5 demonstrate that the
radionuclide inventories in a repository at Yucca Mountain stemming from defense high-level
wastes are expected to be much less than those from commercial spent fuel.
The radionuclide inventory of the repository will change with time due to radioactive decay and
ingrowth of radioactive decay products. For example, inventories of the initially-prominent
fis.sion products Cs-137 and Sr-90, which have approximately 30-year half lives, will decay to
insignificant levels within 1,000 years, while some decay products, such as Pb-210 and Ra-226,
will not achieve peak values until about 100,000 years after repository closure. Activity levels
for very long-lived radioisotopes will maintain low but nearly constant levels for periods on the
order of a million years. Overall, the radioisotope inventory of the wastes placed in the
repository will decrease by about five orders of magnitude during the first 100,000 years after
closure, and remain virtually constant thereafter.
16
Inventory and bumup values were slightly revised in the 1998 Viability Assessment.
5-14
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Table 5-5. Radionuclide Inventories in Spent Nuclear Fuel and High-Level Wastes
Expected to be Disposed in a Yucca Mountain Repository*
Isotope
Ac-227
Ag-108m
Am-241
Am-242m
Am-243
C-14
Cl-36
Cm-243
Cm-244
Cm-245
Cm-246
Cs-135
Cs-137
1-129
Mo-93
Nb-94
Ni-59
Ni-63
Np-237
Pa-231
Pb-210
Pd-107
Pu-238
Pu-239
Pu-240
Pu-241
Pu-242
Ra-226
Se-79
Sm-151
Sn-121m
Sn-126
Sr-90
Tc-99
Th-229
Th-230
U-232
U-233
U-234
U-235
U-236
U-238
Zr-93
BWR Inventory
1.70x JO'5
9.82 x 10-3
2.50 x 103
8.38 x 10°
l.OOx 10'
1.43x10°
1.04x 10-2
7.21 x 10°
3.52 x 102
6.24 x 10-2
l.OSxlO-2
4.32 x 10-'
5.19x10"
2.75 x lO'2
5.56 x 10-"
2.95 x 10-2
1.01 x. 10°
' 1.23 x 102
2.87x10-'
3.53 x 10-5
5.34 x 10-7
8.70 x lO'2
. 1.61 x 103
2.97 x 102
4.64 x 102
3.09 x 104
1.18x10°
2.19 x 10-6
3.73 x 10-'
2.80 x 102
9.52 x 10-'
6.26 x 10-'
3.79x10"
1.22 x 10'
2.07 x 10-7
3.53 x 10-"
1.79 x 10-2
4.30 xlO'5
1.41 x 10°
2.25 x lO'2
2.66 x 10-'
3.18x 10"'
1.98x 10°
PWR Inventory
1.85xlO-5
1.14X10'2
3.67 x 103
1.03x10'
2.09 x 10'
1.46x10°
1.14xlO-2
1.61x10'
9.86 x 102
2.20x10-'
4.39 x ID'2
5.04 x 10-'
6.96 x 10"
3.74 xlO-2
2.76 x 1C'2
1.41 x 10°
4.21 x 10°
5.44 x 102
4.27 x 10-'
3.82 xlO-5
5.37 x 1C'7
1.28x10-'
2.87 x 103
3.53 x 102
5.34 xlO2
4.61 x 10"
2.05 x 10°
2.24 xlO'6
4.96x10-'
3.66 xlO2
6.08x10-'
9.01 x 10-'
4.91 x 10"
1.55x10'
3.66 x 1C'7
3.72 xlO-4
4.16 x lO'2
6.57 x 1C'5
1.56x 10°
2.30 x lO'2
3.37 x 10-'
3.13x 10-'
2.36 x 10°
HLW Inventory
6.17x 10-"
0
1.05 xlO3
1.70x10-'
2.36x10-'
0
0
4.29 xlO'2
2.77x10'
5.47 x 10-4
6.19 x 10'5
2.98 x 10-'
2.97 x 10"
3.44 x 10-6
0
6.93 x lO'5
1.12x10-'
8.13x10°
6.69 x lO'2
1.44x 10-2
0
'2.74xlO-2
9.27 x 102
1.13x10'
7.77 x 10°
3.49 xlO2
1.16 xlO-2
0
1.88x10-'
4.66 xlO2
5.91 x 10'2
5.10x10-'
2.93 x 10"
6.33 x 10°
2.04 x 10-"
2.90 xlOJ
2. 12x1 Q-2
2.41 x lO'2
3.40 xlO-2
2.22 x lO-1
1.11 xlO-3
8.98x10-'
1.55 x 10°
Combined Weighted
Average
7.79 xlO'5
9.76 x 10-3
3.04 x 103
8.68 x 10°
1.54x 10'
1.30x10°
9.94 x 10-3
1.17x 10'
6.89 x 102
1.48x 10"'
2.90 x lO'2
4.61 x 10"'
6.00x10"
3.05 x lO'2
1.62 x lO'2
8.30x10"'
2.78 x 10°
3.57 x 102
3.46 x 10''
1.47x 10J
4.82 x ID'7
1.05 x 10''
2.28 x 103
- 3.01 x 102
4.59 x 102
3.67 x 10"
1.57x10°
2.00 xlO'6
4.26 x 10-'
3.49 x 102
6.62 x 10-'
7.75 x 10-'
4.36 x 10"
1.35x10'
2.07 x 10'5
3.58 x 10-"
3.20xlO-:
2.46 x ID'3
1 .36 x 10°
2.06 x 10'2
2.81 x 10''
2.84 x 10"'
2.16x 10°
*Inventories for spent BWR and PWR fuel are in curies per initial
are for estimated equivalent metric tonnes of heavy metal. Values
assumed by DOE.
metric ton of heavy
are based on burnup
metal. Inventories for HLW
and cooling histories
5-15
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REFERENCES •
i
DOE88 U.S. Department of Energy, Site Characterization Plan, Yucca Mountain Site,
Nevada Research and Development Area, DOE/RW-0199, December 1988.
DOE94a U.S. Department of Energy, Energy Information Administration, Nuclear Fuel
Data Form RW-859, December 1994. ;
DOE95a U.S. Department of Energy, Department of Energy Programmatic Spent Nuclear
Fuel Management and Idaho National Engineering Laboratory Environmental
Restoration and Waste Management Programs Final Environmental Impact
Statement, DOE/EIS-0203-F, April 1995. !
DOE95b U.S. Department of Energy, Office of Scientific and Technical Information,
Nuclear Reactors Built, Being Built, or Planned: 1994, DOE/OSTI-8200-R58,
August 1995. ;
DOE95c U.S. Department of Energy, Draft Waste Management Programmatic \
Environmental Impact Statement for Managing Treatment, Storage, and Disposal
of Radioactive and Hazardous Waste, DOE/EIS-0200-D, August 1995. ;
DOE95d U.S. Department of Energy, Integrated Data Base Report-1994: U.S. Spent
Nuclear Fuel and Radioactive Waste Inventories, Projections, and
Characteristics, Revision 10, September 1995.
DOE95e U.S. Department of Energy, Integrated Data Base Report-1994: U.S. Spent
Nuclear Fuel and Radioactive Waste Inventories, Projections, and
Characteristics, Revision 11, September 1995. \
DOE95f U.S. Department of Energy, Energy Information Administration, World Nuclear
Outlook 1995, DOE/EIA-0436(95), October 1995. \
DOE95g U.S. Department of Energy, Total System Performance Assessment - 1995: An
Evaluation of the Potential Yucca Mountain Repository, TRW Environmental
Safety Systems, Inc., BOOOOOOO-01717-2200-00136, Revision 01, November
1995. !
DOE96a U.S. Department of Energy, Integrated Data Base Report - 1995, Revision 12,
December 1996. I
DOE98 U.S. Department of Energy, Viability Assessment of a Repository at Yucca
Mountain, DOE/RW-0508, December 1998. i
5-16
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DOE99 U.S. Department of Energy, Draft Environmental Impact Statement for a
Geologic Repository for the Disposal of Spent Nuclear Fuel and High-Level
Radioactive Waste at Yucca Mountain, DOE/EIS-02050D, Appendix A, July
1999.
DOE98a U.S. Department of Energy, Report to Congress: Treatment and Immobilization of
Hanford Radioactive Waste, available on Hanford website at www.hanford.org.
EPA85 U.S. Environmental Protection Agency, Draft Environmental Impact Statement
for 40 CFR Part 191: Environmental Standards for Management and Disposal of
Spent Nuclear Fuel, High-Level and Transuranic Radioactive Wastes, EPA
520/1-85-023, August 1985.
NRC88 U.S. Nuclear Regulatory Commission, Code of Federal Regulations, Title 10, Part
60, Disposal ofHigh-Level Radioactive Wastes in Geologic Repositories, as
amended, October 1988.
NWP83 Nuclear Waste Policy Act of 1982, Public Law 97-425, January 7, 1983.
ROD86 Roddy, J.W., H.C. Claiborne, R.C. Ashline, P.J. Johnson, and B.T. Rhyne,
Physical and Decay Characteristics of Commercial LWR Spent Fuel, ORNL/TM-
9591/V2&R1, Oak Ridge National Laboratory, Oak Ridge, TN, 1986.
WCM95a Spent Fuel Goes to Idaho Following State-DOE-Navy Agreement, Weapons
Complex Monitor, October 26, 1995.
WCM95b OE Chooses Reprocessing for Selected Spent Fuel, Weapons Complex Monitor,
December 13, 1995.
5-17
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-------
CHAPTER 6
DOSE AND RISK ESTIMATION
6.1 INTRODUCTION
Ionizing radiation emitted by the radioactive decay of nuclides released into the environment
poses a risk of inducing excess cancers or heritable genetic effects in exposed humans. The curie
(Ci) and becquerel (Bq) are units used to measure the activity of radioactive material, i.e., the rate
atoms are giving off radiation or disintegrating. The curie is equal to 37 billion disintegrations
per second, while the becquerel is equal to one disintegration per second. Exposure can occur
through several "pathways," including inhalation, ingestion, or external irradiation by
radionuclides in the air or deposited on the ground (see Chapter 8).
The risk of a health effect being induced in an exposed individual by a given exposure is
calculated by first estimating the radiation dose to sensitive tissues in the individual, as a function
of age. Depending on the radionuclide in question, its chemical form, and the exposure pathway,
its distribution will vary within the body and with time, leading to a variation in radiation dose
with organ and across time. The dose per unit exposure is referred to as a "dose conversion
factor" (DCF). From the tissue-specific doses, the risks of a radiationr-induced cancer, cancer
death, or genetic effect are calculated using age- and organ-specific "risk factors." The dose
conversion and risk factors are generally calculated from models, as outlined below. The number
of excess cancers in a population is projected using a life-table calculation (BUN81, EPA94),
which corrects for competing causes of death.
6.2 DOSE ESTIMATION
The risk of inducing a cancer in a specific tissue or organ increases with the absorbed dose, i.e.,
the amount of ionization and excitation energy per unit mass deposited in that tissue or organ.
The risk of inducing a genetic effect increases with dose to the testes or ovaries. The absorbed
dose, D, is expressed in gray (Gy) or rad, where 1 Gy = 100 rad. The risk also depends on the
density of ionizations (the number of ionizations per unit path length) produced by the radiation.
The density of ionizations is directly related to the "linear energy transfer" (LET), which is a
measure of the amount of energy per unit path length deposited by a charged particle track in
traversing a material. When the density of ionizations is high, the radiation is referred to as
"high-LET"; conversely, "low-LET" radiation refers to that which is sparsely ionizing.
6-1
-------
Accordingly, a derived quantity called the effective dose is introduced, which is expressed in
units of sieverts (Sv) or rem. The effective dose in a tissue is given by QXD, where Q is a quality
factor (unitless) defined for a specific type of radiation. ;
Note that the absorbed dose is a physical quantity, but that the effective dose is a regulatory
concept determined in part by the choice of Q. Values for Q are assigned based on ;
radiobiological information on the relative biological effectiveness (RBE) of different types of
radiation. Since the RBEs of different types of radiation are not known precisely, the assignment
of Q rests heavily on the subjective judgments of experts on the ICRP. This document is
concerned only with: (1) low- LET radiation from beta particles, gamma rays, or energetic X-rays,
for which Q is taken to be unity and (2) high-LET alpha particles for which Q is taken to be 20
(ICR91, EPA94). In the case of low-LET radiation, 1 Sv = 1 Gy, and 1 rem = 1 rad. It follows
that 1 Sv= 100 rem.
For regulatory purposes, it is useful to introduce certain other measures of "dose." First, there is
the concept of the effective dose equivalent (EDE), which allows one to combine the dose
equivalents to different organs into a single quantity. In this connection, each target organ, /, is
i
assigned a weighting factor, w,, which roughly represents the estimated proportion of the risk
from a uniform, whole-body irradiation occurring in that particular organ. The effective dose
equivalent is then the weighted sum of doses to the individual organs (ICR77):
Second, in dealing with internally deposited radionuclides that remain in the body and irradiate
tissues for extended periods of time, the concept of "committed dose" is introduced (ICR77). For
example, the 50-yr committed effective dose equivalent (CEDE) from a given intake is the
calculated total EDE received over a 50-yr period following that intake. Finally, the annual
committed effective dose equivalent (annual CEDE) refers to the CEDE resulting from one year's
exposure or intake.
When the exposure is external, the dose calculation is a straightforward application of radiation
physics. The radiation doses to target organs in an idealized "reference man" are calculated from
the decay properties of the radionuclides and the well-understood interactions of radiation with
matter (ICR79, EPA89).
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For ingested or inhaled radionuclides, the dosimetry modeling is more complex. It is necessary
to incorporate biokinetic information to describe the distribution and retention of the
radionuclide (and any radioactive decay products) in the body as a function of time after intake.
The irradiation of target tissues by internally deposited radionuclides is further complicated by
the need to consider the cross irradiation of one tissue by radionuclides deposited in another
tissue. Dosimetry models for internally deposited radionuclides have been developed by the
International Commission on Radiological Protection (ICR79, ICR80, ICR81, ICR88). Dose
conversion factors for internal and external radionuclide exposures are tabulated in EPA's
Federal Guidance Reports Nos. 11 and 12, respectively (EPA88, EPA93). The individual
protection standard of 25 mrem/yr that was developed under old dosimetry methods and used in
the 40 CFR Part 191 standards promulgated in 1985 are essentially the same as the 15 mrem/yr
(CEDE) standard for 40 CFR Part 197.
6.3 CANCER RISK ESTIMATION
EPA's current model for estimating radiogenic cancer risks incorporates age- and organ-specific
risk coefficients for low-LET radiation based on data obtained from the Japanese atomic bomb
survivors up through 1985, supplemented by organ-specific data from other sources (e.g., breast
cancer induction in fluoroscopy patients).. For most cancer sites, EPA's methodology involves an
averaging of two sets of coefficients, reflecting two different ways of projecting risk from the
atomic bomb survivors to the U.S. population, which have significantly different baseline rates of
specific cancers (LAN91, EPA94, EPA99, EPA99a).
Aside from breast cancer, for which there is good epidemiological evidence that the dose
response is approximately linear and independent of fractionation (NAS90), it was assumed that
the risks at low doses and dose rates are reduced by a "dose, dose rate effectiveness factor"
(DDREF) of 2 compared to the acute high dose exposures experienced by the bomb survivors.
The value of 2 for the DDREF is consistent with ICRP recommendations (ICR91). For low dose
(or dose rate) conditions, the calculated risk of a premature cancer death attributable to uniform,
whole-body, low-LET irradiation is about 5.75xlO'2/Gy. Neglecting nonfatal skin cancers, which
are usually not serious, the corresponding incidence risk estimate is 8.5*10'2/Gy (EPA99a).
High-LET (alpha particle) risks are presumed to increase linearly with dose and to be
independent of dose rate. Except for leukemia and breast cancer, a relative biological
effectiveness (RBE) factor of 20 is adopted for estimating the risk of high-LET radiation relative
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to that for low-LET radiation at low dose or dose rate conditions. Again the RBE value of 20 is
consistent with the recommendations of the ICRP (ICR91). In view of epidemiological data on
people ingesting or being injected with alpha-emitting radionuclides that deposit in bone, an
effective RBE of 1 was adopted for leukemia; for breast cancer, the high-LET RBE of 110 is used
to be consistent with the DDREF of 1 adopted for this site.
i
The lifetime excess risks of cancer incidence and mortality, for constant exposure rates to over
100 different radionuclides, are tabulated in the Final Version of EPA's Federal Guidance Report
No. 13 (EPA99). The dosimetry models employed in deriving these risk estimates reflect new
ICRP recommendations and incorporate age-specific biological parameters (EPA99). I
6.4 GENETIC EFFECTS
Genetic effects of radiation exposure are defined as stable, heritable changes induced in the germ
cells (eggs or sperm) of exposed individuals, which are transmitted to and expressed only in their
progeny across future generations.
i
The genetic risk of radiation exposure is more subtle than the somatic risk since it does not affect
i
the persons exposed, but only their progeny. Somatic effects are expressed in the exposed
individual over the person's remaining lifetime, while about 30 subsequent generations !(nearly
1,000 years) are needed for near complete expression of genetic effects. Genetic risk is incurred
by fertile people when radiation damages the DNA of the germ cells. The damage, in the form of
a mutation or a chromosomal change, is transmitted to, and may be expressed in, a child
conceived after the radiation exposure. However, the damage may also be expressed in some
subsequent generation(s) or never.
Estimates of the genetic risk per generation are conventionally based on a 30-year reproductive
generation. That is, the median parental age for conception of children is defined as age 30
(approximately one-half the children are produced by persons less than age 30, the other half by
persons over age 30). Thus, the radiation dose accumulated from birth to age 30 is used to
estimate the genetic risks. A basic assumption in assessing radiation genetic risk is that, at low
doses and low dose-rates of low-LET radiation, there is a linear relationship between dose and
the probabi lity o f occurrence of the genetic effect. :
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In the EPA Background Information Document for Radionuclides (EPA84), direct and indirect
methods for obtaining genetic risk coefficients are described, and some recent estimates based ou
these methods are tabulated. Briefly, the direct method takes the frequency of mutation or
occurrence of a heritable defect per unit dose observed in animal studies and extrapolates to what
is expected for humans. These direct estimates are usually used for first generation effects
estimates.
The EPA assessment of risks of genetic effects includes both first generation estimates and total
genetic burden estimates. In developing risk coefficients for genetic effects, EPA has employed
traditional definitions of genetic effects and dose-response relationships. Although the newly
recognized mechanisms of genetic change listed above have future implications for genetic risk
assessment, there are no data upon which to base radiation risk coefficients for these kinds of
damage at this time.
In the NESHAPs Environmental Impact Statement (EPA89), the EPA estimated the low dose-
rate, low-LET doubling dose for genetic effects to be 1.0 Gy (100 rad). That is, 1.6 Gy per
reproductive generation (considered to be 30 years) would double the rate of occurrence of
congenital defects (a defect existing at birth but not hereditary). However, at that time, the
Agency indicated, based on limited human data, that the true doubling dose might be about three
times greater. There is still no consensus on this point.
Neel and Lewis reviewed untoward pregnancy outcomes (UPOs) in the Japanese A-bomb
survivors and compared them to mouse genetic effects data (NEE90a). The gametic doubling
dose for low dose-rate, low-LET radiation in man, in this case, would be 400 rad (NEE90a). In a
companion analysis of mouse genetic data, they estimated a gametic doubling dose in mice of
135 (16-400) rad. The gametic doubling dose for a study where only one sex was irradiated
provides an analog of the "conjoint" parental gonadal dose for comparison purposes. However,
for mice, they recommended a dose-rate factor of 3 for low dose-rate, low-LET radiation, so the
doubling dose would also be 400 rad in mice (NEE90a).
UNSCEAR reviewed the recommendations listed above and concluded that the doubling dose in
humans is most likely between 1.7 and 2.2 Sv (170 and 220 rad) for acute exposure to low-LET
radiation, but 4.0 Sv (400 rad) for chronic exposure (UNS93). However, the UNSCEAR report
also continued to estimate the hereditary effects of exposure to ionizing radiation using a
doubling dose of 1.0 Sv (100 rad), just as in earlier UNSCEAR reports (UNS86, UNS88).
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The EPA assumes a doubling dose of 100 rad (1 Sv) in this document, but again notes; that some
i
estimates of the doubling dose is about four times greater. The EPA estimate for equilibrium
effects is about twice that of recent estimates by BEIR V and UNSCEAR because EPA included
a value for equilibrium multifactorial effects where these others did not. The EPA estimates
incorporate a dose-rate factor of 3 for low-LET radiation as reported in the 1977 UNSCEAR
Report (UNS77).
The projected genetic effects attributable to a given population exposure depend on this
population dynamics of future generations. However, if a stationary population is assumed, the
I
number of effects can be derived from Table 6-1. The dose in the table is that received by
parents in the first 30 years of life, the assumed generation period. Since the average lifetime of
a person in the 1980 stationary population is about 75 years, 40 percent (30/75) of the Population
dose is considered to be genetically significant. Thus, to calculate genetic risk coefficients
comparable to the cancer risk coefficients cited above, the values in Table 6-1 should be
multiplied by 0.4. On this basis, eight serious heritable disorders are expected in the first
generation following a 104 person-Gy population exposure of low dose (or dose rate), }ow-LET
radiation, and 104 such effects would be expected over all generations. The number of serious
genetic effects projected over all generations is then about 20 percent of the excess fatal cancers
projected in the exposed population.
Table 6-1. Estimated Frequency of Genetic Disorders in a Birth Cohort Due to Exposure of
Each of the Parents to 0.01 Gy (1 rad) per Reproductive Generation (30 yr)
Radiation .
Low Dose Rate, Low-LET
High Dose Rate, Low-LET
High-LET
Serious Heritable Disorders !
(Cases per 10* Liveborn) ;
First Generation | All Generations
20
60
90
260 ',
780 ;
690 i
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6.5 DEVELOPMENTAL EFFECTS
6.5.1 In Utero Carcinogenesis
Studies of the effects of in utero X-ray exposures in the U.K. in the 1960s showed increased
childhood cancer as a sequela. The BEIR III Committee reviewed the data and estimated that
there was a risk of 25 x 10"4 excess fatal leukemias per year per Gy exposure (25 x 10"6 per rad) and
28x 10"4 excess fatal cancers of other types (28x 10"6 per rad) (NAS80). The risk starts at birth and
continues for 12 years for leukemias and 10 years for solid tumors (NAS80). Having reviewed
additional data, the BEIR V Committee estimated that the risk was "... about 200 to 250 excess
fatal cancer deaths xlO"4 per Gy [200 to 250 x 10'6 per rad] in the first 10 years of life...." It also
estimated one-half would be leukemias and one-quarter tumors of the nervous system (NAS90).
UNSCEAR estimated a risk of leukemia and solid tumors expressed during the first 10 years of
life of 2X1Q-4 per rad (UNS86). The NRPB estimated a cancer risk of 2.5xlO'4 cases of leukemia
and 3.5x 1Q-4 cases of solid tumors per rad of in utero exposure (STA88). The NRPB in 1993
retained the same cancer risk estimates but concluded about one-half the cases would be fatal and
they would be expressed in the first 15 years of life (NRP93). However, the NRPB also
estimated the lifetime risk would be four times greater than that of the first 15 years.(NRP93).
6.5.2 Brain Teratology
The ICRP published an excellent review of the biology and the possible mechanisms of
occurrence of radiation-induced brain damage in utero (ICR86). ICRP estimates: (1) for
exposures from the 8th through the 15th week after conception, the risk of severe mental
retardation is 4x10"' per Gy (4xlO'3 per rad), with a confidence interval of 2.5x10"' to 5.5x10"'
per Gy (2.5 xlO"3 to 5.5xlO"3 per rad) and (2) for exposures from the 16th through the 25th week
after conception, the risk of severe mental retardation is 1 x 10"' per Gy (1 x 10~3 per rad).
However, a threshold below 50 rad could not be excluded (ICR86).
Effects other than mental retardation and microcephaly have been noted in the Japanese A-bomb
survivors. Schull et al. (SCH88) reported that in individuals exposed prenatally between weeks 8
and 25 of gestation there is a progressive shift downward in IQ score with increasing exposure
and that the most sensitive group is between 8 and 15 weeks gestational age at time of exposure.
The BEIR V Committee estimated a 30 point loss in IQ per Gy exposure (0.3 points per rad)
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consistent with a linear nonthreshold relationship (NAS90). However, even if the effect is linear-
nonthreshold, the response would be too small to be detectable at environmental exposure levels.
Much the same pattern was reported for average school performance, especially in the earliest
years of schooling (OTA88). Finally, a linear-nonthreshold relationship between exposure and
incidence of unprovoked seizures in later life has been found to be consistent with the data for
individuals exposed between 8 and 15 weeks gestational age (DUN88).
In 1986, the United Nations Scientific Committee on the Effects of Atomic Radiatio^ also
reviewed the question of mental retardation as a part of the overall review of the biological
effects of prenatal radiation exposure (UNS86). UNSCEAR, like the ICRP, concluded there was
a risk of severe mental retardation of 4x 10"3 per rad over the period of 8 to 15 weeks after
conception and of 1x 10"3 per rad over the period 16 to 25 weeks after conception (UNS86).
The question of a threshold for central nervous system effects, particularly for the 8 to 15 week
period of gestation, is unresolved. Apparent thresholds in the human data may merely reflect the
statistical uncertainty due to the small number of cases. If, as has been suggested, the effects are
due to improper synaptogenesis in the brain (temporal or spatial) (ICR86, OTA87), itj should be
noted that significant prolongation of cell cycle in matrix cells of the developing telencephalon in
mice (exposed on day 13 of gestation) has been reported following exposures as low as 10 R
r
(KAM78). Exposure of mice to 1 R on day 13 of gestation resulted in an increase in eye and
i
brain abnormalities, but the increase was not statistically significant (MIC78). '
6.5.3 Other Effects of Prenatal Irradiation • -
UNSCEAR estimated: (1) a pre-implantation loss of 1><10"2 per rad during the first two weeks
after conception and (2) a malformation risk of 5><10"3 per rad during weeks 2 to 8 after
conception (UNS86). I
For many of the teratologic effects observed, no threshold has been demonstrated. If a
teratogenic effect of radiation is due to cell-killing effects, then a threshold for that effect is
probable. While early studies of radiation as a teratogen used high exposures and probably
induced effects through cell killing, cell killing may not be required. Patrick cites Zw^lling as
follows: ":.. developmental anomalies appear to be caused by 'failure of proper tissue; interaction
to occur'" (PAT78, ZWI63). For example, a somatic mutation in a single cell, perhaps through
6-8
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clonal expansion, could cause improper tissue interaction with no loss of cells; or, killing a single
cell could cause release of a toxicant that causes an improper local interaction (RUS54, WEI54).
. Jacobsen exposed pregnant mice to 0, 5R, 20R, or 100R on day 8 of gestation and scored skeletal
abnormalities on day 19. He interpreted the dose-effect curve as linear or nearly so and saw no
evidence of a threshold for the types of damage studied (JAC70). He stated: "The observations
made, and in particular that concerning the apparent absence of a threshold dose, indicate that it
is not justified to assume that irradiation with doses of 5 R and less is entirely without effect on
the human embryo in early developmental stages" (JAC70). In another study, exposure of mice
to 1 R on day 8 of gestation resulted in a significantly higher incidence of malformed and
retarded fetuses compared to controls (MIC78). A 1981 review of data on the effects of ionizing
radiation on the developing embryo/fetus reached essentially the same conclusions as Jacobsen
(HHS81). Given the large number of experimental animals that would be required, direct
evidence for a threshold below 5 rad will be difficult to provide.
6.5.4 Summary of Developmental Effects
EPA risk coefficients for estimating prenatal carcinogenic, teratologic, and nonstochastic effects
in man (see Table 6-2) are, with one exception, the same as those published in the 1989
NESHAPs BID (EPA89). The first entry in the corresponding table in the NESHAPs BED lists
"Fatal Cancer" as 6.0x 10'4. The entry should be for "Cancer Incidence." The fatal cancer risk is
about half as great, 3><10"4.
Table 6-2. Possible Effects of In Utero Radiation Exposure
Type of Risk to Conceptus
Cancer Incidence
Mental Retardation3 (exposure at 8-15 weeks)
Mental Retardation5 (exposure at 1 6-25 weeks)
Malformation11 (exposure at 2-8 weeks)
Pre-implantation Loss (exposure at 0-2 weeks)
Risk per Rad
6xlO-4
4x10°
IxlO'3
5x10-'
IxlO0
A threshold for mental retardation following exposure at 8-15 weeks of gestational age may depend on the
mechanism of action.
b A threshold is expected for mental retardation following exposure during the 16-25 week period of gestation and
for many types of malformations following exposures at^arly gestational age.
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REFERENCES
BUNS 1 Hunger, B., JR. Cook and M.K. Barrick, Life Table Methodology for Evaluating
Radiation Risk: An Application Based on Occupational Exposure, Health
Physics, 40:439-455,1981. I
DUN88 Dunn, K., H. Yoshimaru, M. Otake, J.F. Annegers and W.J. Schull, Prenatal
Exposure to Ionizing Radiation and Subsequent Development of Seizures,
Technical Report RERF TR 13-88, Radiation Effects Research Foundation,
Hiroshima, 1988. ;
EPA84 U.S. Environmental Protection Agency, Radionuclides, Background Information
Document for Final Rules, Volume I, Office of Radiation Programs, EPA Report
520/1-84-022-1, 1984.
EPA88 U.S. Environmental Protection Agency, Limiting Values of Radionuclide Intake •
and Air Concentration and Dose Conversion Factors for Inhalation, Submersion,
andlngestion, Federal Guidance Report No.l 1, EPA-520/1-88-020, 1989.
EPA89 U.S. Environmental Protection Agency, Risk Assessment Methodology, \.
Environmental Impact Statement for NESHAPs Radionuclides, Volume I,
Background Information Document, Office of Radiation Programs, EPA Report
520/l-89-005,Washington, DC 1989. . I
I
EPA93 U.S. Environmental Protection Agency, External Exposure to Radionuclides in
Air, Water, and Soil, Federal Guidance Report No. 12, EPA-402-R-93-081, 1993.
EPA94 U.S. Environmental Protection Agency, Estimating Radiogenic Cancer Risks,
Office of Radiation and Indoor Air, EPA Report 402-R-93-076, Washington, DC,
1994.
EPA99 U.S. Environmental Protection Agency, Health Risks From Low-Level
Environmental Exposure to Radionuclides. Federal Guidance Report NO. 13 -
Part 1-Final Version, EPA Report 402-R-98-001, Sept. 1999. |
EPA99a U.S. Environmental Protection Agency, Estimating Radiogenic Cancer Risks,
Addendum: Uncertainty Analysis, EPA Report 402-R-99-003, 1999. '
HHS81 Health and Human Services, Effects of Ionizing Radiation on the Developing
Embryo and Fetus, A Review, Bureau of Radiological Health, Public Health
Service, Food and Drug Administration, HHS Publication FDA 81-8170,
Rockville, MD, 1981. |
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ICR79 International Commission on Radiological Protection, Limits for Intakes of
Radionuclides by Workers, ICRP Publication No. 30, Part 1, Annals of the ICRP,
2(3/4), Pergamon Press, Oxford, 1979.
ICR80 International Commission on Radiological Protection, Limits for Intakes of
Radionuclides by Workers, ICRP Publication No. 30, Part 2, Annals of the ICRP,
4(3/4), Pergamon Press, Oxford, 1980.
ICR81 International Commission on Radiological Protection, Limits for Intakes of
Radionuclides by Workers, ICRP Publication No. 30, Part 3, Annals of the ICRP,
6(2/3), Pergamon Press, Oxford, 1981.
ICR86 International Commission on Radiological Protection, Developmental Effects of
Irradiation on the Brain of the Embryo and Fetus, ICRP Publication 49, Annals
of the ICRP, 16(4): 1-43, Pergamon Press, Oxford, 1986.
ICR88 International Commission on Radiological Protection, Limits for Intakes of
Radionuclides by Workers: an Addendum, ICRP Publication No. 30, Part 4,
Annals of the ICRP, 19(4), Pergamon Press, Oxford, 1988.
ICR91 International Commission on Radiological Protection, 1990 Recommendations of
the International Commission on Radiological Protection, ICRP Publication 60,
Annals of the ICRP, 21(1-3), Pergamon Press, Oxford, 1991.
JAC70 Jacobsen, L., Radiation Induced Fetal Damage, Adv. Teratol. 4, 95-124, 1970.
KAM78 Kameyama, Y., K. Hoshino and Y. Hayashi, Effects of Low-Dose X-Radiation on
the Matrix Cells in the Telencephalon of Mouse Embryos, pp. 228-236, in:
Developmental Toxicology of Energy-Related Pollutants CONF-771017, DOE
Symposium Series 47, Pacific Northwest Laboratories, Richland, WA, 1978.
LAN91 Land, C.E. and W.K. Sinclair, The Relative Contributions of Different Organ
Sites to the Total Cancer Mortality Associated with Low-Dose Radiation
Exposure, in: Risks Associated with Ionizing Radiations, Annals of the ICRP
22(1), Pergamon Press, Oxford, 1991.
MIC78 Michel, C. and H. Fritz-Niggli, Radiation-Induced Developmental Anomalies in
Mammalian Embryos by Low Doses and Interaction with Drugs, Stress and
Genetic Factors, pp. 397-408, in: Late Biological Effects of Ionizing Radiation
Vol. II, International Atomic Energy Agency, Vienna, 1978.
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NAS80 National Academy of Sciences/National Research Council, The Effects on
Populations of Exposure to Low Levels of Ionizing Radiation (BEIR III), National
Academy Press, Washington, DC, 1980.
NAS90 National Academy of Sciences/National Research Council, Health Effects of
Exposure to Low Levels of Ionizing Radiation (BEIR V), National Academy Press,
Washington, DC, 1990.
NEE90a Neel, J.V. and S.E. Lewis, The Comparative Radiation Genetics of Humans and
Mice, Annual Review of Genetics, 24, 327-362, 1990. [reprinted pp. 451 {486 in:
The Children of Atomic Bomb Survivors, A Genetic Study, J.V. Neel and W.J.
Schull, eds., National Academy Press, Washington, DC, 1990.] !
NEE90b Neel, J.V., W.J. Schull, A.A. Awa, C. Satoh, H. Kato, M. Otake and Y. '.
Yoshimoto, The Children of Parents Exposed to Atomic Bombs: Estimates of the
Genetic Doubling Dose of Radiation for Humans, Am. J. Hum. Geneticsj 46:
1053-1072,1990. [reprinted pp. 431-450 in: The Children of Atomic Bojtnb
Survivors, A Genetic Study, J.V. Neel and W.J. Schull, eds., National Adademy
Press, Washington, DC, 1991.] ! '
NRP93 National Radiological Protection Board of the UK, Estimates of Late Radiation
Risks to the UK Population, in: Documents of the NRPB, Volume 4, Number 4,
Chilton, England, 1993. '
OTA87 Otake, M., H. Yoshimaru and W.J. Schull. Severe Mental Retardation Among the
Prenatally Exposed Survivors of the Atomic Bombing of Hiroshima and \
Nagasaki: A Comparison of the T65DR and DS86 Dosimetry Systems, Technical
Report RJERF TR 16-87, Radiation Effects Research Foundation, Hiroshima,
1987. I
OTA88 Otake, M., W.J. Schull, Y. Fujikoshi, and H. Yoshimaru, Effect on School
Performance of Prenatal Exposure to Ionizing Radiation: A Comparison of the
T65DR and DS86 Dosimetry Systems, Technical Report RERF TR 2-88, i
Radiation Effects Research Foundation, Hiroshima, 1988. !
PAT78 , Patrick, C.H., Developmental Toxicology as Input to the Methodology for Human
Studies of Energy-Related Pollutants, pp. 425-440, in: Developmental Toxicology
of Energy-Related Pollutants, CONF-771017, DOE Symposium Series 47, Pacific
Northwest Laboratories, Richland, WA, 1978. '.,
RUS54 Russell, L.B. and W.L. Russell, An Analysis of the Changing Radiation Response
of the Developing Mouse Embryo, pp. 103-149, in: Symposium on Effects of
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Radiation and Other Deleterious Agents on Embryonic Development, J. Cell.
Comp. Physiol., 43, Supplement 1, May 1954.
SCH88 Schull, W.J., M. Otaki, and H. Yoshimaru, Effects on Intelligence Test Score of
Prenatal Exposure to Ionizing Radiation in Hiroshima and Nagasaki: A
Comparison of the T65DR and DS86 Dosimetry Systems, Technical Report RERF
TR 3-88, Radiation Effects Research Foundation, Hiroshima, 1988.
STA88 Slather, J.W., C.R. Muirhead, A.A. Edwards, J.D. Harrison, D.C. Lloyd, and N.R.
Wood, Health Effects Models Developed, in: 1988 UNSCEAR Report, NRPB-
R2'26, National Radiation Protection Board, Chilton, England, 1988.
UNS77 United Nations Scientific Committee on the Effects of Atomic Radiation, Sources
and Effects of Ionizing Radiation, Report to the General Assembly, with Annexes,
Sales No. E.77.IX. 1., United Nations, New York, 1977.
UNS86 United Nations Scientific Committee on the Effects of Atomic Radiation, Genetic
and Somatic Effects of Ionizing Radiation, 1986 Report to the General Assembly,
Sales No. E.86.IX.9., United Nations, New York, 1986.
UNS88 United Nations Scientific Committee on the Effects of Atomic Radiation,
Sources, Effects and Risks of Ionizing Radiation, 1988 Report to the General
Assembly, Sales No. E.88.IX.7., United Nations, New York, 1988.
UNS93 United Nations Scientific Committee on the Effects of Atomic Radiation, Sources
and Effects of Ionizing Radiation, 1993 Report to the General Assembly, Sales
No. E.94.IX.2., United Nations, New York, 1993.
WEI54 Weiss, P., Summarizing Remarks, pp. 329-331, in: Symposium on Effects of
Radiation and Other Deleterious Agents on Embryonic Development, Journal of
Cellular Comparative Physiology, 43, Supplement 1, May 1954.
ZWI63 Zwilling, E., Cell Differentiation and Embryagenesis, pp.75-90, in: Birth Defects.
M. Fishbein, ed., J.B. Lippincott Co., Philadelphia, 1963 (cited by Patrick in
PAT78).
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CHAPTER 7
CURRENT INFORMATION CONCERNING A POTENTIAL WASTE REPOSITORY AT
YUCCA MOUNTAIN
7.1 PRINCIPAL FEATURES OF THE NATURAL ENVIRONMENT
This section describes the principal features of the natural environment at Yucca Mountain and
the surrounding area. This information is based primarily on the site characterization work of the
Department of Energy (DOE). Particular emphasis is given to those aspects of the geology,
mineralogy, structure, hydrology, and climate of the site that are most likely to affect the
performance of a high-level waste repository. The glossary of technical terms at the end of this
BID should be helpful to the reader.
7.1.1 Geologic Features
A description of the important features of Yucca Mountain and the surrounding area provides a
picture of the geologic setting that serves as the context for understanding the repository design.
Important aspects of the geology around the site, such as the presence of faults, seismicity, and
the nature and distribution of rock types, are discussed.
7.1.1.1 Location and Principal Physical Features of the Site (Adapted from DOE95a)
The Yucca Mountain site is located in Nye County, Nevada approximately 150 kilometers (km)
northwest of Las Vegas, Nevada (Figure 7-1). The site is at the southwestern boundaries of the
Nevada Test Site and the adjoining Nellis Air Force Base and about 50 km east of Death Valley
National Monument. The Yucca Mountain Region includes the southern Great Basin in southern
Nevada and an adjacent area in California (Figure 7-2). The Great Basin, which is in the
northern portion of the Basin and Range physiographic province, is bounded geologically by the
margins of the Colorado Plateau to the east and southeast, by the Sierra Nevada and Transverse
Ranges to the west and south, and by the Snake River Plain and flood basalts of the Columbia
Plateau to the north. Typical Great Basin topography consists of north-south mountain ranges
separating narrow structural valleys with internal drainages. The Colorado River, flowing along
the margin of the Colorado Plateau and topographically isolated from Yucca Mountain, provides
the only external drainage. Yucca Mountain is situated in the southern section of the Great
7-1
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Basin, in the Southwest Nevada Volcanic Field (SNVF). This area is bounded on the south by
the Death Valley region and the Mojave Desert of California. Yucca Mountain is a narrow ridge
which trends north-south and extends approximately 20 km from the southern margin pf the
Timber Mountain caldera complex. The area is mapped on the following U.S. Geological Survey
7.5-minute topographic quadrangles: Amargosa Valley, Big Dune, Busted Butte, Crater Flat,
East of Brady Mountain, and Pinnacles Ridge (formerly Topopah Spring NW). '
Figure 7-1. Location of Yucca Mountain (DOE94a)
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Figure 7-2. Boundaries and Larger Subdivisions of the Basin and Range Physiographic
Province. Province boundary is indicated by heavy solid line (HUN74)
Yucca Mountain is an irregularly shaped upland, six to 10 km wide and about 40 km long.
Uplands in the Yucca Mountain area are composed of ridge crests, valley bottoms, and
intervening hill slopes (DOES 8) with dominantly north-trending echelon ridges and valleys
controlled by high-angled faults. The fault blocks, composed mostly of welded fine-grained
volcanic rocks, are tilted eastward. As a result, the fault-bounded west-facing slopes are
generally high, steep, and straight, whereas the east-facing slopes are more gentle and usually
deeply dissected. Except where protected by a resistant rock layer capping the lip slopes, the
ridge crests are mostly angular and eroded. Valleys range from shallow, straight, steeply sloping
gullies and ravines to relatively steep, bifurcating, gently sloping valleys and canyons. Hill
slopes are typically narrow and moderately steep near the crest, with progressively gentler slopes
toward the valley floor. The crest elevation of Yucca Mountain ranges between 1,500 and 1,930
meters (m) above sea level. The summit is about 650 m above the floors of adjacent washes in
Crater and Jackass Flats.
7-3
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The main drainage system for the Yucca Mountain area, including the Timber Mountain area, the
Calico Hills, and the mesas lying to the south of Timber Mountain, is in the Amargosa Valley.
This drainage, east of Beatty, Nevada, carries runoff from the region south through the Tecopa
basin into the southern part of Death Valley. The Amargosa Valley carries significant! runoff
only after extraordinarily heavy precipitation. There are no perennial streams or natural bodies of
surface water on or adjacent to the Yucca Mountain. The major drainages, Solitario Canyon on
the west, Forty Mile Wash on the east, and tributary drainages are primarily on the east flank of
the mountain and flow only briefly immediately after rainstorms (Figure 7-3).
Bedrock exposures are common at higher elevations in the Yucca Mountain Region. Many of
the hill slopes have a discontinuous veneer of blocky talus and wedges of colluvium cover the
lower hill slopes. The rates of erosion in the Yucca Mountain area are lower than in similar arid
areas in the southwestern U.S. and other parts of the world. Conditions contributing to these low
erosion rates include existence of fine-grained volcanic rocks which are relatively erosion-
resistant, insufficient runoff during interpluvial periods to remove hillslope colluviumj and
topography that has not been significantly affected by Quaternary tectonic activity (WHI93).
Regional erosion projections over 10,000 years are less than one meter of down cuttinjg in •
canyons above the potential repository block, and less than 0.02 m of slope retreat (DOE95a).
i
7.1.1.2 Geologic History of the Region (Adapted from DOE95a) ;
The physiography and geomorphic features in the Yucca Mountain area influence the
characteristics of the surface water system, and to some extent, the ground water system as well.
[
The flow of water into, within, and around a repository at Yucca Mountain would directly affect
its ability to contain the waste over time. The composition and chemical behavior of ground
water at Yucca Mountain will be affected by the type, size, and abundances of primary and
secondary mineral phases in the contacting rock formations. Furthermore, the geologic processes
and events important to repository performance and design can only be understood wijhin the
broader context of the geologic history of the region. Current and future geologic processes and
events are a direct product of the area's geologic history; projecting their effect on repository
performance requires an understanding of causes, frequencies, durations, and magnitudes over
time. For example, projecting the potential frequency and magnitude of earthquakes is based on
the historical record of past seismic activity. This information has been developed from records
of past seismicity and geologic studies on the effects of faulting (displacement of strata across
faults, topographic features, etc.) in the vicinity of the site.
7-4 :
-------
Figure 7-3. Physiographic Features in the Yucca Mountain Site Area (DOE88)
7-5
-------
In general terms, the Yucca Mountain Region is characterized by a thick section of Ptecambrian
and Paleozoic sedimentary rocks overlain by a sequence of Tertiary silicic volcanic rocks (see
Figure 7-4). The older rocks have been folded and faulted by a compressional tectonic process
and the entire stratigraphic section subsequently deformed by extensional basin-and-range
tectonics. Uplifted ranges, such as Yucca Mountain, are separated by basins partially! filled
alluvial deposits.
A basement complex of older Precambrian metamorphic and younger Precambrian igneous rocks
is presumed to underlie the area. The basement rocks are overlain by a westward-thickening
accumulation of shallow marine late Precambrian and early Cambrian marine sediments,
quartzite, siltstone, shale, and carbonate rocks. These deposits are interpreted as a rifted
continental margin miogeosyncline, shown in Figure 7-5, formed seaward of the highlands area.
These rocks are locally fossiliferous. Deposition that continued through the Devonian Period is
represented by carbonate and shale with interbedded quartzite and sandstone, thickening from up
to 500 meters in western Utah to at least 6,100 meters in central Nevada. ' !
In late Devonian and early Mississippian time, the Antler Orogeny, a mountain-building event,
formed a north-northeast trending highland area adjacent to the Roberts Mountains Thrust. Large
volumes of sediments eroded from the highlands into a foreland basin in the eastern half of the
Great Basin, forming thick flysch17 deposits adjacent to the highlands and shallow-water shelf
carbonates to the east (Figure 7-6). Erosion of the highlands and deposition into the basin
continued through the Permian Period, decreasing as the mountain-building waned. In Mesozoic
and early Cenozoic time, these rocks were folded and displaced along thrust faults with extensive
fracturing of the brittle rocks in the upper thrust plates. This faulting was accompanied by
intrusion of granitic stocks, uplift, and erosion of the land surface (DUD90). :
Middle and late Cenozoic crustal uplifting and extension in the region occurred over an area
1,500 km long and 500 to 1,000 km wide. The stretching, estimated at 10 to 50 percent of the
original width and locally as great as 100 percent, resulted in northerly trending faults with
sliding and tilting of large crustal blocks, forming the characteristic structure and topography of
the Great Basin. :•
17 Flysch deposits are typified by the widespread sandstones, marls, shales, and clays exemplified by deposits
occurring at the northern and southern borders of the Alps. ;
7-6
-------
'HYGROGEOLOGIC
UNITS
PERIOD
GEOLOGIC FORUfTlOHS
LAVA FLOW &
VALLEY FILL
WELDED - TUFF
AQUIFERS
TUFF & LAVA
FLOW
AOUITARDS
rtasrocatc BJGMJ.M-WVUU.
HJOCEHE BASMT
THIRSTY omrott TUFF
O/SHTS Or XMUUESA * SKULL I
TtuacH uauHTMH Turr
PAHO-BRUSH TUff
• fanarioii
PiMflZSTOHCEHE
AND yOO«S£ft
(0-5 Ua)
UPPER
CARBONATE
AQUIFER
UPPER
CLASTIC
AQUITARD
LOWER
CARBONATE .
AQUIFER
LOWER
CLASTIC
AQUITARD
BEirCD AMCC TUFF
crural eur \
mats or poors SFMHO
HOHSC spfun romunai
rirm'tH UUCSTOHE
OUGOCEHE
(Z4-ST UO)
OEsms care UUESTOMC
MB/MM foniuraii
tone m-H-OataurE » flooenrs urH.Fonvxran
Fir sfKHCS axautrE
cuftEKA cuttirzirE
4HTEWrC VMier UUESTOHE
H*einc ronutnoH .
aooomi UU£STOHE
HOftH fOUfOVH
POBOHlP GROUP
BOHAZ* ana rmiutrioii
CMAVM fOfOlfTIOH
• auAarzire
VOOO CfHTOH FOfUlfraH
srimtHG autrtrztTE
JOHHN1E FOfaifflOH
HOOHOtr oatatarf
pMtauur enour
OMEOUS -UETAHOHffOC BASEItem- COUPl£X
Figure 7-4. Generalized Regional Stratigraphic Column Showing Geologic Formations and
Hydrological Units in the Nevada Test Site Area (Modified from DOE95a). The
repository host rock at Yucca Mountain is in the Tertiary age Paint Brush Tuff.
7-7 '
-------
109
115°
109°
' EASTERN LIMIT OF LOWER PALEOZOIC EUGEOCLINAL ROCKS
' UPPERMOST PROTEROZOIC AND LOWER CAMBRIAN UIOGEOCL1NAL ROCKS
•• MIDDLE CAMBRIAN THROUGH UPPER DEVONIAN UIOGEOCLINAL ROCKS
' |
Figure 7-5. Late Precambrian Through Mid-Paleozoic Paleography of the Great Basin
•(Modified from DOE95a) ;
7-?
-------
121°
109
115°
Figure 7-6. Late Devonian and Mississippian Paleogeography of the Great Basin
(Modified from DOE95a)
7-9
-------
Accompanying these crustal adjustments, volcanic eruptions in the vicinity of Yucca Mountain
formed a series of calderas and deposited numerous thick beds of pyroclastics, tuff, and lava,
aggregating up to three km in thickness near Yucca Mountain. The major episodes of silicic
volcanism ceased about 7.5 million years ago (mega annum; Ma); however, relatively few
basaltic eruptive centers formed in the basins adjacent to Yucca Mountain perhaps as recently as
4,000 years ago, with most of the local basaltic eruptive centers being formed over 75,000 years
ago. |
7.1.1.3 Stratigraphy of the Yucca Mountain Area (Adapted from DOE95a) j
An understanding of the stratigraphy of the rocks at Yucca Mountain and the surrounding area is
important to: (1) designing and constructing the repository, (2) assessing the potential of the
natural barrier to retard the movement of radionuclides from the repository, and (3) describing
the expected behavior of ground water movement through these rocks. For example, tr|e physical
properties of the rocks at the repository horizon determine the effects of heat generatedlby the
radioactive waste on the near-field environment in the postclosure time period. They can also
determine the speed at which-radionuclides can be transported through the repository. >
The stratigraphy of the southern Great Basin is highly varied, with formations ranging in age
from Precambrian to Holocene, that is, from over 500 million years old to 10,000 years' old.
These rocks, briefly described in Table 7-1, are divided into eight general groups based, on age,
lithology, and history. •
At Yucca Mountain, the stratigraphy is dominated by mid-Tertiary rocks of volcanic origin that
erupted from the southwestern Nevada volcanic field. The stratigraphic sequence can be divided
into four general categories based on similarities in lithology, age, and history of deposition or
emplacement: (1) pre-Cenozoic rocks, (2) mid-Tertiary pyroclastic rocks, (3) younger basalt, and
(4) late Tertiary to late Quaternary surficial deposits (Figure 7-7). These categories are; discussed
in the following sections. ;
7-10
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Table 7-1. Stratigraphy of the Southern Great Basin
Older Pirecambrian
Crystalline Rocks
Precambrian and
Lower Cambrian Rocks
Middle Cambrian
through Devonian
Mississippian through
Permian Sedimentary
Rocks
Mesozoic Rocks
Tertiary Sedimentary
Rocks
Tertiary and
Quaternary Igneous
Rocks
Tertiary and
Quaternary Surficial
Deposits
These include extensive exposures of older Precambrian schist and gneiss and younger
Precambrian igneous rocks in eastern Clark and southeastern Lincoln Counties. Outcrops of
Precambrian granite, pegmatite, amphibolite, and gneiss exist in southern Lincoln County.
Schist, gneiss, and gneissic quartz monzonite, possibly as young as late Proterozoic, are
exposed in the Bullfrog Hills and Trapman Hills of southern Nye County. .
Late Precambrian and early Cambrian strata include a westward-thickening prism of quartzite,
siitstone, shale, and carbonate interpreted as a rifted continental margin miogeosyncline. This
prism has been divided into two depositional systems in Nevada: an eastern quartzite and
siitstone system and a western siitstone, carbonate, and quartzite province.
Middle Cambrian through Devonian rocks exposed in the southern Great Basin consist of
carbonates and shales, with interbedded quartzite and sandstone with thicknesses from up to
500 m in western Utah to at least 6,100 m in central Nevada. Strata of middle Cambrian
through Devonian age comprise the Lower Carbonate Aquifer.
Thick flysch* deposits result from erosion of the north-northeast trending highland formed
during the Antler Orogeny in late Devonian and early Mississippian time. This sedimentation
continued through Permian time, declining as the orogeny waned.
Mesozoic sedimentary rocks, locally present only in Clark County, consi- of continental and
marine sandstone, siitstone, and limestone of the Triassic and Jurassic Aztec Sandstone, Chinle
Formation, and Moenkopi Formation. Approximately 30 separate Mesozoic to Tertiary granitic
plutons are exposed in Esmeralda County, west of Yucca Mountain. These range in size from
less than one km2 to the 1 ,000 km2 Inyo Batholith.
Tertiary sedimentary rocks, such as the Esmeralda and Horse Spring Formations, crop out
throughout the southern Great Basin. These consist of poorly to moderately consolidated
alluvial deposits and fresh water limestones in variable thicknesses of up to 1,000 m. They are
commonly. found interbedded with volcanic deposits.
The most prevalent Tertiary igneous rocks of the southern Great Basin are pyroclastic deposits
of rhyolitic to trachytic composition. Eruptions from four calderas at Yucca Mountain between
approximately seven and 16 Ma produced a complex mixture of pyroclastic flow and fall
deposits, epiclastic deposits, and subsidiary lavas approximately 3050 m in thickness at Yucca
Mountain. This was followed by scattered, small-volume basaltic or bimodal basaltic-andesitic
lava and scoria eruptions.
Late Tertiary to Quaternary surficial deposits occur throughout the region as unconsolidated
alluvial fan, pediment, and basin fill deposits of highly variable thickness and character.
* Deposits largely of sandy and calcareous shales.
7-11
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NEVADA
TEST
SITE
1 0123
KEY MAP
LEGEND
Miocene fel sic vo/canlcs
ftalnler Uesa
Paintbrush Croup
tuffocaous rocts
Polrtbrush Group
two roots
Calico Hills formation \
Crator rial Croup
Older and younger ur&s
D Quaternary/Tertiary
a
pitlaaxna bason
Pliocene bosoll
UIocene Patvte
Ussa Turr
Ulocere elastics
Uloccne basalt
Paleozoic cartanatiu
Approximate :
culdera margin'.
Figure 7-7. Simplified Geologic Map Showing the Distribution of Major Lithostratigraphic
Units in the Yucca Mountain Area (Modified from DOE95a).
7-12
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Pre-Cenozoic Rocks
Pre-Cenozoic rocks, believed to consist primarily of Paleozoic sedimentary strata, underlie the
volcanic rocks at Yucca Mountain. Little detailed information is available as to their thickness,
lithology, and contact with overlying stratigraphic units. Exposures of highly deformed
Paleozoic rocks occur at scattered localities in the vicinity of Yucca Mountain, including the
Calico Hills to the east, Bare Mountain to the west, and Striped Hill to the south. Carbonate
rocks have been detected at a depth of 1,244-1,807 m in a borehole two km east of Yucca
Mountain (DOE95a).
In the Calico Hills, exposures of carbonate rocks occur in the upper plate of a gently dipping
thrust fault over a black shale sequence containing minor amounts of siltstone, sandstone,
conglomerate, and limestone. These strata are locally highly folded, making correlation with
stratigraphic units elsewhere in the region uncertain.
At Bare Mountain, there is a varied sequence of pre-Cenozoic sedimentary and meta-sedimentary
rocks, totaling about 6,650 m in thickness and ranging from Precambrian to Mississippian in age.
Fourteen Paleozoic and two Proterozoic formations are represented. Dolomite and limestone
dominate, with minor stratigraphic units of clastic rocks (quartzite, sandstone, and siltstone).
Paleozoic rocks found at a depth of 1,244 to 1,807 m in a borehole two km east of Yucca
Mountain are almost entirely dolomites and have been identified as related to the Lone Mountain
Dolomite and the Roberts Mountains Formation. Seismic reflection data are inconclusive as to
the thickness and extent of pre-Cenozoic rocks underlying Yucca Mountain, but the thickness is
believed to be substantial.
Mid-Tertiary Pyroclastic Rocks
These rocks, resting unconformably on older pre-Cenozoic rocks, compose the portion of Yucca
Mountain most important to the design and performance of the repository because they are the
host rocks for the repository and define the pathways for ground water flow into and out of the
repository. Volcanic rocks ranging in age from about 11.4 to 15.2 Ma form the bulk of the
volcanic sequence, including the host rock of the potential repository, known as the Topopah
Spring tuff (Figure 7-8). The volcanic sequence consists of welded and nonwelded silicic
pyroclastic flow, fallout tephra deposits, and volcanic breccias erupted from nearby calderas in
the southwestern Nevada volcanic field. Non-welded tuffs typically have large primary porosity.
7-13
-------
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However, the large porosity is poorly interconnected resulting in low permeability. The harder,
welded tuffs are commonly more highly fractured and, consequently, have significant bulk
permeability. The principal stratigraphic units are listed in Table 7-2, in order of increasing age
(adapted from DOE94a).
Table 7-2. Principal Stratigraphic Units
Unit
Younger Post-caldera Basalts
Older Post-caldera Basalts
Shoshone Rhyolite Lava
Timber Mountain Group
Ammonia Tanks Tuff
Rainier Mesa Tuff
Post-Tiva/pre-Ranier Rhyolites
Paintbrush Group
Tiva Canyon Tuff
Yucca Mountain Tuff
Pah Canyon Tuff
Topopah Spring Tuff
Calico -Hills Formation
Crater Flat Group
Prow Pass Tuff
Bullfrog Tuff
Tram Tuff
Dacite Lava and Flow Breccia
Lithic Ridge Tuff
OlHer Tnfft - Prp-T itViir RiHap
Age (Ma)
0.27-3.8{a)
8.5-10.5(a)
9
11.45
11.6
12.5
12.7
-
-
12.8
12.9
13.1
13.25
13.45
14.0
14-lfi
(a) Based on information from DOE95a to be discussed subsequently in Section 7.1.1.7. The age of
the older post-caldera basalts ranges from 10.4 to 6.3 Ma; for the younger post-caldera basalts, the
age ranges from 4.9 to 0.004 Ma.
Many of these formations, particularly those in the Prow Pass Tuff, Calico Hills Formation, and
the Paintbrush Group, are further subdivided into members or units. The formations are
summarized below, from oldest to youngest, with an emphasis on thickness, general composition
and minerals important to radionuclide retardation along potential ground water transport
pathways.
a. Pre-Lithic Ridge Volcanics. The oldest known volcanic rocks in the area were deposited
approximately 15 million years ago and are represented in site boreholes by 45 to 350 m
of bedded tuffaceous deposits, pyroclastic flow deposits, and quartz-latitic to rhyolitic
7-15
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lavas and flow breccia. Correlation of these rocks with other rocks in the area is; difficult
because of their heterogeneous character and varying degrees of alteration.
I
b. Lithic Ridge Tuff. This thick, massive pyroclastic flow deposit overlying the older tuffs
appears to represent several eruptive surges and ranges in thickness from 185 m jnorth of
the site to 304 m at the south end of the site. This unit is nonwelded to moderately
welded and has been extensively altered to smectites and zeolites. |
j
c. Dacitic Lava and Flow Breccia. Dacitic lava and flow breccia overlie the Lithic Ridge
Tuff in deep boreholes at the northern and western parts of Yucca Mountain but are
absent elsewhere. Observed thicknesses in boreholes range from 22 m to 249 ml Much
of the unit has been moderately to intensely altered to smectite clays and zeolites.
i
d. Crater Flat Group. This group, overlying dacitic lavas and flow breccias in the northern
part of Yucca Mountain and the Lithic Ridge Tuff in the southern part, includes three
rhyolitic, ash-flow-tuff sheets—the Tram, Bullfrog, and Prow Pass Tuffs, in ascending
order. The Crater Flat Group is distinguished from other pyroclastic units at Yucca
Mountain by the relative abundance of quartz and biotite phenocrysts. I
• Tram Tuff. The Tram Tuff appears to comprise at least 28 separate magmatic
pulses and includes two subunits distinguished on the basis of the relative
abundance of lithic fragments. The lower subunit is rich in these fragmehts
throughout, while the upper unit is poor in lithic clasts. The upper subunit, 126 to
171m thick, is partially welded and has a microcrystalline ground mass.
There are six to 22 m of ash-fall and reworked tuff, primarily comprising; zeolitic
pumice clasts, between the Tram and the overlying Bullfrog Tuff. . i
Bullfrog Tuff. The Bullfrog Tuff is 68 to 187 m thick, consisting mostly of
pyroclastic flow deposits with thin-bedded tuffaceous deposits. North of borehole
USW G-4 (see Figure 7-8), this tuff consists of a moderately to densely vyelded
core enclosed by nonwelded to partially welded zones. To the south, the;tuff is
composed of two welded zones separated by a one-meter-thick bed of welded
fallout tephra.
• Prow Pass Tuff. The Prow Pass Tuff is a sequence of variably welded j
pyroclastic deposits that erupted from an unidentified source between 13:0 and
13.2 Ma. The formation, 90 to 165 m thick across the repository area, consists of
four pyroclastic units overlying a variable sequence of bedded tuffs. Thejse units,
designated Unit 1 through 4 by decreasing age, are characterized by '
orthopyroxene pseudomorphs and the abundance of siltstone and mudstorie lithic
clasts. Unit contacts are defined by fallout tephra horizons and abrupt changes in
sizes and amounts of pumice and lithic clasts. \
7-16
-------
e.
A bedded tuff unit at the base of the Prow Pass Tuff consists of unwelded, altered
tuffaceous deposits with a total thickness ranging from less than one meter to 11
m in boreholes.
Unit 1, a pumiceous pyroclastic flow deposit with an aggregate thickness of 25 to
70 m in cored boreholes, consists of three subunits separated on the basis of their
lithic clast content.
Unit 2 consists of nonwelded to partially welded lithic-rich pyroclastic flow
deposits with an aggregate thickness of three meters to 34 m in cored sections.
The unit has not been subdivided since distinguishing characteristics are lacking;
however, locally preserved ash horizons and abrupt changes in the amount and
size of pumice and lithic clasts suggest at least three flow deposits.
Unit 3 consists of 40 m to nearly 80 m of multiple welded pyroclastic flow
deposits, either separated by thin fallout tephra horizons or defined by abrupt
changes in the amount and size of pumice and lithic clasts. Two of three-flow
deposits have been identified in most core holes but have not been correlated.
Unit 4 is distinguished by comparatively abundant pseudomorphic pyroxene in
pumice clasts and rock matrix and by a comparatively low ratio of flesic to mafic
phenocryst minerals. This unit includes three irregularly distributed subunits.
The aggregate thickness in cored sections ranges from about 4 m to as much as
20.5m.
Calico Hills Formation. The Calico Hills Formation, a series of rhyolite tuffs and lavas,
includes five pyroclastic units overlying a bedded tuff unit and a local basal sandstone
unit in the Yucca Mountain area. The formation thins southward across the site area,
declining from about 290 m in the north to 43 m in the south. Basal beds of the Calico
Hills Formation include two units. One unit consists of a nine- to 39-meter-thick bedded
tuff unit containing coarse-grained fallout, primary and reworked pyroclastic-flow
deposits, and fallout-tephra deposits. The other unit consists of a 0- to 5.5-meter-thick
volcaniclastic sandstone unit with abundant lithic clasts and swarms of altered (to clay
minerals) pumice clasts, interbedded with rare pyroclastic-flow deposits.
The pyroclastic units are composed of one or more pyroclastic-flow deposits separated by
pumice- and lithic-fallout tephra deposits included with the unit lying above. Five units,
designated Units 1 through 5 by decreasing age, can be distinguished on the basis of
textural characteristics (percentages of various clastic material). In the northern part of
Yucca Mountain (below the proposed repository horizon) the formation is high in
zeolites, which compose 60 to 80 percent of the rock. In the southern portion of Yucca
Mountain, the rock remains vitric.
Unit 1 is a nonwelded, lithic rich, pyroclastic-flow deposit ranging from 0 to 58 m thick
in cored sections. Pumice clasts constitute 10 to 15 percent of the unit and lithic clasts
7-17
-------
f.
I
increase from three to seven percent at the top to 15 to 20 percent at the base; phenocrysts
compose seven to 12 percent of the rock. >
Unit 2, 0 to 54 m thick, is a nonwelded, pumiceous, pyroclastic-flow deposit composed of
20 to 40 percent pumice clasts and up to five percent lithic clasts. Fallout deposits at the
base are ash-rich, have a porcelaneous appearance, and are less than one meter thick.
Unit 3 is a nonwelded lithic-rich pyroclastic flow deposit 22 m to 100 m thick in cored
sections. The unit is generally composed of 10 to 40 percent pumice clasts and five to
10 percent lithic clasts. •
Unit 4 is a 0 to 57 m thick nonwelded, pumiceous pyroclastic flow deposit, with pumice
clasts and lithic clasts constituting 10 to 30 percent and one to five percent, respectively.
Thinly bedded ash-fall deposits, reworked pyroclastic-flow tuffs, and tuffaceous
sandstone form a thin basal subunit. ;
Unit 5 is a nonwelded to partially-welded pyroclastic-flow deposit ranging from 0 to 20 m
thick in cored sections. The unit is characterized by a bimodal distribution of pumice
clast sizes—larger, slightly flattened clasts of 20 to 60 mm and smaller equidimensional
clasts of two to 12 mm. The unit is composed of 20 to 30 percent pumice clasts and two
to five percent lithic clasts. \
Paintbrush Group. This group—one of the most widespread and voluminous caldera-
related assemblages in the southwestern Nevada volcanic field—consists of prirnary
pyroclastic flow and fallout tephra deposits, lava flows, and secondary volcanicjastic
deposits from eolian and fluvial processes. !
Eruptive centers for the Topopah Spring and Pah Canyon Tuffs are uncertain, but the
Claim Canyon caldera (see Figure 7-7) is identified as the source of the Tiva Canyon and
perhaps the Yucca Mountain Tuffs. '
• The Topopah Spring Tuff (Figure 7-8) is the host rock for the proposed Yucca
Mountain repository. The tuff has a maximum thickness of about 350 m in the
vicinity of Yucca Mountain. The unit is divided into two members—an^upper
crystal-rich member and a lower crystal-poor member—each of which is
subdivided based on variations in crystal content, phenocryst assemblage, pumice
composition, distribution of welding and crystallization zones, depositiqnal
features, and fracture characteristics.
!
The upper, crystal-rich member is characterized by greater than 10 percent
phenocrysts, with a basal transition zone where the percentage increases; from five
to 10 percent. The member is divided into vitric, nonlithophysal, and local
lithophysal zones.
7-18
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g-
The lower, crystal-poor member is characterized by less than three percent
phenocrysts and is divided into devitrified rocks of the upper lithophysal, middle
nonlithophysal, and lower lithophysal zones and a vitric zone. Below the vitric
zone (the vitrophyre), concentrations of clay and zeolites increase significantly
from alteration of the volcanic glass.
The Pah Canyon Tuff, a simple cooling unit composed of multiple flow units,
reaches its maximum thickness of 70 m in the northern part of Yucca Mountain
and thins southward. This tuff varies from nonwelded to moderately-welded.
Throughout much of the area, vitric pumice clasts are preserved in a sintered or
lithified nondeformed matrix.
• The Yucca Mountain Tuff, a simple cooling unit in the Yucca Mountain area,
varies in thickness from 0 to 30 m. Generally nonwelded, the unit is
nonlithophysal throughout Yucca Mountain but contains lithophysae where
densely welded in northern Crater Flat.
The Tiva Canyon Tuff (Figure 7-8) is a large-volume, regionally extensive,
compositionally-zoned (from rhyolite to quartz latite) tuff sequence that forms
most of the exposed surface rocks exposed at Yucca Mountain. The tuff ranges in
thickness from 100 to 150 m. Separation into crystal-rich and crystal-poor
members and into zones within these members is based on similar criteria and
characteristics discussed above for the Topopah Spring Tuff.
Post-Tiva Canyon, pre-Rainier Mesa Tuffs. A sequence of pyroclastic flow and fallout
tephra deposits occurs between the Tiva Canyon Tuff and the Rainier Mesa Tuff in the
vicinity of Yucca Mountain. The sequence ranges from 0 to 61 m thick and is
intermediate in composition between Tiva Canyon and Rainier Mesa Tuffs.
Timber Mountain Group. This group includes all of the quartz-bearing pyroclastic flow
and fallout tephra deposits that erupted from the Timber Mountain caldera complex about
11.5 Ma (see Figure 7-7). The complex consists of two overlapping, resurgent
calderas—one formed by eruption of the Rainier Mesa Tuff and a younger, nested one
formed by eruption of the Ammonia Tanks Tuff.
The Rainier Mesa Tuff is one of the most widespread pyroclastic units of the
Yucca Mountain area. It is a compositionally-zoned unit consisting of high-silica
rhyolite tuff overlain by a considerably thinner quartz latite tuff restricted to the
vicinity of the Timber Mountain caldera. Exposed thicknesses along the west side
of the caldera are as great as 500 m. The formation is absent across much of
Yucca Mountain, but appears in down-thrown blocks of large faults in valleys on
either side. The tuff is nonwelded at the base, grading upward into partially- to
moderately-welded devitrified tuff.
7-19 '
-------
The Ammonia Tanks Tuff consists of welded to nonwelded rhyolite tuff with a
highly variable thickness of up to 215 m. It is absent across Yucca Mountain, but
is exposed in the southern part of Crater Flat. !
Hydrostratigraphy \
The formal geologic stratigraphy for those rocks near the repository horizon has been reorganized
into four major hydrostratigraphic units for ground water modeling and performance assessment.
The groupings are based primarily on the degree of welding of the tuffs. These units and their
relationship to formal geologic stratigraphy are as follows (descriptions taken from DOE95b):
Tiva Canyon welded (TCw) unit: Consists of the moderately- to densely-welded
zones of the Tiva Canyon geologic member. This unit is characterized by low
matrix porosity (~ 10 percent), low matrix saturated hydraulic conductivity (-10"
1 Ws), and high fracture density (10-20 fractures/m3). |
Paintbrush nonwelded (PTn) unit: Consists of the lower partially-welded to
nonwelded zones of the Tiva Canyon geologic member, partially-welded to
nonwelded Yucca Mountain and Pah Canyon members, the porous interlayers of
bedded tuffs, and the upper partially-welded to nonwelded part of the Top|opah
Spring member. This unit is characterized by high matrix porosity (-40 percent),
high matrix saturated hydraulic conductivity (~ 10"7 m/s), and low fracture' density
(-1 fracture/m3).
Topopah Springs welded (TSw) unit: Consists of the welded zones of the:
Topopah Spring member. This unit is characterized by low matrix porosity (-10
percent), low matrix saturated hydraulic conductivity (-10"7 m/s), and high
fracture density (8-40 fractures/m3). The basal vitrophyre of the Topopah; Spring
member (TSv) is generally identified as a subunit because of its lower porosity as
compared to the TSw unit. •
• Calico Hills nonwelded (CHn) unit: consisting of the moderately-welded to
nonwelded zones of the Topopah Spring member underlying the basal vitrophyre,
the partially-welded to nonwelded tuffs of the Calico Hills formation, and! other
partially-welded to nonwelded tuffs located below the Calico Hills formation (i.e.,
the Prow Pass, Bullfrog and Tram members of the Crater Flat Unit). Portions of
the lower Topopah Spring member are vitrified and zeolitic alteration appbars in
both the lower part of the Topopah Spring member and in the tuffaceous beds of
the Calico Hills. This leads to a further division of this unit into vitric (CHnv)
and zeolitic (CHnz) subunits. The fracture density (2-3 fractures/m3) is similar in
both zones, and the porosity in the vitric tuffs (-30 percent) is marginally higher
than that of the zeolitic tuffs. However, matrix saturated hydraulic conductivity of
7-20
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the CHnv subunit (~ 10"9 m/s) is roughly two orders of magnitude higher than that
of the CHnz subunit.
In some discussions of Yucca Mountain stratigraphy, the stratigraphic column is divided into
thermal/mechanical units, rather than the more formal geologic formations or the
hydrostratigraphic units (see, for example, Figure 6-7 in DOE94a). The boundaries between the
thermal/mechanical units tend to be defined by the interface between welded and non-welded
lithologies and the units are very similar to the hydrostratigraphic groupings.
Younger Basalt
The youngest volcanic rocks in the Yucca Mountain area are the basalts at Lathrop Wells, where
multiple eruptions occurred over a period of about 120,000 years with the latest event occurring
less than 10,000 years ago.
Surficial Deposits
Surficial deposits in the area reflect the effects of erosive processes and affect the surficial
recharge of water to the underlying rocks. Numerous Quaternary/Tertiary surficial deposits have
been defined in the Yucca Mountain area. These include alluvial, colluvial, and eolian deposits.
The alluvial deposits range in age from late Tertiary (probably late Miocene) to late Holocene
and generally consist of sandy gravel (granules to boulders), often with interbedded sands. These
deposits occur along the washes, drainage channels, and valley slopes. The colluvial deposits are
primarily of Quaternary age and generally consist of a thin mantle of angular gravels on slopes
and highlands.
Two deposits of eolian sand ramp are defined, both formed of massive to poorly-bedded sand
with five to 50 percent fine angular gravel. One deposit (late and middle Pleistocene) forms
partially-dissected aprons between gullies on lower hill slopes. The other deposit (Holocene and
late Pleistocene) forms undissected and poorly-exposed sand ramps along Forty Mile Wash.
Summary
The most important rocks affecting the design and performance of the proposed Yucca Mountain
repository are the sequence of Miocene volcanic rocks that overlie, underlie, and are the host
rocks for the repository. These silicic rocks consist of ash-flow and air-fall tuffs produced by
eruptions from the Timber Mountain-Oasis Valley caldera complex. Most of the exposed surface
7-21
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rock over the repository is the 100-150 m thick Tiva Canyon Tuff. Below this, is the Yucca
Mountain Tuff, which is largely nonwelded and up to 30 m thick. The Claim Canyon|caldera
segment lying to the east of the proposed repository site is a possible source for rocks Jin these
units. The repository horizon is in the Topopah Spring Tuff which has a maximum thickness of
350 m in the vicinity of Yucca Mountain. These units are all part of the Paintbrush Group.
i
Next, in descending sequence, is the Calico Hills Formation consisting of rhyolite tuffs and lavas
which, in turn, is underlain by the Prow Pass Tuff in the Crater Flat Group. The Prow, Pass Tuff
is 90 to 165 m thick under the potential repository location. The surface of the water table lies
near the base of this unit. Lower lying units, generally in the saturated zone, include the 68 to
187 m thick Bullfrog Tuff and the Tram Tuff. These two tuffs are separated by six to 22 m of
ash-fall and reworked tuff comprised mainly of zeolitic pumice clasts. ;
7.1.1.4 Major Fault Features of the Yucca Mountain Area (Adapted from DOE95a) ;
The faults present in the site area are important for several reasons. To avoid adverse effects of
fault movement, areas of active fault movement should be avoided when deciding on the location
of surface waste handling facilities for the repository, as well as when designing the underground
waste emplacements locations. The fractured rocks in fault zones can also act as preferential
pathways for ground water movement and radionuclide migration. Their location an^ hydrologic
properties are important for developing an understanding of the flow system and performing
quantitative calculations of ground water movement essential to assessing the repository's
j
performance. i
Faulting and the Structural Setting Around Yucca Mountain \
The location of faults, and the extent of recent movement along these faults, is important to the
location and design of surface facilities and the layout of the underground repository at the Yucca
Mountain site. Seismic conditions in the area show at least some degree of correlation with the
faults observed. Seismic activity could affect surface facilities of the repository. In addition, the
fractured rock zones typical of fault zones often serve as preferential pathways for the: movement
of ground water. Rapid flow of ground water along fractures in the site area has beenlobserved
and DOE's current layout of the repository has been designed to avoid emplacing wastes in areas
where the host rock is prominently fractured (e.g., the Ghost Dance Fault zone).
7-22'
-------
Yucca Mountain consists of a series of north-trending, eastwardly tilted structural blocks that
were segmented by west-dipping, high-angle normal faults during a period of major extensional
deformation. The site is situated near the southern end of the northwest trending Walker Lane
Belt., a zone of northwest-directed shear about 700 km long and 100 to 300 km wide. This Belt
absorbs part of the transform motion of the regional plates and the strain from the extension of
the Great Basin. It parallels the San Andreas fault and the Sierra Nevada Mountains and is
truncated on the south by the east-west Garlock fault (Figure 7-9).
SETTLES WELL FAULT
1932 CEDAR UTa EARTHQUAKE
TONOPAH
EXCELSIOR FAULT ZONE
r LAKE MEAO
••*••• rum r <:V«:TF
FAULT SYSTEM
Figure 7-9. The Walker Lane Belt and Major Associated Faults (DOE88)
Cenozoic deformation probably took place on preexisting structures and is characterized by
strike-slip faulting, regional folding, and large-scale extension (see, for example, STE90). The
current type of deformation in the Walker Lane Belt probably began about five million years ago
as an, overlap between the right-lateral shear caused by the North American and Pacific plates and
7-23
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the gravity-driven extension of the regional uplift in the Great Basin. In the modem
northwest-striking faults move with left-lateral strike-slip or oblique-slip along the
stress field,
planes.
fault
In the Walker Lane Belt, right angle-shear totaling 4.27 to 7.35 millimeters per year (mm/yr) is
distributed along three major faults: the Owens Valley, Panamint Valley-Hunter Mountain, and
Death Valley-Furnace Creek faults. This, along with lesser amounts of slip on other fault
systems to the east, correlates well with the approximate 10 mm/yr of slip estimated from field
measurements.
The major north-trending faults transecting or close to Yucca Mountain are, from west to east,
the Crater Flat, Windy Wash, Fatigue Wash, Solitario Canyon, Stagecoach Road, Ghost Dance,
Bow Ridge, Midway Valley, and Paintbrush Canyon faults (Figure 7-10). Bedrock has been
s
displaced downward and to the west along these faults, which show predominantly dip slip, with
varying amounts of left-oblique slip, along the faults. Estimates of bedrock displacement over
the past 12 million years range from less than 100 m to as much as 600 m, with the displacement
increasing southward along each fault. The faults are projected up to 25 kilometers, but surface
exposures can usually be traced only one kilometer or less. Dips of the fault planes are generally
70 to 75 degrees.
f
Several northwest-trending faults have been identified along valleys, the most prominent being
the Yucca Wash, Sever Wash, Pagany Wash, and Drill Hole Wash faults. A northwest-trending
shear zone, the Sundance Fault, crosses the potential repository site (Figure 7-11). These faults
are thought to be strike-slip faults, with nearly horizontal slickenside lineations and vertical
displacements generally less than five to 10 m. ;
E
Quaternary Faulting in the Yucca Mountain Area !
; [
Of particular concern for the Yucca Mountain site are faults considered to be Type I faults, as
I
classified by the U.S. Nuclear Regulatory Commission (NRC). Type I faults or fault zones are
those subject to displacement and are sufficiently long or located such that they may affect
repository design and/or performance. Evidence of movement during the Quaternary Period, (the
past 1.6 million years) is the primary criterion for identification of these faults.
7-24
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116* 37'30" W
116'30',OO" W
W22V30" W
KCESSIBLC EWIKONUENT SOW/JAW
nrEirriM. REPOsrrom SITE
Figure 7-10. Major North-Trending Faults in the Vicinity of Yucca Mountain (DOE95k)
7-25
-------
rt'irsa- w
Figure 7-11. Index Map of Faults at and near Yucca Mountain (Modified from DOE95k)
7-26
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Studies to identify and characterize faults that may be of concern to the Yucca Mountain facility
have focused on evaluating the potential Type I faults within 100 km of the site, as well as a few
major faults at greater distances. Some 82 known or suspected Quaternary faults and fault
rupture combinations have been identified within 100 km of the Yucca Mountain site (Figure 7-
12). DOE reports that 38 of these are capable of generating apeak acceleration of 0.1 g (the
force of gravity) or greater at the ground surface of the proposed repository site; these are
classified as relevant earthquake sources.18 An updated compilation of faults has been prepared
by the U.S. Geological Survey (USGS) which identifies 67 faults with demonstrable or
questionable evidence of Quaternary movement and the capability of accelerations of at least 0.1
g at an 84 percent confidence limit (WHI96). Significant known or suspected Quaternary faults
located within 20 km of the Yucca Mountain site are briefly described in Table 7-3.19 The more
distant major fault zones include: the Garlock Fault (125 kilometers south), the Owens Valley
Fault (140 kilometers west), the Stewart-Monte Cristo Valley Fault (200 kilometers northwest),
and the Dixie Valley Fault (see page 3.1 -8 et seq, DOE95a).
Several of the north-trending faults show evidence of activity during Quaternary time; the total
displacements on the most active of these is estimated to be less than 50 meters over the past 1.6
million years. Since the late Quaternary Period (<128,000 years), displacements have been as
much as six m but are more commonly.in the one to 2.5 m range. Recurrence intervals on the
faults showing movement in the Quaternary Period fall in the range of tens of thousands of years,
commonly between 30-80 thousand years with slip rates typically in the range of 0.01-0.02
mm/yr. The northwest-trending faults do not appear to have been active.
18 The NRC-supported program of the Center for Nuclear Waste Regulatory Analyses has identified 52 Type I
faults within a 100-km radius of Yucca Mountain (NRC97a).
19 NRC-supported studies have identified 24 Type I faults within a 10-km radius of Yucca Mountain capable of
generating peak accelerations of greater than 0.3 g (NRC97a).
7-27 •
-------
AM - Ash Meadow
AR - Amargosa River
AT - Area Three
BC - Bonnie Claire
BH - Buried Hills
BLR -Belted Range
BM - Bare Mountain
BUL - Bullfrog Hills
CB - Carpetbag
CF - Cactus Flat
CFML - Cactus Flat-Mellan
CGV - Crossgrain Valley
CHV - Chicago Valley
CLK - Chalk Mountain
CP - Checkpoint Pass
CRPL - Cockeyed Ridge-Papoose
Lake
CRWH - Cactus Range-Wellington
Hills
CS - Cane Spring
DV -Death Valley
EPR - East Pintwater Range
ER - Eleana Range
EVN - Emigrant Valley North
EVS - Emigrant Valley South
FC - Furnace Creek
FLV - Fish Lake Valley
GM - Grapevine Mountains
GRC - Groom Range Central
GRE - Groom Range East
GV - Grapevine
HM - Hunter Mountain
ISV - Indian Springs Valley
JUM - Jumbled Hills
KRW - Kawich Range West
KV - Kawich Valley
KW - Keane Wonder
LM - La Madre
MER - Mercury Ridge
MM - Mine Mountain
NDR - North Desert Range
OAK - Oak Spring Butte
OSV - Oasis Valley
PAH - Pahranagat
PEN - Penoyer
PM - Pahute Mesa
PSV - Pahrump-Stewart Valley
PV - Panamint Valley
PVNH - Plutonium Valley-North
-Halfpmt Rarjge
RM - Ranger Mountains
RTV - Racetrack Valley
RV - Rock Valley
RWBW - Rocket Wash'-Beatty Wash
SF - Sarcobatus Fiat
SOU - South Ridge I
SPR - Spotted Range
STM - Stumble
SWF - Stonewall Flat
SWM - Stonewall Mountain
TK - Tikaboo Valley
TM - Tin Mountain
TOL - Tolecha Peak
TP -TownePass
WAH - Wahmonie
WPR - West Pintwater Range
WSM - West Springs Mountain
YF -Yucca Flat
YL - Yucca Lake ,
Figure 7-12. Index Map of Known or Suspected Quaternary Faults in the Yucca Mountain
Region (Modified from DOE95a). Circles are 50 and 100 km radii from Yucca
Mountain (YM). Faults are identified as follows: ;
7-28
-------
co
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CO
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OH
2
PM
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8
oo
Latest Activity
-
^
^
r-
2
Fatigue Wash
Multiple mid- to late-Quaternary events;
1.7 to 2.5 m displacement of Quaternary deposits
^
ctt ?3
•°
fN
r-
^^
J
0
CN
2
c
1
Solitario Can]
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s
r>->
ts
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ex =3
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fig
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CO
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Stagecoach R
c2 co'
^ CO
& s
" C3
s a
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No offset or fracturing of late Pleistocene or Holocene noi
single fracture in one trench. Fracture zone varies up to 2
's
o
l-« ccj
K
u
0
o\
o
oo
^
1
IQ
z
Ghost Dance
No evidence of Quaternary activity found
p*"
1
00
Jg.
1
z
1 Dune Wash
>%
*a>
^
*-*
6
r^
o ^
-2 ^
Most recent event 48±20 ka; cumulative displacement 0.3
recurrence interval 60 to 100 ka; slip rate 0.002 to 0.01 m
J
tN
10
r~-
w}
\o
^
1
1
o
r— 1
2
Bow Ridge
No recognizable ruptures of Quaternary deposits
£
ro
^
!^
^j
2
2
>>
|| Midway Valle
"cc
,_.
£
3
Six to'eight events evident;
Midway Valley excavation: most recent event at 38±6 ka:
«.- e
O >«r4
•O cS
'" 0
w j2
o
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°. S
o j~
1 1
.& a
•3 R
oo
displacement 1.7 to 2.7 m; recurrence interval 20 to 80 ka
0.02 mm/yr;
Busted Butte exposure: Quaternary displacement 4.8 to 7.
0
o
3
VO
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u
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Paintbrush Ca
cs
c-i-
-------
The three major faults in the immediate region of Yucca Mountain are the Ghost Dance fault,
which passes through Yucca Mountain and the proposed repository; the Bow Ridge fault, just to
the east of Yucca Mountain; and the Solitario Canyon fault, just to the west of Yucca jMountain.
According to DOE's interpretation of available data, the Solitario Canyon fault has shown no
significant movement over the last 40,000 to 110,000 years. No movement has occurred during
the last 10,000 years. The most recent surface-rupturing motion on the Bow Ridge fault is
estimated to have occurred 48,000 ±20,000 years ago, with a recurrence interval most| likely in
the range of 60,000 to 100,000 years. There has been no offset or fracture on the Ghost Dance
fault for the past 20,000 years.
7.1.1.5 Tectonics and Seismicity (Adapted from DOE95a)
The fault systems and the seismic history of the Yucca Mountain area must be considered in the
larger context of regional tectonics. By so doing, predictions of future seismic hazards and their
potential effects on the repository, as well as the performance of natural barriers, can be made
with reasonable certainty, within the limits of the available data. This section discusstes what is
currently known about the tectonic setting of the region encompassing the repository site. Data
concerning the seismicity of the area and historic earthquake activity are also presented.
Regional Plate Tectonic Setting \
|
The plate tectonic setting of the southwestern United States is dominated by the interaction of the
North American and Pacific Plates, hi the Yucca Mountain Region, particularly west of Yucca
Mountain, this interaction is complicated by the overlap of right-lateral plate boundary stress
from these plate movements and extensional stress from the Basin and Range tectonics.
Based on geologic and geodetic measurements, the Pacific plate appears to be moving northwest
at approximately 50 mm/yr relative to the North Atlantic plate. The stresses generated from this
movement are distributed to structural features on the North American Plate and contribute to the
tectonic processes (extension or compression of the crust, folding and faulting, etc.) in the region.
About 35 mm/yr of the motion from the Pacific Plate is absorbed by the San Andreas jfault
system; another 5 mm/yr may be absorbed by coastal strike-slip faults parallel to and west of the
San Andreas fault. The eastern edge of the Sierra Nevada microplate (composed of the Sierra
Nevada Mountains and the Great Valley of California) appears to move northwest at
approximately 10 mm/yr. This latter movement, between the eastern edge of the Sierra Nevada
Mountains and the western edge of the Colorado Plateau, is most likely to contribute to the
seismicity and tectonic processes around the Yucca Mountain site (Figure 7-13). Uncertainties in
I
7-30
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Southern Basin and Range
Figure 7-13. Sketch Map of the Western United States Showing Some Major Structural
Features. Symbols (®) at the latitude of Las Vegas give approximate motions toward the NW in
mm/yr relative to a "stable North America." This interpretation suggests that 10 mm/yr of NW
movement occurs between the Colorado Plateau and the crest of the Sierra Nevada Range, 35
mm/yr occurs on the San Andreas Fault, and five mm/yr occurs west of the San Andreas Fault.
This is consistent with the paleoseismic data and historic observations of strike slip faulting in this
region. (Modified from DOE95a)
7-31
-------
the understanding of the regional tectonic processes include: the amount of compression normal
to the San Andreas fault induced by Pacific plate motion (N36°W ±2°), the rate of relative
i
motion between plates, and the amount of motion taken up within the Sierra Nevada microplate.
The timing and mechanisms for producing the crustal extension which characterizes the
structural and physiographic features of the Great Basin are a subject of debate. Several
mechanisms have been proposed for the extensional tectonic processes that produced the major
land forms of the Great Basin. Relatively high-angle, planar, normal faults cutting brittle crust
can accommodate up to 10 or 15 percent of the crustal extension. Normal faults at a high angle
at the surface and curving to lower angles at depth (listric faults) may accommodate much greater
extension. Modeling of very low angle detachment faults suggests extensive crustal thinning that
may accommodate extension of the crust by 200 percent or more.
i •
The typical Basin and Range structures were developed by about 11 Ma. They are tilted fault
block ranges with relatively large displacement, high-angle normal faults exposed at the surface
bounding one or both sides of each range. Scott (SCO90) suggests that rates of fault movement
were highest between 13 - 11.5 Ma and thereafter decreasing over time..
This crustal extension varied across the region in time and space. One thought is that rapid
Miocene extension migrated westward from Yucca Mountain after about 11.5 Ma and may also
have been nonuniform from north to south. Pliocene and later extension, accompanying a
postulated region-wide uplift starting about five million years ago, is more evenly distributed and
is taken up by movement on high-angle normal faults at depth which are coincident Ayith the
Miocene faults expressed at the surface. This belief is consistent with the evidence of the
existence of faulting to depths of 15 km or more indicated by the pattern of hypocenters for the
current seismicity in the region. . :
Structural Features and Seismicity \
; 1
The relationship between specific structural features, particularly faults, and seismicity in the
Basin and Range Province is not entirely clear. The Central Nevada Seismic Belt (CNSB), for
example, is clearly associated with major faults or fault systems showing historic surface rupture.
However, other zones of seismic activity and areas of diffuse activity show no evidence of
historic surface faulting. One example is the east-west seismic belt, which includes the Nevada
Test Site.
7-32
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The apparently poor correlation between earthquakes and faults may be attributable, at least in
part, to several factors: (1) the short historical record relative to the long recurrence intervals for
earthquakes, (2) the difficulty of accurately locating epicenters in this remote area, and (3) the
unknown geometry of faults at depth. Study of the paleoseismic record for the Quaternary Period
suggests that, in the Yucca Mountain Region, recurrence intervals for surface rupture are on the
order of thousands to tens of thousands of years.
Seismology of the Yucca Mountain Area
In the region around the site, there are several zones in which seismicity is concentrated: the
Sierra Nevada-Great Basin Boundary Zone (SNGBZ), the CNSB, the Southern Nevada
Transverse Zone (SNTZ), the Garlock Fault, and the Mojave Block. All of the zones, except the
Mojave Block, are wholly or partially in the Walker Lane Belt, a major tectonic element of
southwestern Nevada. In addition, there is a broad distribution of seismic activity that is not
associated with any known major tectonic feature throughout much of the Great Basin.
The Walker Lane Belt tectonic element (Figure 7-9) consists of nine structural blocks acting
more or less independently. The belt is defined by a style of faulting within and bounding the
blocks which ranges from northwest-trending right-lateral slip (the Pyramid Lake, Walker Lane,
and Inyo-Mono blocks) to northeast-trending left-lateral slip (the Carson, Spotted Range-Mine
Mountain, and Lake Mead blocks) to east-west trending left-lateral slip (Excelsior-Coaldale
block). Cumulative lateral offset on individual major faults ranges from a few kilometers up to
100 kilometers and faults rarely extend to adjacent blocks.
The Walker Lane Belt probably developed in the Mesozoic Period and is still active. Most of the
faults show evidence of Cenozoic movement and numerous zones exhibit Quaternary and
Holocene offset (STE90). Although the recurrence interval for the late Quaternary faulting is
generally thousands to tens of thousands of years, recurrence may be on the order of decades in
some sections of the seismic zone, e.g., the CNSB.
Of the four seismic zones identified in the Walker Lane Belt, the SNTZ is nearest to the Yucca
Mountain site and is the most significant to repository performance. Although the other zones
exhibit recent seismic activity, they are further removed from the Yucca Mountain site and are
less likely to affect the repository.
The Southern Nevada Transverse Zone, which includes Yucca Mountain, is an arcuate belt of
seismicity about 150 kilometers wide, extending from the southern region of the Intermountain
7-33
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Seismic Belt (in southwestern Utah) to the Mammoth Lakes area in California. Historic
earthquakes in this zone have been of moderate magnitude with no documented surface rupture.
Earthquake events include the 1902 Pine Valley, Utah (ML 6.3)20, the 1966 Caliente-Clover
Mountain, Nevada (ML 6.0), and the 1992 Little Skull Mountain, Nevada (ML 5.6) nejar the
proposed site (see Table 7-3). '
Seismic Distribution ' '
i
Studies of the large Great Basin earthquakes suggest faulting on steeply dipping fault planes that
penetrate the upper 15 kilometers of crust as the focal mechanism for many of the earthquakes
observed. In general, mainshock hypocenters for earthquakes of magnitude seven or greater in
this region can be located on the down-dip projection of the surface rupture observed!along faults
identified in the field, suggesting that large Great Basin events occur on steeply dipping planar
faults at depths less than about 15 kilometers. i
Three—with perhaps two additional possible—seismic gaps (areas of no recent seismic activity)
have been identified in the western Great Basin. These gaps occur between the rupture zones of
•[
major historic earthquakes and contain structures that show evidence of prehistoric activity.
Seismic gaps are generally considered to be significant in plate-boundary regions but their
relevance for interplate regions such as the Great Basin is not clear. These gaps may represent
areas of prolonged low or no seismic activity or areas where stresses are not being released by
fault movements.
Significant Historical Earthquakes '
i
i
Figure 7-14 depicts the epicenters for earthquakes of magnitude 3 and greater occurring within
320 kilometers of the proposed site from 1850 through 1992. These data show a clustering of
seismicity in the CNSB and the SNGBZ, as well as in the southern Mojave Desert and along the
San Andreas fault zone. In addition to those identified in the figure, numerous small magnitude
earthquakes have occurred in clusters or as isolated events throughout much of Nevada. The
Garlock Fault and a large portion of the southern Great Basin appear to show relatively little
seismic activity during this period.
20 ML is a measurement of the magnitude of the seismic event. See Table 7-4 for a definition of this and other
magnitude measures.
7-34 :
-------
Table 7-4. Significant Earthquakes within 320 km of Yucca Mountain Site Since 1850
Owens Valley, CA, 1872
March 26, 1872; estimated at Mw 7.8 to ML. 8.0*a; considered largest historic event
of the Basin and Range; surface ruptures along 90 to 110 km on Owens Valley
fault; average net oblique slip of 6.1 ±2.1 m and up to four m vertical
displacement; liquefaction of unconsolidated sediments.
Wonder, NV, 1903
Fall 1903; estimated magnitude 6.5; rupture of the Gold King fault; ruptures of five
to 16 km with fissures up to 1.5 m wide and 1.5 m deep in alluvium; in the same
area as the 1954 Fairview Peak-Dixie Valley earthquakes.
Cedar Mountain, NV,
1932
December 21, 1932; Ms 7.2; about 61 km of discontinuous faulting in a belt six to
14 km wide; displacements up to 1.8 m horizontal and 0.5 m vertical; analysis
indicated main shock was two sources occurring about 20 seconds apart; an Mw 6.7
event and a second Mw 6.6 event; series of seven moderate events in this part of the
CNSB from 1932 to 1939.
Excelsior Mountains,
NV, 1934
January 30, 1934; ML 6.3 (Mw 6.1); on Excelsior-Coaldale section of the Walker
Lane belt; about 60 km west-southwest of the 1932 event; foreshock of ML 5.6
preceded mainshock by 45 min.; surface rupture 1.4 km in length and less than 13
cm vertical displacement. An ML 5.5 earthquake occurred on August 9, 1943,
approximately 40 km southeast.
Rainbow Mountain,
Stillwater, NV, 1954
July 6, 1954; two events of M 6.6 and M 6.4 in Rainbow Mountain area were
followed on August 24 by the Stillwater M 6.8 event initiating a six-year period of
10 events greater than M 5.5 in the CNSB.
Fairview Peak-Dixie
Valley, NV, 1954
December 16, 1954; an ML 7.3 event on the Fairview fault followed four minutes
later by an ML 6.9 event rapturing the Dixie Valley fault; diffuse fracture zone
covering an area 100 km by 30 km from Mount Anna to the northern part of Dixie
Valley; displacements four m right lateral and three m vertical on Fairview Peak
fault and over two m vertical in Dixie Valley.
Caliente-Clover Valley,
NV, 1966
On August 16, 1966; ML 6.0; near Caliente, Nevada, about 210 km east-northeast
of Yucca Mountain. The source depth is estimated at 6 km; with the focal
mechanism a strike-slip motion on steeply dipping plates oriented either north-
northeast or west-northwest.
Mammoth Lakes, CA,
1978-1980
An ML 5.8 earthquake midway between Bishop and Mammoth Lake in October,
1978, was followed 18 months later (May, 1980) by a swarm-like sequence of four
events (ML 6.5, ML 6.0, ML 6.7, ML 6.3) within two days. This sequence was
accompanied by inflation of the resurgent dome in the Long Valley caldera.
Activity continued with moderate earthquake swarms in the southern part of the
caldera with spasmodic tremor sequences usually associated with magma injection
at depth. The Chalfant sequence, discussed below, occurred to the east in 1986.
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Table 7-4. Significant Earthquakes within 320 km of Yucca Mountain Site Since 1850
(Continued) i
Chalfant Valley, CA,
1986
On July 21, 1986, an ML 6.6 earthquake occurred in the Chalfant Valley in eastern
California about 15 km north of Bishop with about 10 km of rupture along the
White Mountains fault zone. The source-depth was located 11 km below the
surface and the focal mechanism indicates right lateral slip on a plane oriented
north-northwest dipping 70° southwest. |
Landers, CA, 1992
The Landers sequence began April 23rd with the ML 6.2 Joshua Tree earthquake,
followed by a sequence of 6000 events. On June 28,1992, an Ms 7.6 earthquake
near Landers, California, ruptured sections of several mapped north- to inorthwest-
trending faults and several concealed unmapped north-trending faults in the south-
central portion of the Mojave block. An extensive aftershock sequence followed,
extending 85 km north of the mainshock and 40 km to the south. The sequence
included the Ms 6.7 Big Bear earthquake three hours after and 30 km west of the
mainshock. Surface rupture extended for 85 km, with displacement averaging two
to three meters across the rupture zone, up to 6.7 m on the Emerson fault, and
minor rupture of faults within 30 km of either side of the main rupture zone. The
Lander event was followed by a sudden increase in seismic activity in the western
U.S. up to 1250 km from the mainshock, with an intense cluster of events in the
Walker Lane belt. This included the ML 5.6 Little Skull Mountain earthquake on
June 29, 1992, approximately 20 km SE of Yucca Mountain. j
Eureka Valley, CA, 1993
On May 17,1993, an ML 6.1 earthquake occurred 30 km southeast of Bishop,
California. The hypocenter was located nine kilometers below the surface in the
southern part of Eureka Valley. Preliminary analysis indicates normal faulting on
a northeast striking plane, perhaps paralleling a north-northwest trending inferred
Quaternary fault in the area.
*• Terms used for earthquake magnitude in the table above include:
ML Local magnitude; this is' the original Richter scale, developed in California for earthquakes with
epicentral distances less than 600 km and focal depths less than 15 km; uses waves with periods of
about 1 s; saturates at M = 7.25;
Ms Surface-wave magnitude; suitable for global distance; uses waves with 20 s periods; saturates at
about M = 8.6; [
Mw Moment magnitude; based on seismic moment (M0 = I^AD), where u = shear modules, A = area of
fault rupture, and D = fault displacement; Mw = 2/3 log M0-10.7; does not saturate; |
M This is assumed to be local magnitude. !
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Figure 7-14. Magnitude 3 or Greater Earthquakes Within 320 Km (200 Miles) of Yucca
Mountain from 1850 to 1992 (Modified from DOE95a)
Earthquakes occurring since 1850 within 320 km of the Yucca Mountain site with magnitudes
greater than 6 are summarized in Table 7-3. These either resulted in surface rupturing or
represent the largest event in a particular seismic-source zone. The most recent strong
earthquake (ML =5 or greater) in the vicinity of Yucca Mountain was the Little Skull Mountain
(ML = 5.6) event in June 1992, associated with the Landers, California earthquake earlier that
year.
Studies of ground motion from recorded seismic activity around Yucca Mountain and of surface
features susceptible to ground motion effects, suggests that Yucca Mountain has not been subject
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to ground accelerations at the surface in excess of 0.2 g for over several tens of thousands of
years. At the depth at which waste is likely to be emplaced in the repository, the effects of
ground motion would be expected to be significantly less. These ground accelerations; do not
present excessive demands on seismic facility design requirements for the repository pr its
associated surface facilities. '
The largest seismic event in the immediate area of Yucca Mountain since 1978 was an ML 2.1
event on November 18,1988, centered 12 km northwest of the proposed repository location. An
earthquake of magnitude Mw 5.7 occurred on June 29,1992, beneath Little Skull Mountain
approximately 20 km southeast of Yucca Mountain. This earthquake is the largest ever recorded
(in about 100 years of records) in the vicinity of the site. It caused minor structural damage to
the Yucca Mountain project field office near Yucca Mountain but had no apparent effect on
geologic features near the mountain. j
Based on a return period of 12,700 years, Bechtel Nevada estimates that for the adjacent Nevada
Test Site there is a 0.55 probability of at least one earthquake of magnitude 6.8 or greater
occurring in the next 10,000 years (SHO97). !
DOE has not considered'seismicity to be a significant factor in repository safety performance.
Seismic effects are not considered in previous total system performance assessments (DOE94a,
DOE95b) because DOE believes that they will have virtually no effect underground. Dowding
and Rozen (DOW78) examined empirical evidence of damage to 71 rock tunnels in Alaska,
California and Japan from earthquake shaking. From this analysis, the authors concluded that,
for peak surface accelerations which would cause heavy damage to above ground struptures,
there was only minor damage to tunnels. No tunnel damage was observed for peak surface
accelerations of less than approximately 0.2g and only minor tunnel damage occurred when the
peak surface acceleration was less than 0.5g.
DOE quantitatively analyzed the variation of ground motion with depth using both stochastic and
empirical methods (DOE94e). Peak surface accelerations were shown to be reduced by a factor
of two at a depth of about 400 m.
! I
DOE considered tectonism in the TSPA-VA released in 1998, including the effects of parameter
variability (DOE98). NRC included the effects of fault displacement impacts and seismic
rockfall impacts on waste packages in TPA 3.1 (NRC97c). ;
' i f
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In its 1996 Phase 3, Yucca Mountain Total System Performance Assessment, the Electric Power
Research Institute (EPRI) did not include consideration of earthquakes since it was concluded
that "...tectonic activity is not expected to significantly impact repository integrity" (EPR96).
The National Academy of Sciences (NAS) supports DOE's view that seismic effects on
underground excavations are usually less severe than on surface facilities (NAS95, p. 93). In
addition, NAS states that while the timing of seismic effects is unpredictable, the consequences
of such events are boundable for performance assessment purposes (Ibid., p. 94). The NAS
further notes that it is possible for the hydrologic regime to be affected either adversely or
favorably by seismic events.
The technical community did not agree with DOE's position on structural deformation and
seismicity presented in TSPA-95. Subsequently, in May 1996, a meeting of involved groups was
held to review and seek agreement on defensible tectonic models based on available data. The
group included DOE, NRC, the Advisory Committee on Nuclear Waste (ACNW), the Nuclear
Waste Technical Review Board, the USGS, the State of Nevada, the EPRI, and the Center for
Nuclear Waste Regulatory Analyses (CNWRA) (NRC97a). Of 11 proposed models, the group
agreed that only five were supported by existing data. Agreement on the five supportable models
was riot unanimous nor was agreement on the relative importance of the five models. In
addition, some of the models may be independent and some may be subsets of others. The five
viable alternative models are:
• Deep detachment fault (12-15 km)
• Moderate detachment fault (6-8 km)
• Planar faults with block deformation
• Pull-apart basin21
• Amargosa shear
The pull-apart basin model proposed by the USGS and the Amargosa shear model proposed by
the State of Nevada are based on buried or blind seismic sources at Crater Flat and involve the
greatest seismic risk. These seismic sources are not included in DOE's Probabilistic Seismic
Hazards Analysis which was used as a partial basis for the conclusions reached in TSPA-95.
Depending on proximity to the repository, the Amargosa shear could result in an earthquake with
magnitude Mw>7.8 and accelerations exceeding 1 g (NRC97a). More recently, CNWRA stated
21 A pull-apart basin is a structural depression formed by localized extension along strike-slip fault zones. The
basin is formed in the brittle upper crust above a horizontal detachment in the lower crust (NRC97a).
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that apatite-fission-track dating from Bare Mountain and Striped Hills does not support the
USGS reconstruction of the Amargosa shear model (McK96). CNWRA believes that the pull-
apart basin model is more tenable but requires additional direct observations of basin-bounding
and cross-basin strike-slip faults.
I
Additionally, DOE argued that future tectonic events are unlikely to significantly alter the
hydrologic characteristics of the Yucca Mountain site. This argument is based on the position
that the current state of faults and fractures at the site is the result of cumulative tectqnic events.
However, CNWRA posits that a single tectonic event can cause significant changes in hydrologic
characteristics. The DOE argument is valid only for characteristics resulting from cumulative
events and not for the most recent single tectonic event (NRC97a).
I F
7.1.1.6 Fractures (Adapted from DOE95a) ! !
!
1 j
Closely allied with tectonic issues is the consideration of fractures in the rocks surrounding the
repository. An extensive fracture network can provide fast paths both for influx of water into the
repository for overlying strata and egress of water potentially contaminated with radionuclides
through underlying strata. To develop an understanding of fractures, studies have been
conducted to examine the age and connectivity of fractures primarily in a portion of the Tiva
Canyon Tuff. Outcrop studies were conducted for a number of units. The studies were designed
to define the general orientations of fracture sets over all of Yucca Mountain and to establish the
relationship of fracture sets to regional tectonic history. A few studies of the vertical continuity
of fractures have been conducted in the Paintbrush nonwelded unit. These are designed to
examine changes in fracture pattern as a function of stratigraphy (DOE95a). ,
i
Four sets of tectonic fractures with consistent orientation were identified within the Paintbrush
Group. In addition, a set of sub-horizontal joints with variable strikes and dips of less than 10
degrees exists. These fracture sets may have originated as extension joints, many of which have
been subsequently been reactivated. It has been postulated that the fractures developed as a
i
mountain-wide response to far-field stresses rather than local movement of structural blocks.
However, data to support this hypothesis conclusively are limited (DOE95a).
Fracture widths are defined both by rock wall separation and actual fracture aperture.; Rock wall
separation is the distance between the fractured surfaces without reference to any infilling with
secondary minerals. Aperture includes the effects of any infilling and is the amount of open
space remaining. Wall separations are typically one to 10 mm from the surface to a depth of
about 200 m. Surface fractures are 50 to 75 percent filled with caliche which reduces the
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aperture to one to two mm. Below about 10m from the surface, the fractures are 40 to
50 percent filled, primarily with quartz and calcite (DOE95a).
Studies of surface fractures have led to the following general conclusions (DOE97c, SWE96):
• Fracture intensity is a function of lithology, variation in the degree of welding in
the tuffs, and, to a lesser extent, proximity to faults
Connectivity of the fracture network also depends largely on the degree of
welding and the lithology
• Width and intensity of fractured zones vary around faults and are related to fault
complexity
The degree of welding within the Paintbrush Group has the greatest effect on the overall
character of the fracture network with fracture intensity and network connectivity behlg least in
nonwelded or poorly-welded units.
Subsurface studies have indicated that correlation with surface features diminishes as the depth
increases because:
• Some faults which displaced units in the Topopah Spring Tuff became inactive
before the overlying Tiva Canyon Tuff was deposited
• Many faults are discontinuous so that the displacement may die out between
observation points
• Faults commonly spread upward resulting in differing surface and subsurface
geometries (DOE97c)
7.1.1.7 Volcanism (Adapted from DOE95a)
To assess the possibilities of disruptive volcanic events, the nature and history of volcanism in
the area must be understood. Yucca Mountain consists of silicic volcanic rocks originating from
the Timber Mountain caldera complex to the north. A resurgence of silicic volcanism is unlikely
since the activity that formed the rocks at Yucca Mountain ceased millions of years ago.
However, basaltic volcanism has taken place more recently. Basaltic volcanism is commonly
accompanied by the intrusion of dikes into the surrounding rocks and could pose the potential for
intrusion into the repository itself if such volcanism occurred close to the repository. Magmatic
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intrusions could mobilize waste and/or alter ground water pathways. The volcanic history of the
Yucca Mountain area is discussed below. <
Yucca Mountain is composed of Miocene volcanic rocks erupted from the overlapping Silent
Canyon, Claim Canyon, and Timber Mountain calderas between 11 and 15 million years ago.
The silicic volcanic tuffs that comprise Yucca Mountain are typical of mid-Tertiary basin arid
range extensional tectonics in southern Nevada. Yucca Mountain, at the depth of the proposed
repository, is comprised of units of the Paintbrush Tuff, a major outflow ignimbrite of the Claim
Canyon caldera segment of the Timber Mountain caldera complex (Figure 7-15). During the late
Neogene (two to 10 Ma) and Quaternary (0 to two Ma) Periods, small-volume, mostly
polygenetic, basaltic centers produced lava flows, air falls, and cinder cones in the area. The
silicic and basaltic volcanism are described below.
Silicic Volcanism \
The-silicic volcanism in the Yucca Mountain area is part of an extensive, time transgressive pulse
l
of mid-Cenozoic volcanism that occurred throughout much of the southwestern United States.
Yucca Mountain is in the south-central part of the SNVF, a major Cenozoic volcanic field
covering an area of over 11,000 km2. Magmatism in the region was distributed in linear belts
parallel to the convergent plate margin during the Mesozoic Era. hi the southwestern United
States, a pause or disruption in the belts about 80 Ma formed the Laramide magmatic gap or
hiatus, which lasted until renewed silicic magmatism began in the northeastern part of the Great
Basin about 50 Ma. Sites of eruptive activity migrated south and southwest across parts of
Nevada and Utah, with eruptive centers distributed along arcuate east-west trending volcanic
fronts. The most intensive eruptions were at the leading edge of the migrating front, with the
I
most voluminous silicic volcanic activity in the Yucca Mountain area occurring between 11 and
15 Ma. Silicic magmatic activity in the area ceased about 7.5 to 9 Ma. The Yucca Mountain
area marks the southern limit of time-transgressive volcanic activity.
Between 10 and 13 Ma, there were two significant changes in the regional volcanic and tectonic
patterns: the southern migration of volcanism halted and the composition of the volcanic activity
changed. Diminished silicic-eruptive activity migrated in less systematic patterns to the
southwest and southeast, leaving a conspicuous amagmatic gap from the southern edge of the
Nevada Test Site south to the latitude of Las Vegas.
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saT Mile*
TIVA CANYON MEMBER DASHED WHEKE INFEKKHD
Figure 7-15. Index Map Showing Outlines of Calderas in the Southwestern Nevada Volcanic
Field and the Extent of the Tiva Canyon and Topopah Spring Tuffs of the
Paintbrush Group (Modified from DOE95a)
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Should volcanism occur in the future, the type of volcanism (basaltic or silicic) is potentially
significant, since silicic eruptions are more explosive. The DOE claims that there has been no
silicic volcanism in the Yucca Mountain Region since about 7.5 Ma at the Stonewall Mountain
caldera more than 100 km northwest of Crater Flat and since nine Ma at the closer Black
Mountain caldera (60 km northwest of Crater Flat). Consequently, DOE has concluded1 that the
potential for future silicic volcanism is negligible (DOE96e). However, work by NRCsuggests
that silicic pumice with an age of 6.3 ±0.8 Ma (based on zircon fission track data) existed
beneath basalts in Crater Flat. This is at odds with the DOE position that post-caldera Silicic
eruptions had not occurred near the proposed repository site (NRC97a). Subsequently, NRC
reported that, based on argon isotope dating, the age of the silicic material was 9.1 ±3 Ma, which
correlates with the eruptions from the Black Mountain caldera (NRC97b). On the basis of this
information, NRC concluded that silicic volcanisnji did not need to be considered in evaluating
the probability and consequences of igneous activity at Yucca Mountain. I
I
Basaltic Volcanism
Two episodes producing basaltic-volcanic rocks have been defined in the Yucca Mountain area,
both occurring after the majority of the silicic volcanism ended. The first, marked by basalt of
the silicic episode (BSE), consists of basalt-rhyolite volcanism postdating most silicic eruptions
of the Timber Mountain-Oasis Valley (TM-OV) complex. The second episode is comprised of
spatially-scattered, small-volume centers marked by scoria cones and lava flows of alkali basalt,
ranging in age from about 10 Ma to less than 10,000 years. These post-caldera basaltsiof the
Yucca Mountain Region are divided into older post-caldera basalts (OPB) and younger post-
caldera basalts (YPB). The locations of basalts in the Yucca Mountain Region with ages of less
than 12 Ma are shown in Figure 7-16 (NRC96). (The cited ages of some of the occurrences
reported by NRC differ slightly from those reported by DOE. The differences are not ;
substantive.) .
The BSE crops out throughout the Yucca Mountain area and is identified by several
characteristics: (1) a close association (in time and space) with activity of the TM-OV complex,
(2) all centers of the BSE are large-volume eruptive units (<3km3 dense-rock equivalent—the
largest centers are in the ring-fracture zone of the Timber Mountain caldera), and (3) a wide
range of gepchemical composition. The BSE occurs in three major groups: i
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Mafic Lavas of Dome Mountain (age 10.3 ±0.3 Ma) are exposed in the moat
zone of the Timber Mountain caldera and comprise the largest volume of basaltic rocks
Figure 7-16. Distribution of Basalts in the Yucca Mountain Region with Ages of Less Than 12
MA (NRC96). Dotted line defines boundary of Yucca Mountain/Death Valley
isotopic province where basalts have same relatively unique isotopic structure.
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I f I
• Basaltic Rocks of the Black Mountain Caldera overlap some units offtie
caldera in age :
• Basaltic Volcanic Rocks, Yucca Mountain Area include the basaltic andesite of
Skull Mountain (dated 10.2 ±0.5 Ma), the basalts of Kiwi Mesa, and Jackass Flats
The second episode of basaltic volcanism, marked by the post-caldera basalt of the Yucca
Mountain Region, occurred at sites either well removed from the eruptive centers of the TM-OV
complex or younger than the silicic-magmatic activity. These sites generally consist of small
volume (<1 km3) centers marked by clusters of scoria cones and lava flows. ,
The OPB were'produced along either north-northwest trending Basin and Range faults or at the
intersection of Basin and Range faults with the ring-fracture zone of older calderas. These range
in age from 10.4 to 6.3 Ma and are represented at four localities:
i
• Rocket Wash, thin, basalt lava flows (8.0 ±0.2 Ma) occur at the edge of the ring-
fracture zone of the Timber Mountain caldera • '
• Pahute Mesa, three separate but related basalts (with ages ranging from 8.8 ±0.1
to 10.4 ±0.4 Ma) occur at the intersection of faults with the ring-fracture, zone of
the Silent Canyon caldera
• Paiute Ridge, dissected scoria cones and lava flows (8.5 ±0.3 Ma) are associated
with intrusive bodies occurring at the interior of northwest-trending graben; the
related Scarp Canyon basalt (8.7 ±0.3 Ma) crops out west of Nye Canyon
Nye Canyon, three surface basalts (6.3 ±0.2 Ma, 6.8 ±0.2 Ma, and 7.2 ±0.2 Ma)
and a buried basalt (8.6 Ma) occur in the Canyon. :
The second eruptive cycle, resulting in the YPB, usually occurred at clusters of small-volume
centers aligned along predominantly northeast structural trends. These eruptions occurred from
4.9 Ma to as recently as 0.004 Ma and are represented at the following localities (in decreasing
age): '•
• Thirsty Mesa, a thick accumulation of fluidal lava and local feeder vents erupted
onto a pre-existing Thirsty Canyon Group ignimbrite (welded tuff) plateau (ages
of 4.6, 4.68 ±0.3, and 4.88 ±0.4 Ma are reported for various samples) :
• Amargosa Valley, cuttings from a buried basalt gave ages of 3.85 ±0.05 and 4.4
±0.07 Ma
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Southeast Crater Flat basalt lavas (4.27 to 3.64 Ma) are the most areal-extensive
oftheYPB
Buckboard Mesa basaltic andesite (3.07 ±0.29 to 2.79 ±0.10 Ma) erupted from a
scoria cone in the northeast part of the ring-fracture zone of the Timber Mountain
caldera and from nearby fissures
Quaternary Basalt of Crater Flat consists of a series of four northeast trending
basalt centers extending along the axis of Crater Flat including the Little Cones
(0.76 ±0.20 to 1.1 ±0.3 Ma), the Red and Black Cone centers (1.55 ±0.15 to 0.84
±15 Ma and 1.09 ±0.3 to 0.80 ±0.06 Ma, respectively), and the Makani Cone
(1.66 ±0.522 to 1.04 ±0.03 Ma)
Sleeping Butte Centers are two small volume (<0.1 km3) basaltic centers about
2.6 km apart with an estimated age of 0.38 Ma based on recent argon isotope
dating measurements
Lathrop Wells Center, the youngest and most thoroughly studied center of
basaltic volcanism, involved multiple eruptions over more than 100,000 years
Three alternative models involving various chronologies of volcanic events'have been proposed
by DOE to explain the eruptive history of the Lathrop Wells volcanic center. These include a
four-event eruption model (eruption at >0..!3, 0.08 to 0.09, 0.065, and 0.004 to 0,009 Ma), a
three-event eruption model (eruptions at 0.12 to 0.14, 0.065, and 0.004 to 0.009 Ma), and a two-
event eruption model (eruptions at. 0.12 to 0.14 and 0.004 to 0.009 Ma). Exact dating of the
eruptions has been problematic and the exact number and timing of the eruptions is not certain,
but the youngest eruption is believed to be less than 10,000 years old. This most recent activity
was restricted to minor ash deposits (TRB95).
Summary
The majority of the silicic volcanic rocks that form the most important units in the Yucca
Mountain stratigraphic section were deposited about 11 to 15 Ma. This silicic volcanism ceased
about 7.5 Ma. Silicic volcanism was followed by two subsequent episodes of basaltic volcanic
rock formation. In the first episode, basalts of the silicic episode were deposited about 10 Ma. In
the second or post-caldera episode, smaller eruptions occurred beginning 8 to 10 Ma and
continuing to near present time. The youngest basaltic rocks at the Lathrop Well volcanic center
have ages between 4,000 and 9,000 years.
"~ This value appears to be an anomaly and will be investigated further.
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Both DOE and NRC agree that a future occurrence of silicic volcanism is highly unlikely and
therefore the consequences of such an event need not be considered in system performance
assessment. However, DOE and NRC have not reached agreement on the treatment o( igneous
activity associated with possible future basaltic volcanic events.
Given the history of volcanism in the Yucca Mountain Region, there is some probability that a
volcanic event can either intersect the repository footprint and directly affect the waste or that a
nearby intrusive dike can indirectly affect the natural and engineered barriers. In TSPA-93
(DOE94a), DOE used available data to estimate the impact of indirect magmatic effects, such as
heating or attack by aggressive volatiles on waste packages, when contact of the waste; packages
with magma does not occur. Assuming that the waste packages were vertically emplaced, such
that the thermal loading they produced was 57 kW/acre, the magmatic effect on peak drinking
water doses is virtually indistinguishable from a case in which magmatic effects are not
considered.
In subsequent activities to address the stochastic uncertainty associated with the possibility that a
future magmatic event may intersect the repository, DOE convened a panel of 10 experts and
used a formal elicitation process to develop disruption23 probability estimates (DOE96f). Results
of the elicitation include (DOE97a): • ;
• A mean annual disruption probability of 1.5xlO"8 ;
A 95 percent confidence interval of 5.4x 10'10 to 4.9x 10'8 ;
Upper and lower bounds of 10'10 to 10'7 • \
I i
The NRC has taken a different tack in establishing the probabilities of volcanic disruption. The
NRC approach considers spatial patterns of basaltic volcanism, regional recurrence rates of
volcanic activity, and structural controls on volcanism in the Yucca Mountain Region (NRC96).
Using two different measures to assess the impact of structural controls on volcanism (density of
high dilation-tendency faults and horizontal gravity gradients), two methods to assess spatial-
temporal distributions (near-neighbor and Epanechnikov kernel methods) and regional recurrence
rates varying from two to 10 volcanoes per million years, calculated probabilities based on
NRC's bounding approach ranged from IxlO"8 to 2x10~7 volcanic disruptions per year (NRC96).
23 Disruption is the physical intersection of magma with the potential repository volume (DOE97a).
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Based on a homogeneous Poisson model (i.e., with a time invariant rate), the probability of at
least one volcanic disruption event occurring in 10,000 years, using DOE's estimated maximum
(95 percent confidence) disruption rate of 4.9x10'Vy, is 0.0005. Based on the maximum
disruption rate estimated by NRC of 2xlO"7/y, the probability of at least one disruption is 0.002 in
10,000 years.
In its 1996 Phase 3, Yucca Mountain Total System Performance Assessment, EPRI did not
include consideration of volcanism (EPR96). This position was based on an assessment made by
one member of the expert panel — one of 10 volcanologists sponsored by DOE — who
estimated that the annual probability of a hiagmatic intrusion into the proposed repository is 1.0 x
io-8.
Scientists at UNLV, supported by the State of Nevada, have considered a number of alternative
modeling approaches to volcanism. (See, for example, HO96 and HO95.) Using a non-
homogeneous Poisson model (i.e., with a time varying rate), Ho estimated the probability of at
least one disruption in 10,000 years to lie between 0.0014 and 0.03.
DOE investigated the significance to repository performance of basaltic igneous activity in the
TSPA-VA (DOE98, Volume 3, Section 4.4.2). Scenarios evaluated included impact of an event
where the waste package is breached by the magma and waste is transported to the surface;
impact of a magmatic event where the repository footprint is not intercepted but groundwater
pathways are altered; and impact of a magmatic event where 0 to 170 waste packages are
breached resulting in an enhance source term but no direct transport of waste to the surface. The
probability of direct surfaces releases was estimated to be essentially zero for the first 10,000
years due to the ability of the waste package to withstand magmatic attack over the assumed 5 to
40-day period of the intrusive event. Peak dose rates for direct surface releases are several orders
of magnitude less than for the TSPA-VA base case after one million years. Peak dose rate
CCDFs for the enhanced source term scenario are lower than the base case at both 100,000 and
one million years but the scenario can result in spikes in the dose rate that are greater than the
base case. DOE estimates that over 10,000 years, there is less than one chance in 1,000 that any
igneous activity occurs. If an igneous event into the repository occurs, there is a 60 percent
probability that the source term for groundwater transport of radionuclides would be enhanced.
If the magmatic event does not intersect the repository footprint, the consequences are negligible.
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7.1.1.8 Geologic Stability Issues ; :
The NAS Committee reporfstates that the Yucca Mountain site will exhibit long-term geologic
stability on the order of one million years (NAS95). This implies that the contribution of
geology to overall system performance can be assessed for that time period. The Committee
therefore concludes that there is no need to arbitrarily select a shorter compliance evaluation
period, such as 10,000 years. The Committee recommends "...that compliance assessment be
conducted for the time when the greatest risk occurs, within the limits imposed by long-term
stability of the geologic environment."
I
This section examines the Committee's assertion of long-term geologic stability and related
issues. Factors addressed include characteristics of the geologic and hydrologic systems implied
by the Committee's concepts of "stable" and "boundable;" validity of the assertion of stability;
and the significance of stability to the occurrence, magnitude, and evaluation of peak dose.
Geologic stability does not imply absence of geologic activity or absence of changes in geologic
processes, but rather that any changing characteristics of the system do not introduce
uncertainties of sufficient magnitude to compromise the ability to perform credible analyses of
future repository performance. • |
Characterization of Geologic Stability by the NAS Committee
The NAS report (NAS95) does not specifically define geologic stability. The existence of
stability is discussed six times in the report, in different ways: :
• The geologic record suggests that [the time frame during which the \
geologic system is relatively stable or varies in a boundable manner] is on
the order of one million years. (Executive Summary, page 9)
• ...the long-term stability of the fundamental geologic regime [is] on the
order of one million years at Yucca Mountain, (page 55)
• The long-term stability of the geologic environment at Yucca Mountain ...
is on the order of one million years, (page 67)
• The time scales of long term geologic processes at Yucca Mountain are on
the order of one million years, (page 69)
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• The time scale for long-term geologic processes at Yucca Mountain is on
the order of approximately one million years, (page 72)
The geologic record suggests that [the time frame over which the geologic
system is relatively stable or varies in a boundable manner] is on the
order of about one million years, (page 85)
These characterizations of geologic stability are quite similar, although some are expressed in
terms of the geologic regime itself and others are described in terms of the processes that operate
on or within that regime. These two assertions are not necessarily the same. For example,
characteristics of the geologic regime that are important to peak dose evaluation might remain
stable while tectonic and other natural processes and events continue in the future, even varying
from past characteristics. Alternatively, natural processes and events may continue in the future
as they have occurred in the past (i.e., the processes and events exhibit stability), while the effects
they produce may change the features of the geologic regime that are important to peak dose
evaluation. Conditions in which past and continuing tectonic movement produces differential
movement of deep geologic structures might cause changes in the hydrologic regime important to
the occurrence of the peak dose. The various expressions of stability used in the Committee's
report imply no significant change in either the geologic regime or in the processes and events
that affect the characteristics of that regime.
The Committee's report does not explicitly justify the assertion of million-year stability by
providing a synopsis and interpretation of the geologic record. Some of the references cited in
the report contain information about the geologic record (e.g., DOE's Site Characterization Plan
for the Yucca Mountain site (DOE88)), but none of the cited references interprets the record to
indicate a million-year stability of the geologic regime or the processes associated with it.
Existing Documentation Related to Stability
Existing documentation does not directly address long-term stability of the natural features of
Yucca Mountain and its environs. Until revision of the EPA and NRC regulations for Yucca
Mountain was initiated, the DOE documents containing information about the geologic features
of the Yucca Mountain site anticipated that evaluations of site suitability would be made in
accord with DOE's 10 CFR Part 960 Site Suitability Regulations and anticipated that safety
performance of a repository at the site would be evaluated in terms of EPA's 40 CFR Part 191
regulations and NRC's 10 CFR Part 60 regulations. Under this regulatory framework, the time
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period of concern is 10,000 years. [The NRC's 10 CFR Part 63 regulations and EPA's 40 CFR
Part 197 standards retain this time period. See Section 7.3.11 for a discussion of EPA's
rationale. ;
i [
The 10,000-year time frame for compliance with EPA's 40 CFR Part 191 regulation was selected
by the Agency because it was short compared to long-term factors, such as tectonic motion, that
might affect and change in ways that could not be characterized, the natural environment
conditions important to regulatory compliance evaluations. On the other hand, the time period
was long enough to bring into consideration, at least in principle, factors such as seismicity that
are important in geologic time scales and might affect repository performance. !
"
The DOE has, in many Yucca Mountain project documents, implied geologic stability or the
equivalent for time periods of 10,000 years. The State of Nevada believes, however, that the
record does not justify such a conclusion. For example, the State asserted in its comments
(NEV85) on DOE's draft Environmental Assessment (DOE84) for the Yucca Mountain site, that
DOE's conclusion that "neither major tectonic activity nor the resumption of large-scale silicic
volcanic activity in the area near Yucca Mountain is likely in the next 10,000 years" is ;
premature, based on existing evidence. The State also asserted that "possible hydrovolcanic
activity at Yucca Mountain has not been sufficiently evaluated" (NEV85, Volume II, p^ge 125).
DOE and others have reported a variety of topical studies concerning geologic and hydrologic
phenomena that are relevant to stability of the geohydrologic regime (potential for climate
change and its effects are discussed in Section 7.1.3). Topics addressed include: j
Potential for water table rise (SZY89, ARN96, KEM92, NAS92, DOE98)
• Tectonic movement and its potential effects (BAR96)
Seismicity and its potential effects (CAR91, ARN96) !
Volcanism and its potential effects (DOE96e, DOE96f, HO95, HO96, BAR93)
Potential for rockfall in drifts and its effects (CRW96, DOW98) =
Potential for changes in the fracture network and fracture flow (MAT97)
>
Work was recently initiated, and is ongoing, that attempts to use data from fluid inclusions to
estimate the potential for heated, ascending fluids to reach the repository horizon in the future
(DOEOO). Fluid inclusions are small droplets of the solutions that form minerals that a;re trapped
as defects in the growing crystal. 1
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As discussed in Section 7.3.10, information from the studies cited above and other sources will
be used in DOE's integrated consideration of features, events, and processes that can affect
repository performance in the TSPA evaluations for the Site Recommendation Considerations
Report (SRCR) and the Site Recommendation (SR).
The effect of these phenomena on uncertainty in performance assessment results and on the
potential to evaluate compliance with regulatory standards at far-future times when peak dose is
predicted to occur is discussed in Section 7.3.11.
In general, the documents of record show controversy concerning the stability of the geologic
regime and associated natural processes and events at the Yucca Mountain site and the effect of
natural processes and events on repository performance. The controversy stems both from
opposing interpretations of the available data by DOE and the State of Nevada and by differing
definitions of geologic stability. To some extent, the opposing viewpoints reflect the institutional
positions of the parties involved; nonetheless, the uncertainties in the data permit alternative
interpretations to be made and controversy to persist.
Interpretation of the Geologic Record Related to Stability
The geologic history of the area provides the basis for assertions concerning the stability of the
geologic regime for Yucca Mountain and its vicinity. Site characterization activities for DOE's
Yucca Mountain project, and other activities unrelated to the Yucca Mountain project (e.g.,
commercial characterization of natural resource potential), have yielded an extensive data base
concerning geologic features and the geologic record of the region. The most comprehensive
data available for assessing the geologic stability of the Yucca Mountain site are contained in the
Yucca Mountain Site Description (CRW98a).
Such data do not, however, definitively resolve the question of the long-term stability of the
geologic regime and its impact on projections of repository system performance. Such issues can
be resolved only in context, through the expert judgment of the involved parties. The NAS
Committee's assertion of long-term geologic stability at Yucca Mountain for the next million
years is an example of expert judgment.
The basis for the Committee's judgment of the geologic stability of Yucca Mountain over the
next one million years is the conclusion that the properties and processes of the geologic regime
7-53 :
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important to repository performance "...are sufficiently understood and stable over the long time
scales of interest to make calculations [of repository performance] possible and meaningful"
(NAS95, page 68). The relevant properties and processes include the radionuclide inventory of
the waste, the influx of water to the repository, migration of the water and its contained waste
materials from the repository to the ground water, and subsequent dispersion and migration of
contaminated ground water to the regional biosphere. The Committee considers it possible, for
example, to estimate, with acceptable uncertainty, Concentrations of wastes in ground water at
various locations and times for the purpose of a bounding safety assessment.
i
The assertion of geologic stability implies a judgment that the basic features of the geologic
regime that affect waste release and transport will remain as they are, or change in a limited and
reasonably predictable fashion, over the next million years. In other words, phenomena that
would substantially and unpredictably change the current, relevant geohydrologic regime are not
expected. Such phenomena would include tectonic motion, seismicity, and volcanism sufficient
to change the features of the geologic regime that govern radionuclide release and transport.
The Committee's assertions also imply that the geologic and hydrologic features of the site and
region can and will be characterized in a way that allows repository performance to be reliably
projected on the basis of current conditions. Two pf the parameters cited by the Committee as
important to predicting the performance of the repository—water influx to the repository and
dispersion and migration of ground water in the biosphere—have been demonstrated by DOE
modeling studies (e.g., those for the Total System Performance Assessment for the Viability
Assessment; TSPA-VA, DOE98) to be highly important to estimating potential health effects
from the repository. However, these two parameters are currently among the least well-known of
the parameters related to repository performance. \
i
The DOE performance assessment reports indicate that these hydrologic parameters will be
extremely difficult to evaluate reliably. As DOE notes in the TSPA-VA, direct observation of
water infiltration rates is not possible. Consequently, the TSPA-VA treats the infiltration rate to
the repository as an uncertain parameter. Bounding values, consistent with the NAS
Committee's concept of bounding, can be established, but the bounds may have to be njarrowed
considerably from present ranges to be meaningful to the process of determining compliance.
I
This situation raises an issue not addressed directly by the NAS Committee: Can key
performance-related parameters be adequately characterized? The long-term geologic stability of
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the Yucca Mountain site may be less important to evaluating repository performance than the
actual values of those parameters most significant to its performance. As the example given
above demonstrates, the variability of a parameter such as infiltration rate presents an obstacle to
characterizing reliably the long-term risks to the critical group. In addressing the overall question
of long-term repository performance, the uncertainty associated with these factors may be much
more significant than the uncertainty associated with the long-term geologic stability of the site.
Summary of Evidence for Stability
The information presented in this chapter generally supports the NAS Committee's assertion that
the fundamental geologic regime at Yucca Mountain will remain stable over the next one million
years. The overall picture that emerges from the data is that the site and region had a highly
dynamic period of volcanism, seismicity, and tectonic adjustment in the past, but these processes
and events have matured into a system in which the magnitudes, frequencies, locations, and
consequences of such phenomena can be bounded with reasonable confidence relative to
assessing the long-term repository performance.
The possible exception to this finding is the chance that on-going processes and events are
producing differential changes to the geologic and hydrologic regimes that are currently
unrecognized but could affect repository performance and potential radiation risks for affected
populations in the future. For example, on-going tectonic processes and movements could
potentially have different effects on the geologic and hydrologic regimes near the surface and at
depth, and the at-depth changes may not be readily recognizable. At present, tectonic movement
in the area varies by location but falls generally within the range of four to 10 mm/year
(DOE95a). Over one million years, an annual tectonic movement of 10 mm/year will produce a
total translation of location of about 5 miles. If all of the elements of the geologic and hydrologic
regime important to repository performance and dose estimation do not move together in space
and time, the differential movement could invalidate the results of performance and exposure
assessments. The potential for differential movement and its consequences are not yet addressed.
Perspective on the Significance of Stability of the Geologic Regime
A judgment that the geologic regime at Yucca Mountain will be stable for one million years
enhances confidence in the results of model-based assessments of the effects of natural processes
and events over that time frame on repository performance. Long-term natural phenomena may
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not, however, control repository performance or uncertainties in performance assessment results.
Uncertainties in other factors involved in performance projections may ultimately control the
reliability of the projections.
The existence of long-term geologic stability can assure reliable estimation of long-term peak
doses only if stability-related issues are confirmed to dominate repository performance and
numerical values of relevant parameters have been established with confidence. As discussed
subsequently in Section 7.3, DOE's total system performance assessments indicate that;the rate
of infiltration of water to the repository and the dilution and dispersion characteristics of ground
water containing radioactive contamination released from the repository are among the .dominant
factors in repository performance and dose assessment. The finding that these are among the
most important performance parameters has been sustained throughout the evolution of TSPA
evaluations and the repository design (see Sections 7.3.1 through 7.3.10).
The DOE's performance assessments to date for Yucca Mountain have emphasized release of
nuclides from the repository over a 10,000-year time frame, in response, to the requirements of
EPA's 40 CFR Part 191 regulations, which were applicable until enactment of the WEPP Land
Withdrawal Act. Experience in evaluating repository performance over a 10,000-year time frame
(DOE94a, DOE95b) has shown that repository conditions must be assessed at, or near, the time
when key performance parameters, such as temperature, may be at their peak values. The
10,000-year time frame encompasses the time of highest uncertainty in the effect of repository
design factors important to waste isolation and safety performance. These uncertainties may
have a greater effect on predicting long-term repository performance and regulatory compliance
than a natural process or event, such as an earthquake or a volcanic eruption. This is due to the
high degree of uncertainty in the "nominal" dynamics and performance of the repository's
barriers and the low probability of a major natural process or event occurring. ;
I
Beyond 10,000 ye'ars, however, the technical factors associated with repository design [features
that dominate performance issues earlier may become less important to determining regulatory
compliance at the time of peak dose. If the engineered barrier system is likely to have failed in
the long term, radionuclides will be available for transport to the environment. The DOE
performance assessment report by Intera, Inc. (DOE94b) states that variations in assumptions and
conditions for waste package degradation produce less than a 20 percent variation in results for a
10,000 year assessment period and less than a 10 percent variation in results for a 100,000 year
period. Supplemental calculations in DOE94c show that peak doses and releases at the
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accessible environment boundary over a one million-year period are generally unaffected by
waste package lifetimes up to 100,000 years. As discussed in Section 7.3.11, it is in the time
period beyond 10,000 years that the issue of long-term geologic stability becomes more
important to repository performance.
7.1.2 Hydrologic Features
7.1.2.1 Unsaturated Zone Hydrology
The region beneath the surface of Yucca Mountain in the vicinity of the proposed repository is
characterized by a very thick unsaturated zone, ranging in thickness from about 500 to 750 m.
The variable thickness is produced by the combined effects of rugged topography and a sloping
water table. The presence of a thick unsaturated zone is desirable for siting an underground
waste repository because ground water, and any contaminants it might carry, generally travels
more slowly through the unsaturated zone than through the saturated zone. The thicker the
unsaturated zone, the longer contaminants will take to reach the water table.
In this document, and in the literature generally, the term unsaturated flow actually means
partially-saturated flow, since by definition there can be no water flow through a totally dry
medium. Unsaturated ground water flow is more complex than fully-saturated flow because it
involves the simultaneous movement of water, air and water vapor. For unsaturated media, the
measure of permeability is called the effective hydraulic conductivity. The effective hydraulic
conductivity, and hence the rate of fluid flow, through any given partially-saturated porous
medium depends on the degree of saturation of that medium. The higher the saturation, the
greater the quantity of water that can flow through it, all other factors (saturated hydraulic
conductivity, hydraulic gradient, etc.) being equal. As the degree of saturation reaches
100 percent, the effective hydraulic conductivity approaches fully-saturated hydraulic
conductivity. The dependency between degree of saturation and effective hydraulic conductivity
is complex, due to the nonlinearity of the relationship.
The dependence of unsaturated flow on the degree of saturation is important to understand when
reading the following sections of this document because some of the phenomena described are
not intuitively obvious. An example of this is described later, where it is stated that water
moving downward in the partially-saturated zone encounters zones of increased effective
porosity, which may act as barriers to further downward flow. It may at first seem
counterintuitive that a zone of increased porosity could act as a flow barrier until one considers
that a geological zone with a high porosity possesses a low capillary suction potential. If this
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zone is overlain by a zone which has a lower porosity and thus a higher capillary potential, water
entering the upper zone will be retained there as a result of capillary equilibration. These
conditions will prevail until the gravitational force overcomes the capillary force in the upper
zone as more water enters, which usually happens when the bottom of the upper zone becomes
nearly saturated, allowing water to flow into the lower zone. j
A sequence of nonwelded porous tuffs that overlies the Topopah Spring Member (Section 7.1.1)
may act as a natural capillary barrier to retard the entrance of water into the fractured tuffs. A
similar sequence of nonwelded tuffs underlies the Topopah Spring Member. These underlying
nonwelded tuffs locally contain sorptive zeolites and clays that could be an additional barrier to
the downward transport of some radionuclides from a repository to the water table.
The proposed repository is surrounded by and crossed by numerous strike-slip and normal faults
with varying amounts of offset (LBL96). The repository would be located largely, if not entirely,
within what is known as the "central block" as described below (see Figure 7-8). The structural
geology of this block is less complex than in the surrounding area, although one extensive, nearly
vertical normal fault has been mapped in the block (Ghost Dance Fault). The central block of
Yucca Mountain is a large block beneath the center of the Yucca Mountain ridge and is bounded
on its west side by the Solitario Canyon fault, a major north-striking normal fault witK greater
than 100 m of offset. West of this fault is a chaotic, brecciated and faulted west-dipping zone
caused by drag on the fault. A zone of imbricate normal faults forms the eastern boundary of the
central block. These faults are west-dipping and have vertical offsets of about two to five m.
Northwest striking strike-slip faults also occur in the area, such as the one forming the northern
boundary of the central block, beneath Drill Hole, Wash. The concept of a central block should
not, however, be taken to imply that the central block or the proposed repository area is free of
faults (USG84a).
Unsaturated Zone Hydrogeologic Units
The detail of the layered volcanic rock sequence beneath Yucca Mountain is very complex. The
various rock units can be separated into a small or large number of units depending up, on the
scale and aims of a particular study. For the purposes of this document, the unsaturated zone is
considered to consist of six hydrogeologic units, based on their physical properties. This
grouping and the description of the six units are based primarily on USG84a, except where
otherwise referenced. Additional data regarding matrix and fracture properties are presented in
the hydrogeologic database developed in DOE95c. ;
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The physical properties within each formation vary considerably, largely due to variation in the
degree of welding of the tuffs. In most cases, physical property boundaries do not correspond to
rock-stratigraphic boundaries. However, it is the physical properties that largely control water
occurrence and flow; the hydrogeologic subunits into which the volcanic sequence is separated
are different than the lithological units outlined in Section 7.1.1.3. The hydro-geologic units are,
in descending order, Quaternary Alluvium (Qal), the Tiva Canyon welded unit (TCw), the
1 Thicknesses from geologic sections of Scott and Book (1^84).
' Scott and others (19831.
3 Inferred from physical properties.
Figure 7-17. Unsaturated Zone Hydrogeologic Units (USG84a) 't
Paintbrush nonvvelded unit (PTn), the Topopah Spring welded unit (TSw), the Calico Hills
nonwelded unit (CHn), and the Crater Flat unit (CFu). Figure 7-17 illustrates these
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hydrogeologic units and some of their characteristics. They are described in detail in the
following paragraphs.
Structural features, although they are not hydrogeologic units in the same sense as stratigraphic
units, are mappable, have certain measurable hydraulic characteristics, and may have a
significant effect on unsaturated zone flow. Because these structural features are regarded as
important components of the unsaturated hydrologic system, they are described later in this
section.
i
Qal. Unconsolidated alluvium underlies the washes that dissect Yucca Mountain and forms the
surficial deposit in broad inter-ridge areas and flats nearby. Thickness, lithology, sorting, and
permeability of the alluvium are quite variable; particles range in size from clay to boulders, and
in places the unit is moderately indurated by caliche. Alluvial and colluvial deposits generally
have small effective hydraulic conductivity, large specific retention, and large effective porosity
as compared to the fractured rocks. Therefore, a large proportion of the water infiltrated into the
alluvial and colluvial material is stored in the first 'few meters of the soils and is lost to;
evaporation during dry periods. The saturated permeability of alluvium generally is substantial
compared to the tuff units. i
TCw. Lying immediately beneath the Qal is the Tiva Canyon welded unit, consisting of
devitrified ash-flow tuffs ranging from 0 to 150 m in thickness across the site. The TCw is the
densely to moderately-welded part of the Tiva Canyon Member of the Paintbrush Tuff. This unit
is the uppermost stratigraphic layer that underlies much of Yucca Mountain; it dips 5° to 10°
eastward within the central block, resulting in a relatively planar eastward-sloping, dissected land
i
surface. The unit is absent in some washes and is about 150 m thick beneath Yucca Crest. This
unit has a fracture density of 10 to 20 fractures/m3 and small matrix permeability. Saturated
matrix hydraulic conductivity has been estimated at about 2x10~6 m per day (m/d); the effective
hydraulic conductivity is thought to be lower, as saturation is estimated to range from 60 - 90
percent. Neither bulk rock nor fracture hydraulic conductivities are well characterized for this
unit.
PTn. The Paintbrush nonwelded unit is situated below the TCw unit and consists of the
nonwelded and partially welded base of the Tiva Canyon Member, the Yucca Mountain Member,
the Pah Canyon Member, the nonwelded and partially-welded upper part of the Topopah Spring
Member, and associated bedded tuffs. All are part of the Paintbrush Tuff. The unit consists of
thin, nonwelded ash-flow sheets and bedded tuffs that thin to the southeast from a maximum
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thickness of 100 m to a minimum thickness of about 20 m. The unit dips to the east at 5° to 25°;
the dip at any location depends on the tilt of the faulted block at that site. In the central block,
the dip rarely exceeds 10°. In the vicinity of the central block, this unit crops out in a narrow
band along the steep west-facing scarp along Solitario Canyon.
Tuffs of this unit are vitric, nonwelded, very porous, slightly indurated, and in part, bedded. The
unit has a fracture density of about one fracture/m3. Saturated hydraulic conductivities of five
core samples of the matrix have a geometric mean of about 9.0x10"3 m/d. Porosities average
about 46 percent, but some porosities are as much as 60 percent. The rocks of this unit are
moderately saturated, with an average value of about 61 percent. However, water contents are
relatively large; the mean volumetric water content is about 27 percent and the mean water
content by weight is about 19 percent. The maximum values reported are: saturation, 80 percent;
volumetric water content, 42 percent; and water content by weight, 36 percent.
TSw. The Topopah Spring welded unit consists of a very thin upper vitrophyre, a thick central
zone consisting of several densely welded devitrified ash-flow sheets and a thin lower vitrophyre
of the Topopah Spring Member of the Paintbrush Tuff. The unit, which varies from 290-360 m
in thickness, is densely- to moderately-welded and devitrified throughout its central part. The
TSw contains several lithophysal cavity zones that generally are continuous, but vary appreciably
in thickness and stratigraphic position. The TSw is also intensely fractured. !
The Topopah Spring Member is the thickest and most extensive ash-flow tuff of the Paintbrush
Tuff. The central and lower densely-welded, devitrified parts of the Topopah Spring welded unit
are the candidate host rock for a repository. This part of the unit contains distinctive subunits
that have abundant lithophysal gas cavities within the central block. The saturated hydraulic
conductivity of the matrix of this unit generally is small and has a mean of abbut 3.0xlO"6 m/d.
Because of the densely fractured nature of this unit, bulk hydraulic conductivity is substantially
greater than matrix hydraulic conductivity. Saturated horizontal hydraulic conductivity of the
rock mass is about one m/d for a 120-meter interval of the TSw that was packed off and tested at
Well J-13 (see Figure 7-18 for bore hole locations), about six km east of Yucca Mountain.
Because of the marked contrast between the matrix and the bulk hydraulic conductivities in this
unit, values of the bulk hydraulic conductivity from Well J-13 (USG83) and borehole UE-25a#4
probably represent the hydraulic conductivity of the fractures in this unit. The large bulk
hydraulic conductivity of this unit probably promotes rapid drainage of water. The amount of
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flow carried in the fractures with respect to the matrix has been estimated to range between 10
95 percent (GEO97). I
Figure 7-18. Locations of Deep Boreholes in the Vicinity of Yucca Mountain (USG96a)
The effect of lithophysal cavities on the hydrologic properties of the TSw is not well understood.
Total porosity is much greater where lithophysal cavities are more abundant than in those
sections that are free of these cavities. Overall unsaturated hydraulic conductivity probably is
decreased by the presence of these cavities. These cavities commonly are several centimeters in
diameter, filled with air, and form capillary barriers with the fine grained matrix. In effect, the
cavities decrease the transmissive cross-sectional area, decrease effective porosity, and
consequently, decrease the effective hydraulic conductivity.
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CHm. Beneath the TSw unit is a series of non- to partially-welded ash-flow tuffs called the
Calico Hills nonwelded unit. Locally, these may be vitric (CHnv) or zeolitized (CHnz). The
CHn includes the following components, in descending order:
1. A nonwelded to partially-welded vitric layer, locally zeolitic, that is the lowermost
part of the Topopah Spring Member of the Paintbrush Tuff.
2. Tuffaceous beds of Calico Hills.
3. The Prow Pass Member of the Crater Flat Tuff, which is nonwelded to partially-
welded where it occurs in the unsaturated zone beneath the central block.
4. The nonwelded to partially-welded upper part of the Bullfrog Member of the
Crater Flat Tuff where it is above the water table.
In the vicinity of the central block, this unit crops out in a narrow band along the steep west-
facing scarp along Solitario Canyon. Both vitric and devitrified facies occur within the CHn. As
described below, the permeability of the vitric facies is substantially greater than that of the
devitrified facies. Alteration products in the devitrified facies include zeolites (most abundant),
clay, and calcite (rare). Because this facies is mostly zeolitic, it is hereafter referred to as the
zeolitic facies. Thickness of the zeolitic facies generally increases from the southwest to the
northeast beneath Yucca Mountain. Beneath the northern and northeastern parts of the central
block, the entire unit is devitrified and altered.
Both the vitric and zeolitic facies of the CHn are very porous, with a mean porosity of about
37 percent for the vitric facies and 31 percent for the zeolitic facies. Saturations in this unit
generally are greater than 85 percent, with a mean value for the zeolitic facies of about
91 percent.
A significant difference exists in values of vertical hydraulic conductivity of the matrix between
the vitric and zeolitic facies of the CHn. The mean vertical hydraulic conductivity of the matrix
of the vitric facies is 4.0xlO~3 m/d. The geometric mean of the vertical hydraulic conductivity of
the matrix of the zeolitic facies is about S.OxlO"6 m/d. The marked contrast in vertical hydraulic
conductivities of the two facies probably is the result of extensive argillization in the zeolitic
facies, which tends to decrease permeability.
CFu, In approximately the southern half of the central block, the lowermost unit in the
unsaturated zone is the Crater Flat unit. This unit consists of the unsaturated welded and
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underlying nonwelded parts of the Bullfrog Member of the Crater Flat Tuff. No differentiation is
made between the welded and nonwelded components of the Crater Flat unit because of the
limited extent of the unit in the unsaturated zone beneath the central block, and therefore, its
probable limited effect on the unsaturated flow system. Beneath the central block, the thickness
of the CFu ranges from 0 to 160 m. Little is known about the unsaturated hydrologic properties
of the unit, but it is assumed that the properties are similar to those of the nonwelded arid welded
counterparts higher in the section. ! :
Structural Features I
As previously described, the central block of Yucca Mountain is bounded on three sides by
faults. Because these major faults and fault zones transect the full thickness of the unsaturated
zone, they may by hydrologically significant either jas flow barriers or as flow pathways. The
variation in unsaturated hydraulic properties of these features have in most cases not been
measured. However, some inferences can be made, based on the physical properties of the
welded and nonwelded tuff units and on observations of drill cores.
i
The welded units are relatively brittle. Open faults have been observed in cores even from below
the water table. Conversely, the nonwelded units generally are more ductile than the welded
units and more readily produce a sealing gouge material. Fault zone's are less common in the
Calico Hills nonwelded unit. In general, hydraulic [conductivity varies greatly along the faults
and is greater in welded units than in nonwelded urjits (USG84a).
I ' '
Knowledge of the permeability of the numerous faiilts which cross Yucca Mountain is important
because some faults may ac{ as conduits for rapid vertical flow in the unsaturated zone. This
possibility is especially critical in areas in which su!ch faults may intercept large amounits of
lateral flow and divert this flow downward, potentially into the repository. Evidence for the
permeability of the faults in and around the proposed repository area is mixed. Studies:
performed to date indicate that particular faults are barriers, while other faults are more
permeable (LBL96). It is also possible that a particular fault may be relatively impermeable in
some areas of the fault plane, and relatively permeable in others. Factors which may reduce
permeability of faults include development and alteration of fault gouge, deposition of fracture
coating materials on fault surfaces, and the juxtaposition of permeable and nonpermeable units
by movement along the fault plane. Faulting can also create zones of enhanced permeability
where the rock around the faults is highly fractured or brecciated.
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Studies in the Exploratory Studies Facility (ESF) indicate that the permeability of the Bow Ridge
fault is about the same as measured with air permeability testing of highly permeable bedded tuff
formations or highly fractured welded units. Also, the geothermal profile in borehole ONC#1
shows that the geothermal profile is offset by several degrees as the borehole passes through the
Bow Ridge fault zone. This indicates that the fault may be highly permeable to gas or moisture
flow which decreases the temperature in that region (LBL96).
Evidence from other faults indicates that they may act as low permeability barriers. For instance,
the water body observed at borehole SD-7 is thought to be perched over a zeolitic layer and
prevented from moving laterally by the presence of the Ghost Dance fault. A similar hypothesis
has been invoked to explain perched water in a borehole intersected by a splay of the Solitario
Canyon fault. This conclusion is corroborated by pneumatic pressure data taken in borehole UZ-
7a, which appear to show a degree of anisotropy in the fault which is consistent with a
permeability barrier, at least in the horizontal direction (LBL96).
Another indication that some faults at the site may act as permeability barriers is obtained from
potentiometric surface measurements. For instance, the potentiometric surface elevation on the
western side of the Solitario Canyon fault is approximately 40 m higher than on the eastern side
of the fault. This gradient could only be maintained if the Solitario Canyon fault is somehow a
permeability barrier to flow (LBL96).
The ESF has provided data and observations regarding the structural features within Yucca
Mountain. Prior to the construction of the ESF, detailed geological and structural cross-sections
were prepared. As-built cross sections prepared from data and observations from the ESF show
that geologic sections drawn prior to construction compare favorably with results from tunneling.
These findings indicate that the lithostratigraphy, and to a lesser extent structure, of this are well-
characterized and predictable. Detailed information on the results of ESF geological mapping is
available in BOR96 and BOR96a. These publications provide detailed fracture pattern analysis
including measurements of trace length, orientation, continuity, roughness, aperture, and mineral
infilling. From ESF studies, three main fracture sets are reported; two are approximately vertical
and strike north-south, and east-west, while the third fracture set is close to horizontal. BOR96
reports that the open distance between fracture faces averages 2.3 mm over the entire fracture
population. The largest aperture is 91 mm, although this is anomalously large in this population;
67 percent of the fractures are closed (0 mm). For fractures with an aperture greater than zero,
the average is 7.2 mm. The fracture population includes measurements from the Tiva Canyon
Tuff, the Paintbrush Tuff, and the Topopah Spring Tuff. The repository horizon is generally
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more fractured, containing an average of about four fractures per meter, but typically ranges from
about two to six fractures per meter (LLNL96). , j
A common feature in some horizons in the volcariic rocks are lithophysal cavities, which are
voids in the rock presumably created by gases exsolved from cooling lavas and pyrocl;astic
deposits. In the Tiva Canyon and Topopah Spring Tuffs, lithophysae are mostly concentrated
into stratiform zones, but they also occur adjacent to lithophysal zones and sporadically in
nonlithophysal zones. The cavities range in size from less than one centimeter (cm) to greater
than 1.4 m. Fractures demonstrate several different relationships with lithophysal cavities.
Fractures that intersect and terminate in lithophysal cavities are common. This, and other
evidence, suggest that lithophysal cavities may locally influence fracture propagation (BOR96,
BOR96a).
I
Ground Water Flow In The Unsaturated Zone \ ;
'• \
Water flow and storage in the unsaturated zone is three-dimensional and is controlled; by the
structural, stratigraphic, thermal, and climatological setting. The dynamics of water-air-vapor
flow in the layered, fractured rock unsaturated zone beneath Yucca Mountain are complex and
highly uncertain at this time. In the unsaturated zone, water is present both in liquid and vapor
phases within the interstitial, fracture, and lithophysal openings. Hydrogeologic features that
probably affect flow significantly in the unsaturajted zone include the presence of fractured
porous media, layered units with contrasting properties, dipping units, bounding major faults, and
a deep water table. These features probably result in the occurrence of phenomena such as flow
in both fractures and matrix, diversion of flow by capillary barriers, lateral flow, perched ground
water zones, and vapor movement. ;
Infiltration Rates ;
The ultimate source of water in the unsaturated zone at Yucca Mountain is precipitation on the
mountain. The spatial and temporal relationships between infiltration and recharge are complex,
because of the hydrogeologic variability of Yucca Mountain. Some water that infiltrates returns
to the surface by interflow; another part is returned to the atmosphere by evapotranspiration. A
small quantity that is not evaporated, or discharged as interflow, percolates deep into, the
unsaturated zone and becomes net infiltration or percolation. The terms "infiltration" and
"percolation" are used frequently, sometimes interchangeably, in literature about the Yucca
Mountain unsaturated zone. For the purposes of this report, "infiltration" is used to describe the
7J66
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amount of water which enters Yucca Mountain at the ground surface, while "percolation" is used
to describe the amount of water which actually penetrates deep enough into the mountain to
reach the repository horizon and below. The difference between the two terms lies mainly in the
partitioning of part of the infiltration flux into the vapor phase, which may then be recirculated to
the atmosphere.
At Yucca Mountain, the infiltration rate is both spatially and temporally variable. Because the
quantity of net infiltration that percolates through different paths is quite variable, estimated
average recharge rates do not represent percolation rates through specific flow paths. Spatial
variations of infiltration depend mostly on variations in the properties of surficial units,
topography, the intersection of faults with the surface, and the presence of local fracturing.
Temporal variations in infiltration rate are related to the seasonality and relatively infrequent
precipitation events in the arid climate of Yucca Mountain. Temporal variations in the
infiltration rate have also occurred over a much larger time span, reflecting long term climate
changes.
Knowing the temporal and spatial variability of the percolation rates is crucial to modeling
efforts because of the importance of the relationship of infiltration rate to horizontal and vertical
permeabilities of the various units and the effect this has on whether or not significant lateral
flow occurs in the unsaturated zone. The higher the actual infiltration rate, the greater the
likelihood of significant lateral flow. Such lateral flow could result from a combination of two
factors. The first factor is that infiltrating water may encounter zones of lower relative
permeability as it moves downward. The second factor is that in many of the units, the relative
permeability is far greater in the direction parallel to bedding than the direction perpendicular to
it. The anisotropic permeability may cause lateral flow of mounded water away from the area in
which it accumulates. Lateral flow is important because it could transmit water to structural
features which would then move the water downward, possibly acting as a conduit to divert large
amounts of water flowing downward through a small area. Such flow paths could direct water
into and through the repository or away from it.
The actual quantity of net infiltration or percolation beneath the surface of Yucca Mountain has
not been accurately determined. The percolation flux is a difficult parameter to determine for
low flux regions such as Yucca Mountain. There are currently no reliable direct measurements
that can be made to determine this important parameter (LBL96). Existing estimates have been
obtained from a mixture of indirect methods involving field testing and modeling of various
processes at different scales. Data exist to suggest that the flux reaching the repository horizon
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through the matrix is relatively small. Relatively low matrix saturations measured in the upper
portion of the TSw suggest that much of the moisture which infiltrates into the TCw does not
reach the TSw (LBL96). Data from the ESF show that no weeping fractures were found, even in
the region where perched water is found in boreholes. (Note, however, that because of
ventilation equipment inside the ESF, much of any such moisture might be removed from the
ESF as water vapor). Furthermore, no moisture was observed infiltrating into the radial
boreholes of Alcove 1 of the ESF after storm events, even though the boreholes are located close
to the land surface in the highly fractured and broken TCw formation (LBL96). However, other
data suggest that the percolation flux may reach the repository level mainly through episodic
fracture flow. These data include observation and testing of extensive bodies of perched water
located below the repository horizon, as well as measurements of bomb-pulse isotope levels from
atmospheric nuclear testing which show that some water in the unsaturated zone is relatively
young (LBL96). ] .
Estimates of net infiltration vary from slightly negative (net loss of moisture from the mountain)
to about 10 mm/yr (LBL96). USG84a reports that net infiltration flux probably ranges from 0.5
to 4.5 mm/year, based on estimates of earlier workers for various localities in the Yucca
Mountain area. Flint and Flint (FLI94) provide preliminary estimates of spatial infiltration rates
that range from 0.02 mm/yr, where the welded Tiya Canyon unit outcrops, to 13.4 mm/yr in areas
where the Paintbrush nonwelded unit outcrops. The bulk of the area above the repository block
is underlain principally by the Tiva Canyon member. The DOE's 1995 Total System ,
Performance Assessment (DOE95b) concludes that, if the predominant flow direction's vertical,
then the average infiltration through the repository block, using the average infiltration rates of
Flint and Flint (FLI94), would be 0.02 mm/yr. If, on the other hand, the predominant how
direction has a significant lateral component due to material property heterogeneity and/or
anisotropy and the sloping nature of the hydrostratigraphic unit contacts, then the average net
infiltration rate over the repository block could be as high as some weighted average of the
infiltration rates inferred from FLI94. The 1995 TSPA (DOE95b) also reports that the average,
spatially-integrated infiltration rate is about 1.2 nhm/yr; most of this infiltration occurs along the
Paintbrush outcrop in the washes north of the repository block.
Recently, several lines of evidence have converged to alter the prevailing view regarding the
magnitude of infiltration/percolation rates beneath Yucca Mountain, with the most recent
estimates being revised upward from previous work. The newer estimates of percolation are
around five mm/yr, with a range of one to 10 mm/yr (LANL96, LBL96). Recent isotopic
analyses of rock samples from the ESF are consistent with a percolation rate of five mm/yr
I
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(LANL96, LBL96). Profiles of temperature vs. depth of water in boreholes are consistent with a
range of infiltration rates from one to 10 mm/yr (LBL96). Three-dimensional modeling results of
percolation flux at the repository horizon using the latest available spatially varying infiltration
map Indicate percolation fluxes on the order of five to 10 mm/yr. The expert elicitation panel
estimates for mean infiltration rates range from 3.9 to 12.7 mm/y (GEO97). The effect of
uncertainty in infiltration and percolation flux rates is examined in the discussion of the
unsaturated zone conceptual model.
Conceptual Model(s)
The first detailed conceptual model of unsaturated zone flow at Yucca Mountain was proposed in
USG84a. Since then, the majority of the data collected has been in general agreement with these
ideas and concepts (LBL96). Most subsequent conceptualizations of unsaturated zone behavior
are largely refinements of this model, revised to accommodate newly-acquired data (Figures 7-19
and 7-20). Newly-acquired data include isotopic analyses, concentration ratios of ions .dissolved
in matrix rocks and perched water zones, calcite fracture fillings, and thermal modeling of
vertical temperature gradients. Perhaps the most significant change from early conceptual
models has been the recent acquisition of new isotopic data which indicate the presence of "fast
paths" for water moving through the unsaturated zone. This topic is discussed in more detail in a
subsequent section..
The following presentation of the unsaturated zone flow conceptual model is taken primarily
from USG84a. Where appropriate, the published literature is referenced when describing
refinements or revisions that have been made to the USG84a model. The following conceptual
model is presented as if it were an established physical reality. Bear in mind, however, that the
proposed model is probably not the only reasonable description that could be made of the system.
Following the description of the conceptual model is a discussion of critical unknowns, their
effects on unsaturated zone flow, and results of numerical modeling studies.
Percolation of infiltrated water through the exposed fractures of the Tiva Canyon welded unit is
relatively rapid because of the large fracture permeability and small effective porosity of this unit
compared to the alluvial material. Therefore, a large proportion of the infiltrated water normally
is percolated sufficiently deep within the fractured tuff to be unaffected by the evaporation
potential that exists near the surface. Depending on the intensity of the infiltration, percolation
downward through the Tiva Canyon welded unit may occur without a significant change in rate.
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Ghost Dance f
*•' FouH 1 VosiotAl Spolicl
^x^^^ 1 Tefnpardtn(flroli
^""biir I !
Figure 7-19. Early Conceptual Model of Ground Water Flow in the Unsaturated Zone at Yucca
Mountain (USG84a)
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A ALLUVIUM
TC TIVA CANYON WELDED UNIT
P PAINTBRUSH NONWELDED UNIT
TS TOPOPAH SPRING WELDED UNIT
CH CALICO HILLS NONWELDED UNIT
NOTE: NOT TO SCALE
CF CRATER FLAT UNIT
DIRECTION OF LIQUID FLOW
DIRECTION OF VAPOR MOVEMENT
PERCHED WATER i
Figure 7-20. Current Conceptual Model of Ground Water Flow in the Unsaturated Zone at
Yucca Mountain (LBL96)
A small proportion of the water percolating through the fractures slowly diffuses into the matrix
of the Tiva Canyon welded unit. Downward flow in the matrix is very slow because of the small
effective hydraulic conductivity of the matrix. During dry periods, some of the diffused water
flows back into the fractures and probably reaches the land surface by vapor diffusion. The mass
of water involved during this process is likely to be negligible compared to the percolating water.
The densely fractured Tiva Canyon unit, with small matrix porosity and permeability, overlies
the very porous, sparsely fractured Paintbrush unit. A marked contrast in material properties
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exists at the contact between these two units; depending on the magnitude of the infiltration flux,
this contrast could impart a significant lateral component of flow. Flow of water through
fractures of the Tiva Canyon unit occurs rapidly until it reaches the contact. At this point, the
'velocity is significantly decreased because of the greater effective porosity and lesser hydraulic
conductivity of the Paintbrush unit. As a result, lateral, unsaturated flow of water above this
contact can occur. Perched water may occur above this unit if displacement along faults has
created significant differences in permeability on opposite sides of the fault. :
I
The saturated hydraulic conductivity of the Paintbrush nonwelded unit in the direction of dip is
10 to 100 times greater than saturated hydraulic conductivity in the direction normal to the
bedding plane. The combination of dipping beds and differences in directional permeability
creates a downdip component of flow. The magnitude of this component depends on the
magnitude of the principal hydraulic conductivity ratio. The permeability contrast may be
sufficient to decrease vertical percolation into the underlying Topopah Spring welded unit to
almost zero. In this case, water would flow laterally downdip until structural features are
encountered that create perching conditions or provide pathways for vertical flow.
i
As water moves downward through the PTn, the effect of high porosity and low fracture density
progressively moves water from fractures into the matrix. Except for areas where fast paths may
exist (such as faults), beyond a certain depth in the PTn, flow may be almost entirely in the
matrix. Travel times through the matrix of the PTh are thought to be relatively long because the
matrix of this unit appears to act as a "sponge" which dampens out episodic infiltration pulses.
Water flows from the matrix of the Paintbrush nonwelded unit into the fractures or matrix of the
underlying Topopah Spring welded unit. Owing to the thickness of this unit, it is hypothesized
by ROB96 that water moving through the fracturek eventually diffuses into the matrix and moves
very slowly downward. An exception is the second subunit of the TSw (ROB96). In contrast to
this conceptualization, the unsaturated zone expert evaluation panel estimated that up to 95
percent of the flow in the Tsw could remain in the fractures (GEO97). •
Flow enters the Calico Hills nonwelded unit either from the matrix of the Topopah Spring
welded unit or through structural flowpaths. How much flow occurs in the fractures of the lower
part of the Topopah Spring unit is unknown, and therefore their potential to contribute to flow
into the Calico Hills unit is also uncertain.
The nature of flow at the contact between the Topopah Spring welded unit and the Calico Hills
nonwelded unit depends on whether the vitric or zeolitic facies of the Calico Hills unit is present.
-------
The permeability and effective porosity of the vitric facies are much greater than those of the
matrix of the Topopah Spring unit, which may result in a capillary barrier where those units are
in contact. Conversely, the permeability of the zeolitic facies is about the same as for the matrix
of the Topopah Spring unit, resulting in continuity of matrix flux across the contact.
Flux within the Calico Hills unit may occur with some lateral component of downdip flux,
because of the existence of layers with contrasting hydraulic conductivity in the unit. A large
scale anisotropy probably is caused by intercalation of tuffs with alternately large and small
permeability and 'by compaction.
Water that flows downdip along the top of the Calico Hills unit slowly percolates :into this unit
and slowly diffuses downward. Fracture flow is known to occur near the uppermost layers of the
Calico Hills unit, but diffusion into the matrix may remove the water from the fractures deeper in
the unit and thereby limiting flow mostly to within the matrix, except along the structural
flowpaths. It is possible, however, that fractures provide significant avenues for rapid flow
through this unit. Beneath the southern part of the block, the Crater Flat unit occurs between the
Calico Hills unit and the water table. Included are the welded part and underlying nonwelded
part of the Bullfrog Member of the Crater Flat Tuff.
Fluxes along many structural flowpaths are probably larger than within the units they intersect.
The Calico Hills unit is more ductile than the overlying Topopah Spring unit, which may give the
Calico Hills unit fracture sealing properties. In addition, because of the lesser shear strength of
this unit compared to that of the Topopah Spring, gouge formation along faults and shear zones
is more common. These properties may result in a smaller fracture conductivity in the Calico
Hills unit. In the case where the structural flowpaths are hydraulically continuous across the
upper contact of the Calico Hills unit, water would be more likely to flow downward without a
significant change in its path until it reaches the water table. In cases where the structural flow
paths are discontinuous across the upper contact, flow may be diverted downdip along this
boundary. Intermediate conditions between the two extreme cases are also possible. Recent
numerical modeling (LBL96, ROB96) of flow through the unsaturated zone has provided
important insights into the possible characteristics of flow in each subunit of the unsaturated
zone. Some of these insights are discussed in the following paragraphs.
Discussion of Unsaturated Zone Conceptual Flow Model and Modeling of the Unsaturated Zone
Under current conceptualizations the net infiltration rate through the unsaturated zone beneath
Yucca Mountain is one of the most critical parameters for determining the nature of flow in the'
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unsaturated zone, yet it is one of the least well characterized. Numerous modeling studies, based
on varying conceptual models, have been performed to simulate unsaturated flow beneath Yucca
Mountain (e.g., DOE94a, DO.E95b, LBL96, ROB96). Sensitivity analyses performed in these
studies indicate that uncertainty in the amount of net infiltration accounts for as much as 90
percent of the variability in the results.
The magnitude of infiltration flux has a significant bearing on the potential for lateral unsaturated
flow beneath Yucca Mountain. In the Paintbrush nonwelded unit, the overall hydraulic;
conductivity parallel to bedding is 10 to 100 times greater than that in the direction normal to the
bedding plane. At higher flux rates, the potential vertical flow rate of some units is exceeded,
thereby inducing a significant lateral component onflow to the infiltration flux. Some authors
have examined the possibility of "focused recharge," a phenomenon in which surface rainfall
runoff is directed to areas where faults intersect the surface. Significant amounts of recharge
may infiltrate into these zones, which may induce lateral unsaturated flow in the underlying units
(LEH92). One obvious area where this may be occurring is the northern extension of Solitario
Canyon fault, which bounds Yucca Mountain on the west. As previously described, lateral flow
could direct water to structural flow paths, which may then redirect the flow vertically
downward, providing a "fast path" and potentially reduced travel times to the saturated zone.
There is growing evidence to suggest episodic water flow at Yucca Mountain may take place
along "fast paths" (LBL95, FAB96, LBL96). Data obtained from recent sampling conducted
within the ESF tunnels drilled into Yucca Mountain provide compelling evidence that hot only
does flow occur along "fast paths," but that such flpw is capable of moving considerable
distances over a relatively short time frame. The amount of water which may be infiltrating by
fast paths is obviously of critical importance to predicting repository performance. Samples
taken in the ESF tunnel show elevated concentrations of some radionuclides, principally
chlorine-36, as well as lesser amounts of tritium and technetium-99 (FAB96). Chlorine-36 is a
radioactive isotope produced in the atmosphere and carried underground with percolating ground
water. High concentrations of this isotope were added to meteoric water during a period of
global fallout from atmospheric testing of nuclear devices, primarily in the 1950's. This "bomb-
pulse" signal can be used to test for the presence of fast transport paths (FAB96). j
Testing for bomb-pulse radionuclides was conducted by collecting and analyzing rock 'samples
from the ESF. Systematic samples were collected ;every 200 m, and feature-based samples v/ere
collected whenever a structural feature such as the intersection of the tunnel with a fault, was
recognized. The results of the testing indicate that most of the samples had 36C1 ratios ranging
from 400e-15 to 1300e-15. The analysis in LANL96 indicates that although many samples
7-74 ;
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showed 36C1 ratios above present day atmospheric levels, it is believed that they represent
Pleistocene water which entered the system when the 36C1 ratios of infiltrating water were higher
than they are today. Samples with 36C1 ratios above 1500e-15 were interpreted as containing a
component of bomb pulse water, indicating that at least a small proportion of the water at those
locations is less than 50 years old. Locations at which multiple samples showed indications of
bomb-pulse 36C1 ratios appear to be associated with the Bow Ridge fault zone, the Drill Hole
Wash fault zone, and the Sundance fault zone (ROB96). The most significant result of the 36C1
testing is that some water travels to the repository horizon in less than 50 years. It is important to
recognize, however, that these results do not indicate that all water travels this quickly in the
unsaturated zone. The 36C1 data do not indicate what fraction of the water now in the unsaturated
zone has traveled by fast paths, nor do they by themselves indicate the magnitude of infiltration
fluxes. Age dating, numerical modeling, and other lines of evidence suggest that travel times for
most of the unsaturated zone are on the order of thousands to tens of thousands of years (LBL96).
Recent numerical modeling studies (LBL96, LANL96, ROB96) suggest two important
requirements for rapid (less than 50 years) transport of 36C1 to the ESF: (1) a continuous, high
permeability pathway must exist to depth, and (2) a means of focusing infiltration and
maintaining flux to the pathway must exist for a sufficient time. The eastward dip of the highly
permeable PTn unit allowed strong lateral flow which was subsequently diverted downward at
faults in these simulations. The strong lateral, down dip flow in the PTn was subsequently
channeled into local permeability highs. In both the Paintbrush and Calico Hills units several
vertical "fast paths" developed in response to these conditions. The recent modeling suggests
that where the PTn is relatively thick, it was necessary to modify fracture properties to represent
greater fracture densities and/or fracture apertures in order for bomb-pulse 36C1 to migrate to the
ESF in less than 50 years (ROB96).
The presence of perched water has implications for travel times, flow paths, and fluxes of water
through the unsaturated zone. Analysis of water from several perched water zones documents a
number of important findings, including perched water compositions that are out of equilibrium
with pore water, showing little fracture/matrix interaction (DOE96d). This indicates that the
perched water probably reached its present location without extensive travel through and
interaction with the rock matrix, thus suggesting that this water had traveled relatively quickly
through the unsaturated zone. Recently-measured tritium concentrations in perched water are at
background levels, therefore suggesting that perched water is older than thermonuclear weapons
testing. Also, preliminary data from isotope testing of perched water samples from boreholes
UZ-14 and SD-7 indicates an apparent residence time of about 10,800 years, with corrected ages
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ranging from 5,000 to 10,800 years (LBL96). A detailed conceptual model of perched water is
presented in LBL96. |
Radionuclide Transport in the Unsaturated Zone \
The travel time of radionuclides beneath Yucca Mountain is a function of both physical and
chemical processes and interactions between fluid and rock. In terms of physical processes,
radionuclides travel by gas phase and liquid phase advection, dispersion, and diffusion.
Radionuclide travel times to the accessible environment are a function of the percolation flux
distribution in the unsaturated zone and the advective flux distribution in the saturated zone, as
well as the hydrostratigraphy along the ground water flow paths between the repository and the
accessible environment. The percolation flux distribution within the Topopah Spring
hydrostratigraphic unit (and other unsaturated zone units below it) is a function of the infiltration
rate and the complex mechanism of ground water flow in the unsaturated zone. Chemical
influences on radionuclide travel times include rejtardation processes involving liquid and gas
phase diffusion, ion-exchange, adsorption on solids, surface complexation, colloidal suspension,
chemical reactions, mineral alteration and dehydration reactions, radioactive decay, and
precipitation/dissolution reactions. ;
In particular, the key conceptual uncertainty in the transport of radionuclides through the
unsaturated zone at Yucca Mountain is the presence of fracture flow and transport which might,
if fracture pathways are continuous and interconnected, lead to the formation of "fast paths" to
the underlying saturated zone. \
i
Uncertainties in chemical retardation mechanisms and the lack of rock/radionuclide interaction
data also lead to considerable uncertainty in predicting future repository performance. For
instance, in TSPA (DOE95b), modeling efforts hkve simulated fluid/rock interactions that can
serve to chemically retard the transport of radionuclides with a simple equilibrium (infinite
capacity) distribution coefficient (Kd) model. Generally, values for distribution coefficients are
related to both the chemical nature of the individual hydrostratigraphic unit and to the properties
of the radionuclide. Since distribution coefficients are used to model such a wide variety of
phenomenological processes, they are modeled in TSPA-95 as stochastic parameters with a high
degree of uncertainty. This process results in a broad range of predicted times it would take
radionuclides to travel from the repository to the water table. Radionuclides that are little
affected by chemical retardation (e.g., I, Tc) could reach the water table within the same time
frame as the ground water. Alternatively, Kds used in TSPA-95 for a number of radionuclides
(i.e., Am, Ra, Cs, Sr) result in travel times to the water table that are 50,000 times greater than
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those for the ground water. Plutonium exhibits significant sorption on all types of Yucca
Mountain tuffs, with sorption coefficients often in excess of 100 cubic centimeters per gram
(cc/g) (ROB96). Detailed analysis of laboratory data for 237Np showed that a nominal sorption
coefficient of 2.5 cc/g could be used in the clinoptilolite-rich zeolitic rocks, with a value of 0 cc/g
elsewhere. Measured Kd values for 79Se are on the order of one cc/g. Sorption of uranium,
similar to 237Np, is significant only for zeolitic tuffs (ROB96).
Recent numerical modeling of the role of rapid transport through fractures was studied for 237Np
(ROB96). For peak dose criteria, the model indicates that the peak may be a result of rapid
radionuclide transport through fractures. However, this does not mean that most of the
radionuclides travel through fractures. According to this model, 10 percent of the source
radionuclides typically travel rapidly in the fracture system, while 90 percent traveled much
slower in the matrix material. (Other conceptualizations suggest that up to 95 percent of flow is
in the fractures.) These results must be interpreted with the realization that the distribution of the
simulated flux between the fractures and matrix is entirely the result of the parameters used to
characterize the system. The Calico Hills, the primary unit through which radionuclides must
travel to get to the water table, is poorly characterized; nothing is known of its fracture hydraulic
properties.
Simulations of 36C1 ratios and 14C in the unsaturated zone indicate that infiltration rates between
one and five mm/yr are more consistent with the field measurements than infiltration rates on the
order of 0.1 mm/yr (ROB96). The environmental isotope simulations also helped provide a
reasonable explanation for the bomb-pulse 36C1 ratios measured in the ESF. This explanation
involves disturbance of the PTn (e.g., faulting) which led to increased bulk fracture
permeabilities and provided a local hydrologic environment conducive to rapid fracture flow of a
small fraction of the total infiltrating flux. The flow in the fractures associated with these
disturbances is rapid enough to transport solutes from the ground surface to the ESF in less than
50 years.
When flow and transport in fractures is simulated using a particle tracking method, a bimodal
distribution of travel times is obtained — an early arrival through fractures, followed by a much
delayed breakthrough of radionuclides that traveled through the matrix (ROB96). Although
ROB96 predicts that the percentage of the total radionuclide inventory that travels rapidly to the
water table is small, the radionuclide flux entering the saturated zone is at its greatest level
during this period, and thus the peak dose is controlled by fracture transport. Migration of
radionuclides through fractures is likely to be retarded by diffusion and in some cases adsorption.
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ROB96 noted that there is an inverse relationship between infiltration rate and arrival time of
first breakthrough peak.
i
Due to sparse data and limited or nonexistent testing of the CHn, characterization of fracture
hydrologic properties in this unit is based on speculation and application of theoretical
relationships (ROB96). Model simulations indicate that the nature of fracture flow in the Calico
Hills is critical to characterizing the performance of the site. Changes in estimated hydrologic
property values estimated for these units have considerably altered the simulated flow and
transport behavior through the unsaturated zone natural barrier.
7.1.2.2 Hydrologic Characteristics of Saturated Zone Units
I'
In contrast to the unsaturated zone in which the floW of water is considered to be primarily
vertical, ground water flow in the saturated zone at Yucca Mountain is principally in the
horizontal direction. This consideration, coupled with the fact that it is the saturated zone in
which most downgradient radionuclide transport from a repository would occur, requires the
description of saturated zone hydrology to cover ari area much greater than Yucca Mountain
itself. Thus, while the discussion of unsaturated zone hydrology is conveniently limited to the
Tertiary volcanic rocks beneath the proposed repository, this section broadens in scope to include
not only the saturated volcanic rocks, but also the adjacent Paleozoic carbonates and th|e alluvial
I
basin fill deposits. Because of the complex three-dimensional geometric relationships of these
geologic materials, the BID breaks the description of saturated zone hydrology into two parts.
Section 7.1.2.2 is restricted to a description of each of the three individual geologic materials
(volcanic rocks, alluvium, and Paleozoic carbonates) and their hydrogeologic properties; Section
7.1.2.3 attempts to describe the geometric and hydrologic relationships of the various units to one
another and to present an integrated picture of regional ground water flow.
Before beginning a detailed description of the hydrologic properties of the individual aquifer
units, it will be helpful for the reader to keep in mihd the following information while reading
this section. As previously described, Yucca Mountain is composed of a thick sequence of
Tertiary volcanic rocks. Beneath Yucca Mountain1^ the thickness of these rocks is more than
1,800 m (SPE89). The Tertiary volcanic sequencej is underlain by complexly folded and faulted
Paleozoic sedimentary rocks, including thick sections of carbonate rocks (SPE89). The
Paleozoic rocks beneath the volcanic section are water-saturated and capable of transmitting
ground water, probably over great distances. Bounding Yucca Mountain on three sides are
downdropped basins filled with alluvial deposits eroded from the surrounding mountains. Water
recharged in the higher altitude areas north of Yucca Mountain flows generally southward
7-78
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through the volcanic, carbonate, and alluvial aquifers toward discharge areas located in the
southern Amargosa Desert and in Death Valley.
Volcanic Aquifer
At Yucca Mountain, where the volcanic rocks may or may not be fractured and where the
hydrologic properties can change significantly in a single stratigraphic unit, stratigraphic units are
useful only in a very general sense for defining hydrogeologic units. The volcanic rock section
beneath Yucca Mountain has been divided informally into the four hydrogeologic units shown in
Figure 7-21: (1) the upper volcanic rock aquifer, (2) the upper volcanic confining unit, (3) the
lower volcanic aquifer, and (4) the lower volcanic rock confining unit. Note that the boundaries
of these hydrogeologic units do not correspond necessarily to stratigraphic or thermal/mechanical
units as defined by other studies. Ground water flows through all of these units to some degree
(where saturated); these hydrogeologic unit designations serve primarily to distinguish between
zones which transmit relatively large quantities of ground water ("aquifers") and zones which
transmit lesser, but not necessarily insignificant, amounts of ground water ("confining units")
(DOE95e; USG94a). • -
The largely nonwelded and intensely altered lower volcanic section, the Lithic Ridge Tuff and
older tuffs, is a confining unit. The variably-welded Crater Flat Tuff constitutes an aquifer of
moderate yield. The tuffaceous beds of Calico Hills are largely nonwelded and are zeolitized
where saturated; however, this unit is significantly less altered than the lower volcanic section.
Where saturated, it is generally a confining unit, but locally parts of the formation are permeable.
The Topopah Spring Member of the Paintbrush Tuff is predominantly densely welded and has
abundant lithophysal horizons. It contains the zones of greatest primary and secondary
permeability and constitutes the most productive aquifer in the tuff section, where it is saturated
(FRI94). Units of the lower volcanic aquifer generally are completely or mostly in the saturated
zone. Because it is deeper, increased lithostatic load probably accounts for part of the difference
between the two aquifers, but the lower aquifer also tends to be less fractured than the upper
volcanic aquifer. The lower volcanic aquifer is also more altered, which accounts for the
decreased permeability (USG96a).
7-79
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TYI
01
UE
O
200
40O
eoo
BOO
1000
1200
1400
ieoo
1BOO
2000
HCAL SELECTED SATURATED ZOH
.PTH GEOLOGIC HYDROCEOLOCI
IN UNITS UNITS
TERS
'
,„
o
tn TopOpCft Sfmg
S Tirff
J>
1
«.«.*-*.
Pro. Pen Tuir
9-
3
_ BJfrog TuH
a
•i
"o
t> Iron Tull
Flow breccias
and lavas
Lithic Ridge Tuff
Older tuffs, lavas,
and breccias
?
Eleano Formatian
Cambrian to
Devonian
formations
ProterazcMC rocks
....
> »
Upper
volcanic
aquifer
uppv xHecnic
Lower
volcanic
aquifer
Lower
volcanic
confining
7
Upper elastic
confintig unit
Lower
carbonate
aquifer
Proterozoic
confining
unit
E TYP
: • DD
I
MET
LAND SURFACED 0-
100
1 200
300
\ 400
\
s
\ 500
\
•\
\ 600
\
\. ;
\ 700
\BOO
\
1000 '
I
CAL SELECTED THERMAL/ SATURATED ZONE
TH GEOLOGIC MECHANICAL. HYDROGEOLOCIC
II - UNITS UNITS UNITS
ERS (Orllz end olhen.
. IMS)
Paintbrush Group
Aluvium
v^ Cinjli
Trfl
Topopah
Spring
Tuff
Cofco KIs
Formalion.
§•
O
o
I
t
a
Pro. Pai
. Tufl
BJHroq Tuff
Tram loll
EXPAND
EO L
UO
PTn
TSwl
TSw2
n>]
CHnl
CHn2
CHn3
PPw
CFUn
BFw
CMM
OMC
CRUn3
TRw
PPER UMTJ
Upper
volcanic
aquifer
Upper
volcanic
confining
unit
i
[
Lower
volcanic
aquifer
i
i
Figure 7-21. Saturated Zone Hydrostratigraphy of Volcanic Rocks (USG96a)
7-80
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The physical properties within each formation vary considerably, largely due to variation in the
degree of welding of the tuffs. The nonwelded tuffs are characterized by having a relatively large
primary porosity, but low permeability. This low permeability results from small pore sizes and
the presence in many nonwelded units of secondary alteration minerals (primarily zeolites and
clays). The welded tuffs are typically very hard and densely welded. The welded tuffs are
commonly more highly fractured than the nonwelded units. The fractures in the welded tuffs
endow them with a significant bulk permeability. For this reason, many of the welded tuff units
are capable of transmitting greater quantities of water than their nonwelded counterparts
(USG84a). !
The occurrence of the water table is not restricted to any one hydrogeologic unit. .Directly
beneath Yucca Mountain, the water table occurs primarily within the Calico Hills Formation and
toward the southern end of Yucca Mountain in the underlying Crater Flat Tuff. To the east of
Yucca Mountain, in the vicinity of Forty Mile Wash, the water table occurs in theTopopah
Spring member of the Paintbrush Tuff. The occurrence of the water table in different
hydrostratigraphic units is attributable to three factors: (1) the vertical displacement of
hydrostratigraphic units by the numerous faults that dissect the area, (2) the eastward dip (five to
10 degrees) of the volcanic units, and (3) the variable elevation of the water table. See USG93a
and USG84b for graphical depictions of the relationship of the water table to stratigraphic units
and FRI94 for a map of the geology at the water table.
Aquifer Geometry
The thickness of the volcanic units is greatest to the north of Yucca Mountain toward the
eruptive centers of the Timber Mountain Caldera Complex (USG85a; USG90a), diminishing
gradually from the eruptive centers to zero thickness at the limits of the southwest Nevada
volcanic field. The thickness of the volcanic deposits also varies considerably for two reasons.
First, these units were deposited on a topographic surface of considerable relief. Second, erosion
and postdepositional structural events have significantly modified their original distribution and
thickness (USG85a, p. 8). In the vicinity of Yucca Mountain, the only direct measurement of the
thickness of the volcanic sequence has been at Well UE-25p#l, where the thickness was
measured to be 1,244 m. Seismic reflection studies have not yielded definitive data, owing to
absorption of reflected energy by the thick volcanic cover (USG85a). Drill hole USW H-l,
located immediately north of the proposed repository boundary, was drilled to a depth of 1,829 m
entirely in volcanic rocks. Thus, the thickness of the volcanic sequence at the north end of Yucca
Mountain may exceed 2,000 m.
7-81
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The saturated thickness of the volcanic unit has been measured only at Well UE-25p#l. At this
location, the water table is 752.6 m above mean sea level (MSL) and the bottom of the volcanic
sequence was encountered at 129.1 m below MSL,| giving a saturated thickness of the volcanic
rocks of approximately 881.7 m (USG84c). Other information can be used to provide a crude
approximation of the saturated thickness of the volcanic units. For example, the elevation of the
water table beneath Yucca Mountain ranges from 1029 m above MSL at the northern part of
Yucca Mountain to 729 m above MSL at the southern end of Yucca Mountain, a difference of
300 m (USG94a). Assuming that the bottom of the volcanic sequence beneath Yucca Mountain
is 129 m below sea level everywhere (which it assuredly is not), the saturated thickness of the
volcanic sequence would range from about 1,158 to 858 m.
The subsurface extent of the volcanic units south of Yucca Mountain is not reliably known
because the volcanic rocks dip under and are covered by alluvial deposits of the Amargosa
Desert. See Figure 7-15 for an illustration of the generalized extent of the volcanic rocks in
southern Nevada and Figure 7-22 for a schematic cross-section showing the southward thinning
of the volcanic units. Aeromagnetic maps suggest that the volcanic rocks pinch out at about the
latitude of Lathrop Wells, and therefore, alluvial deposits constitute most or all of the cover in
the Amargosa Desert (USG85a). Further evidence for the disappearance of the volcanic rocks is
provided by two oil exploration wells drilled in the Amargosa Valley (DRI94). These two wells
were drilled through alluvium into the underlying carbonate aquifer without encountering any
volcanic rocks. USG85a, p. 12, notes that the "southward thinning of the volcanic rocks has
been placed in question by recent north-south unreversed seismic refraction measurements.
Preliminary profiles suggest that some highly magnetized volcanic rocks may indeed thin as
proposed, but that an underlying rock sequence of less magnetized volcanic rocks may continue
southward far beyond Lathrop Wells." USG91a notes the presence of rhyolitic volcanic units
within the Amargosa Basin, although the genetic relationship of these units, if any, to the
volcanic rocks that comprise Yucca Mountain is not clear.
Bare Mountain, Ideated approximately nine kilometers to the west of Yucca Mountain across
Crater Flat, consists of Paleozoic rocks. Tertiary volcanic rocks are known to lie beneath the area
may be located at the eastern bounding fault of Bare Mountain. To the north and east of Yucca
Mountain, the volcanic sequence continues for several to several tens of kilometers.
7-82
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South
YUCCA MOUNTAIN
Center ol
dsposdsile CHOCOLATE MOUNTAIN
—>-.'. ' .••.•-•••.•- »-..•..•• >•..-..•• vnuxiMc- n ... ,-.••;,....,.•., .-..;
JW^rr^.^^^
s?..' ••• •;*.;• •? • .;* • 0..':v • •?._• .v • •*.&?*&%«••;•••• •? • "»-••?• "vrtcfa- •»--.V »•-.! • ;-.!r
Lower Clostic Aquitord
tie. Sedimentary)
Kforneters
Upper Clostic Aquitard
tie. Sedimentary)
Figure 7-22. Schematic North/South Cross-Sectional Illustration of Thinning of Volcanic Units
Beneath the Amargosa Desert (USG85a)
Hydraulic Conductivity
Rock properties largely control the characteristics of water occurrence and flow in the saturated
zone. Rock properties, in turn, are dependent on eruptive history, cooling history, post-
depositional mineralogic changes, and structural setting. Permeability of ash-flow tuffs is in part
a function of the degree of fracturing, and thus, the degree of welding. Densely-welded tuffs
fracture readily; airfall tuffs do not. Therefore, the distribution of permeability is affected by
irregular distribution of different tuff lithologies and is a function of proximity to various
eruptive centers. Permeability is also a function of proximity to faults and fracture zones
(USG82a).
The most reliable method for determining aquifer hydraulic properties are pumping tests,
especially those in which drawdowns are measured and analyzed in wells other than those being
pumped. More than 150 individual aquifer tests have been conducted at and around Yucca
Mountain since the 1980s. Most hydraulic data were from tests conducted in the lower volcanic
aquifer and in the lower volcanic confining unit. Very few data were available for the upper
confining aquifer and the upper volcanic confining unit. Almost all the tests were single-
borehole tests in specific depth intervals and included constant-discharge, fluid-injection,
pressure-injection, borehole flow meter, and radioactive tracer tests. Multiple-borehole tests
have been conducted only at the C-well complex (USG96b, DOE96a). Most reported values of
7-83
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hydraulic conductivity available in the published literature were calculated from transmissivity
values calculated from single-borehole pumping tests and should be regarded as "apparent
hydraulic conductivity." Single-borehole tests doinot record drawdown data from a large enough
sample of the aquifer to be considered reliable. Drawdown data in the pumped well may be
affected by a variety of factors such as fractures, well efficiency, borehole storage, gravity
drainage, and borehole plumbing. USG96b reported that transmissivity and apparent hydraulic
conductivity values determined using multiple-borehole hydraulic tests tend to be much
higher—about two orders of magnitude—than values reported for single-borehole tests
conducted at the same borehole.
Laboratory permeameter testing has been conducted on core samples taken during drilling of
boreholes at Yucca Mountain. Welded units were reported to have matrix hydraulic
conductivities with geometric means ranging frorn 2-OxlO'6 to S.OxlO'6 m/day and bulk hydraulic
conductivities of 0.09 to 10.1 m/day. The nonwelded units have variable hydraulic
conductivities, with geometric means ranging from 2.6x10'5 to 3.0xlO'2 m/day (USG84a).
USG91b reports that, for Well USW H-6, water production during pumping tests was coincident
with fractured, partially, and partially- to moderately-welded tuff units. The reverse was not
necessarily true; that is, not all fractured partially-welded tuff units produced water. USG91b
also states that for Well USW H-6 "porosity and permeability of these rocks is generally
inversely related. Porosity is greatest near the top and bottom of ash flow tuff units and is the
least near the center. Permeability, as indicated by water production, is greatest near the center of
units, where the degree of welding is greatest."
Hydraulic conductivity of the Topopah Spring Member, as determined from aquifer testing of a
120 meter interval of Well J-13, located about five miles east of the crest of Yucca Mountain, is
about one m/d. Below the Topopah Spring Tuff Member, tuff units are confining beds.
Hydraulic conductivities of units tested below thel Topopah Spring Member at Well J-13 range
from 0.0026 to 0.15 m/d (USG83). !
Beneath Yucca Mountain, the Topopah Spring Member is above the water table. Wells installed
in Yucca Mountain are open to the upper volcanic aquitard (Calico Hills Formation) and the
lower volcanic aquifer (Crater Flat Tuff). Pumpijig tests conducted in these wells derived water
primarily from the Bullfrog and Tram Members of the Crater Flat Tuff (USG91b). Hydraulic
conductivities calculated from single-borehole pumping test data are shown in Table 7-5.
7-84
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Table 7-5. Hydraulic Conductivities Calculated from Pumping Test Data
^^•^•-Wellt ; ;' -~\ ^[^^.-K^m/day) "•• '- ':
UE-25b#l
USWH-4
USW H-6
USWG-4
0.46
0.3-1.1
0.85
1.34
Source
USG84d
USG85c ;
USG91b
USG86
In addition to the cautions expressed above regarding the accuracy of single-borehole pumping
test analyses, it is important to recognize that the values of hydraulic conductivity presented here
are average values for the entire pumped interval in the well. Borehole flow surveys, in
conjunction with acoustic televiewer logging, indicate that the volcanic rocks are highly
inhomogeneous in the vertical direction and that the majority of water yielded from the wells
derives from a few highly fractured water-bearing zones of limited thickness. The hydraulic
conductivities shown above are likely to significantly underestimate the actual horizontal
hydraulic conductivity of the water-bearing zones and to overestimate the hydraulic conductivity
of the less transmissive zones. USG91b estimates hydraulic conductivities for specific intervals
within the volcanic section. The authors calculated a hydraulic conductivity of about 9.1 m/d for
a 15.2-meter section of the Bullfrog Member and 6.7 m/d for a 10.4.-meter section of the Tram
Member.
As previously stated, multiple-borehole tests have been conducted only at the C-well complex
(USG96b, DOE96a). The pumping tests at this location involved pumping of selected horizons
isolated by inflatable packers. In this way, transmissivity and hydraulic conductivities can be
calculated for individual members of an aquifer or confining unit. The following description of
transmissivity and hydraulic conductivity data is taken directly from DOE96a.
The results of four pumping tests conducted from June 1995 to May 1996 indicate that the
transmissivity of the Calico Hills interval typically is 100-200 ft2/d; the transmissivity of the
Prow Pass interval typically is 400-700 ft2/d; the transmissivity of the Upper Bullfrog interval
typically is 400-1,000 ft2/d; and the transmissivity of the Lower Bullfrog interval typically is
18,000-20,000 fVVd. The pumping tests conducted in 1996 indicate that transmissivity is about
the same from UE-25 c#l to UE-25 c#3 as it is from UE-25 c#2 to UE-25 c#l (DOE96a).
Horizontal hydraulic conductivities were calculated from computed transmissivities by dividing
the transmissivity by the thickness of the transmissive rocks in the interval. Horizontal hydraulic
conductivity typically is one to five ft/d in the Calico Hills interval and five to 10 ft/d in the Prow
7-85
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Pass interval. The horizontal hydraulic conductivity of the Upper Bullfrog interval typically is
two to three ft/d from UE-25 c#l to UE-25 c#3 aijtd eight to 10 ft/d from UE-25 c#2 to UE-25
c#3. The horizontal hydraulic conductivity of the Lower Bullfrog interval typically is ;70-90 ft/d
from UE-25 c#l to UE-25 c#3 and 150-210 ft/d from UE-25 c#2 to UE-25 c#3. Composite
horizontal hydraulic conductivity from UE-25 c#2 to UE-25 c#3 consistently was found to be
twice the composite value from UE-25 c#l to UEp25 c#3. Ratios of vertical to horizontal
hydraulic conductivity were determined to range downward from 0.08 to 0.0008 in the Calico
Hills, Prow Pass, and Upper Bullfrog intervals, l^ote that the anisotropy in calculated hydraulic
conductivities between UE-25 c#2/#3 and UE-25 c#l/#3 is opposite of that predicted on the basis
of prevalent fracture orientations. The layout of the three boreholes to form a triangular pattern,
with boreholes UE-25 c#l/#3 located along a line estimated to be the major semiaxis of the
permeability tensor and UE-25 c#2/#3 along a linje estimated to be the minor semiaxis of the
permeability tensor (USG96a, p. 48). One possible explanation for this can be found in the
relative distances of the wells from each other. Well #1 is twice the distance from #3 (pumped
I
well) than is well #2; the apparent anisotropy may result from fracture/channeling effects
associated with sampling the aquifer at different scales.
l
Porosity .
In terms of bulk porosity, the volcanic sequence may be considered to consist of two different
types of tuffs: welded and nonwelded (or bedded). The welding process generally reduces the
matrix porosity. Therefore, the welded tuffs typically have a lower porosity than the non-welded
tuffs (USG75, USG84a). The welded tuffs are alko more highly fractured than their nonwelded
counterparts. USG84a reports that welded units have a mean fracture density of eight to 40
fractures per cubic meter and mean matrix porosities of 12 to 23 percent. The nonwelded units
have a mean fracture density of one to three fractures per cubic meter and mean matrix porosities
of 31 to 46 percent. In both rock types, however,:matrix porosity probably comprises the
majority of bulk porosity because fracture porosities, even in the more highly fractured units, are
reportedly quite small (USG85d). USG85d, using a theoretical model to calculate fracture
porosity, reports a fracture porosity of tuffs penetrated by Well USW H-4 ranging from 0.01 to
0.1 percent. Matrix porosities probably decrease with depth due primarily to lithostatic loading
and formation of secondary minerals (SPE89).
7-86
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Effective Porosity
Effective porosity is that portion of the total porosity that contributes to saturated flow. Many of
the volcanic rocks are characterized by relatively small pore sizes and lack of inter-connectedness
of pores; thus, the effective porosity is normally significantly less than the total porosity.
USG84a, p. 18, reports that preliminary laboratory studies of the vitric facies of the Calico Hills
unit show that only about five percent of the pore space is large enough to contribute
significantly to flow under saturated conditions. USG85d, p. 28, considers that fracture porosity
is a reasonable estimate of effective porosity. USG83, p. 13, reports that effective porosities in
samples of welded tuff, vitrophyre, and zeolitized clayey pumiceous tuff range from 2.7 to 8.7
percent.
Storage Properties
Numerous pumping tests have been conducted in water wells completed in the volcanic rocks at
Yucca Mountain and may be used to estimate storage properties. However, most calculations of
storage coefficients for the volcanic rocks are based on single well pumping tests which generally
do not produce reliable estimates of storage properties. The ground water storage characteristics
of the fractured tuffs at Yucca Mountain are complex (USG85d). Estimates of storage properties
of the volcanic rocks vary widely, depending partly upon the lithology and the degree of
hydraulic confinement of the unit being tested. A particular hydrostratigraphic unit may be under
unconifined conditions at one location and under confined conditions at another. USG91b
calculates a storage coefficient of about 0.2. USG93a, p. 78, calculated storage coefficients for
the more densely welded units that ranged from IxlO"5 to 6x10~5; for nonwelded to partially-
welded ash flow tuff zones storage coefficients were estimated to range from 4x10~5 to 2X10"4.
Composite storage coefficients calculated from the multiple-borehole C-well tests ranged from
0.001 to 0.004 (DOE96a).
The degree of confinement of the volcanic aquifers and confining units varies in ways that are
consistent with the geology of the intervals and their distance below the top of the saturated zone
(USG96b, DOE 96a). Beneath Yucca Mountain, the water table is either within or below the
Calico Hills interval (upper volcanic confining unit); this interval typically responds to pumping
as an anisotropic, unconfined aquifer. The underlying Prow Pass and Upper Bullfrog intervals
(part of the lower volcanic aquifer) respond to pumping as either a leaky, unconfined or fissure-
7-87
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block aquifer. The Lower Bullfrog, isolated by intervals of nonfractured rock, typically responds
to pumping as a nonleaky, confined aquifer. :
Recharge and Discharge
I
Precipitation is the primary source of recharge to the volcanic aquifer (USG86; USG84).
Snowmelt in the Timber Mountain area to the north of Yucca Mountain, as well as on Yucca
Mountain itself, provides some of the precipitation-derived recharge. The occasional intense
rainstorms experienced in the area also provide a source of recharge to ground water. However,
because so much of the water that falls either evaporates immediately or is directed into steep
i
channels along the flanks of the mountains to the permeable talus and alluvial deposits at the
base of the mountain, the extent of this contribution is less certain.
Various methods have been employed to estimate the amount of precipitation that recharges the
saturated zone beneath Yucca Mountain (NDC70; USG84e; USG82b). The most frequently
employed approach is to divide the recharge area ihto a number of zones by altitude and to
assume higher precipitation at the higher altitude zones. Some fraction of this precipitation,
usually less than 10 percent, is then assumed to recharge the underlying saturated zone.
Enhancements of this method allow for variable infiltration fractions to account for factors such
as topography, rock type, and vegetation. In the volcanic system, recharge is more easily
quantified than discharge, and discharge is usually! calculated by assuming that outflows are
equal to inflows. This assumption is necessary, but questionable. Some researchers have raised
the possibility that the volcanic aquifer may still be equilibrating to a long term pulse of higher
recharge during the wetter climate of the Pleistocene (about 10,000 years ago) (USG85f,
USG96a). This possibility is not inconsistent with apparent ground water ages of 9,000 to
15,000 years calculated for the volcanic aquifer (lJSG93a; USG83). NDC70 estimated that the
maximum recharge for Crater Flat and Jackass Flats is three percent of the precipitation rate, or
about 4.5 mm/y. USG84a considers this the upper bound for the recharge rate that may be
occurring in certain parts of the saturated zone beneath Yucca Mountain, estimating that recharge
ranges from approximately 0.5 to 4.5 mm/year. Recent evidence, discussed previously, indicates
that the percolation flux through the unsaturated zone probably ranges from one to 10 mm/yr, and
averages approximately five mm/yr. Most of this percolation flux would be expected to recharge
the saturated zone. i
7-88
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An upward hydraulic gradient from the underlying Paleozoic carbonate unit to the volcanic units
(measured in Well UE-25p#l) indicates the potential for flow in the carbonate rocks to move into
the overlying volcanic units. Additional evidence of upwelling flow from the carbonate aquifer
includes zones of elevated ground water temperature and carbon isotopic relationships. Elevated
temperature measurements from the upper saturated zone indicate the possibility of upwelling
from the carbonate aquifer along the Solitario Canyon fault and in the area between the Bow
Ridge and Paintbrush Canyon faults (USG96a, FRI94). Stuckless et al. (STU91) used the
relationship of the 13C/12C ratio to the 814C of the ground water to argue for at least three sources
of water under the mountain. They tentatively identified the three sources as: (1) lateral flow
from the tuff aquifer to the north; (2) local recharge, probably introduced dominantly by flow in
flash-flood watercourses on the eastern side of Yucca Mountain (Forty Mile Wash); and (3)
water that upwells from the deep carbonate aquifer into the tuff aquifer. Savard (SAV94) has
documented recharge to the volcanic aquifer from intermittent streamflow in Forty Mile Wash.
In a saturated zone ground water model developed by the USGS, areal recharge had to be
specified along Forty Mile Wash for the model to adequately simulate measured potentiometric
levels in the vicinity of Yucca Mountain (USG84e). :
Potential pathways by which ground water leaves the volcanic units include downgradient
outflow, pumping, outflow to the carbonate aquifer, and flow into the unsaturated zone. Of the
four pathways, flow into the unsaturated zone, where it occurs, is probably among the least
significant (USG96a). There is no direct evidence that water from the volcanic units flows into
the carbonate aquifer. Vertical hydraulic gradients, where measured, indicate the potential for
flow is from the carbonate aquifer to the volcanic aquifer. The DOE states that the "current
conceptual model for the regional ground water flow system considers that ground water in the
volcanic rocks beneath Yucca Mountain moves generally southward and discharges in the
subsurface into the valley fill alluvium as the volcanic section thins and ultimately pinches out
south of Yucca Mountain" (DOE95f). Currently, water is pumped from the volcanic aquifer
from two wells, J-12 and J-13, located in Jackass Flat near Forty Mile Wash. These wells supply
water for part of the Nevada Test Site, as well as for all site characterization activities at Yucca
Mountain, including human consumption.
Paleozoic Carbonate Aquifer
Thick sequences of carbonate rock form a complex regional aquifer system or systems that are
largely undeveloped and not yet fully understood. Secondary permeability in this sequence has
7-89
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developed as a result of fracturing and enlargement of existing fractures by solution. The area
underlain by carbonate rocks is characterized by relatively low volumes of runoff. Flow can be
complex and may include substantial interaction with volcanic and basin fill aquifers (USG75).
Due to the extensive, thick cover of volcanic rocks and alluvium in the vicinity of Yucca
Mountain, the local characteristics of the Paleozoic sequence are not well known. In eastern
Nevada, the Paleozoic sequence of sedimentary rocks is commonly divided into four general
hydrogeologic units: the lower clastic aquitard, the lower carbonate aquifer, the upper clastic
aquitard, and the upper carbonate aquifer. Evidence from drill hole data and geologic mapping in
surrounding mountain ranges indicates that only the lower carbonate aquifer may be present in
the vicinity of Yucca Mountain and to the. south. ;
Aquifer Geometry : \
\
Evidence suggests that the lower Carbonate aquiferj underlies the entire area. Exposures of
Paleozoic rocks at the perimeter of the study area include Bare Mountain to the west of Yucca
Mountain, the Funeral Mountains south of the Amargosa Desert, and the Specter Range to the
east and southeast. Further evidence comes from three drill holes which have penetrated the
overlying units to reach saturated carbonate rocks -i- borehole UE-25p#l on the eastern flank of
Yucca Mountain, which penetrated through Tertiary volcanic rocks into the underlying carbonate
sequence, and two oil wildcat wells drilled near Amargosa Valley. Additional information
regarding these wells is provided in Table 7-6.
I
Examination of the altitudes of the top of the carbonate aquifer in Table 7-6 indicates that the
buried surface of the buried carbonate aquifer is quite irregular. This variability is probably a
combination of relief of the original erosional surface of the carbonate units coupled with
structural offsets produced by faulting. >
Saturated thickness of this aquifer is largely unknown; USG75 indicates that water circulates
freely to depths of at least 1,500 feet beneath the top of the aquifer and up to 4,200 feet below
land surface. The effective flow thickness of the aquifer depends, in part, upon the lithostatic
i
pressure at depth, which in turn depends on the thickness of the column of rock overlying the
carbonate aquifer.
7-90'
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Table 7-6. Borehole Location and Depth Data for Wells Drilled to the Lower Carbonate
Aquifer in the Vicinity of and Downgradient of the Yucca Mountain Area
Well ID*
UE-25 p#l
Federal-
FederhoffS-1
Federal-
Federhoff25-l
Latitude &
Longitude
36°49'38"/
116°25'21"
36°35'32"/
116°22'54"
36°37'07"/
116°24'26"
Surface
Altitude (m)
1,114.9
772.9
783.9
Depth to Carbonate
Aquifer (m)
1,244
259
671
Altitude (MSL) of Top of
Carbonate Aquifer (m)
-129.1
513.9
112.9
*Note: Information for well UE-25 p#l obtained from USG84c. Information on oil exploration wells
obtained from DRI94.
Hydraulic Conductivity
Interstitial permeability of the carbonate rocks is negligible; essentially all of the flow transmitted
through these rocks is through fractures. Permeability measurements of the carbonate rocks are
reported as transmissivity values, as opposed to hydraulic conductivity values, because the
thickness of the carbonate unit through which water is flowing is not well known. Estimates of
fracture transmissivity range from 1,000 to 900,000 gallons per day per foot (USG75). USG75
reports the results of six pumping tests in the lower carbonate aquifer. The average calculated
transmissivity was 13,000 gallons per day per foot.
Porosity -
USG75 reports that total porosity determinations were made for 16 samples of the lower
carbonate rocks. Total porosities ranged from 0.4 to 12.4 percent with an average of 5.4 percent.
Fracture porosity .of the rock is estimated to range from 0 to 12 percent of rock volume.
Effective Porosity
Due to the extremely low matrix permeability of the carbonate rocks, effective porosity can be
approximated as,the effective porosity of the fractures. Many of the fractures in the carbonate
units are partially filled with clay or other materials which reduce both fracture permeability and
7-91
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effective porosity. USG75 reports that effective
lower carbonate rocks ranged from 0.0 to 9.0
porosity values determined for 25 samples of the
percent, with an average of 2.3 percent. '
Storage Properties
USG75 reported that, based on examination of rock cores, the effective fracture porosity of the
lower carbonate aquifer is probably a fraction of one percent; accordingly, the storage coefficient
under unconfined conditions is not likely to exceed 0.01. Because of the extremely low effective
porosity of the carbonate rocks, the specific storage under confined conditions probably ranges
between 10"5 and 10'6 per foot. Where the aquifer js several thousand feet thick the storage
coefficient may be as large as 10"3. j
Recharge and Discharge
Direct areal recharge to the carbonate aquifer occurs where these rocks are exposed at the
surface. The highest amounts of areal recharge are expected to occur in highland areas where
precipitation levels are highest and where the highly fractured rocks are exposed at the surface.
Recharge to the carbonate units may also derive from downward infiltration through overlying
volcanic or alluvial deposits. The relationship of fjlow potential in the carbonate aquifer to that in
the overlying units is not well known. No downward gradients have been measured between the
carbonate aquifer and overlying units in the study area. This would seem to indicate that the
recharge areas for the carbonate aquifer are located relatively far away from Yucca Mountain.
North of the proposed repository area is an area of [relatively high hydraulic gradient, measured in
the saturated volcanic rocks. One proposed explanation for this high hydraulic gradient is an
inferred east-west striking graben which provides a conduit for ground water flowing in the
i
volcanic aquifer to drain into the underlying carbonate aquifer (FRI94). If this is the case, then
the carbonate aquifer is being recharged by flow from the overlying volcanic units at this
location.
The only measurements of potential in the carbonate aquifer made near Yucca Mountain indicate
vertically upward hydraulic gradients over wide arbas of the carbonate unit. Over at least part of
the study area (in borehole UE-25 p#l) and beyond (specifically in the Amargosa Desert east of
the Gravity and Specter Range Faults), upward hydraulic gradients have been measured between
the carbonate aquifer and overlying units. These upward hydraulic gradients indicate the
potential for upward flow, but do not demonstrate that such flow is occurring in these areas.
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Discharge from the carbonate aquifer would occur in those areas where such flow actually
occurs. FRI94 describes anomalously high ground water temperatures measured beneath Yucca
Mountain in the saturated volcanic aquifer which indicates upward flow (discharge) from the
carbonate aquifer into the overlying volcanic units may be occurring in the vicinity of the
Solitario Canyon Fault.
One major discharge location for flow in the regional carbonate aquifer is at Ash Meadows,
located southeast of Yucca Mountain. It is not clear, however, whether discharge at Ash
Meadows includes any ground water that has flowed beneath Yucca Mountain (this point is
discussed in more detail in Section 7.1.2.3). Additionally, Death Valley, located about 60
kilometers south-southwest of Yucca Mountain, is regarded by many researchers as the base
level or terminus for the entire regional system and, as such, accommodates discharge from the
carbonate aquifer (USG88a). There are also numerous small, relatively low flow springs located
throughout eastern Nevada, though to a lesser extent in the study area, which represent discharge
points from the carbonate aquifer(s) (USG75).
Alluvial Aquifer
Valleys, topographic basins, and other topographic and structural lows are filled with variable
thicknesses of unconsolidated, often poorly-sorted sand and gravel deposits. Lacustrine and
eolian deposits are found locally. Basin-fill deposits are generally 2,000 to 5,000 feet thick, but
in some basins exceed 10,000 feet in thickness. Basin-fill ground water reservoirs are restricted
in areal extent, generally being bounded on all sides by mountain ranges. Beneath the central
parts of the deeper valleys, the water table is encountered in the alluvium. At and near the valley
margins, the alluvium is relatively thin and the water table occurs in the underlying consolidated
rocks. ;
In the Yucca Mountain area, several basin-fill aquifers or potential aquifers exist. These are:
Crater Flats, west of Yucca Mountain; Jackass Flats, east of Yucca Mountain; and Amargosa
Valley, located south of Yucca Mountain. The Amargosa Valley aquifer is substantially larger
and more significant as an aquifer than the Crater Flats and Jackass Flats basins (USG91a).
Farther to the south, across the Funeral Mountains, lies the Death Valley alluvial aquifer.
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Aquifer Geometry
The intermontane alluvial basins tend to be elongated in a north-south direction and are of
roughly the same dimensions as the mountain ranges that separate them (FIE86). The alluvial fill
thickens toward the center of the basins. The Crater Flats and Jackass Flats alluvial basins are
bounded on their northern sides by mountainous aireas at approximately the latitude of the north
end of Yucca Mountain. Crater Flat is bounded at its southern end by a small, southeast trending
ridge of rock outcrops. Topographic map patterns and satellite photographs (DOE95g) suggest
that the Crater Flat Basin may be closed. The Jackass Flats basin does not have a well-defined
southern terminus; it appears to have an outlet at its southwestern end which merges into the
larger, northwest trending Amargosa Desert Basin|. The Amargosa Basin is bounded on its
northwest end by the Bullfrog Hills and on its southwestern boundary by the Paleozoic carbonate
sequences of the Funeral Mountains. Both the Crater Flats and Jackass Flats alluvial basins are
bounded below by their contact with Tertiary volcanic rocks (USG88b; USG83). South'of Yucca
Mountain, the volcanic sequence thins and probably pinches out (USG85a). If so, alluvial
deposits may rest directly on top of Paleozoic carbonate units in the southern part of the basin.
As previously described, two oil exploration wells drilled in the Amargosa Desert, near the town
of Amargosa Valley, went through sedimentary (mostly alluvial) deposits into the carbonate
aquifer. The thickness of the alluvial deposits at these wells was 259 m and 671 m, respectively
(See Table 7-6). The exact nature of the sediments through which these wells were drilled is not
clear, as drilling logs were not examined. DRI94 refers to the sediments both as "alluvium" and
as "Neogene." Czamecki and Wilson (HST91, p. 22) refer to deep (600 m) boreholes in the
south-central Amargosa Desert which terminated in "Tertiary basin-fill sediments" underlying
the Quaternary alluvial fill, thus opening the possibility that the Quaternary alluvial basin-fill
sediments do not directly overlie the Paleozoic carbonate sequence, but are instead separated
from it by an unknown thickness of undifferentiated Tertiary sediments.
i
Thicknesses of the deposits in the three alluvial basins in the study area are not well known due
to the scarcity of drill holes that penetrate the entire alluvial sequence. Two drill holes in Crater
Flat (USW VH-1 and USW VH-2) penetrate through the alluvial cover into volcanic rocks.
Thickness of the alluvium in drill hole USW VH-2 is approximately 305 m, with a depth to water
of 164 m. In Jackass Flats, Well J-13 penetrated approximately 137m of alluvium prior to
entering Tertiary volcanic rocks; the alluvium was not saturated at this location (USG83). Most
of the wells drilled in the Amargosa Valley are water wells for irrigation and water supply. Since
most of these wells encountered sufficient water ii} the alluvium, drilling was not carried through
I
7-94
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to the underlying units; thus, direct evidence for the thickness of the Amargosa Basin alluvial
deposits is lacking. Indirect evidence (geophysical methods) indicates that the thickness of the
alluvial cover in the southern Amargosa Desert may be as much as 1,585 m (USG89).
Saturated thickness and depth to water varies considerably among basins and within a given
basin. In basins where significant discharge areas exist (typically manifested as dry lakes or
playas), depth to water may be only a fraction of a meter to a few meters. Other alluvial basins
may have no saturated zone at all. hi the Amargosa Basin, south of Yucca Mountain, the water
table in some irrigation wells is about 56 m deep. Considering that the basin maybe over 1500
m deep, the thickness of the saturated zone in the Amargosa Basin could be over 1500 m. A
study conducted in the Amargosa Basin area (USG89) concluded that at least 85 percent of the
alluvial thickness in the Amargosa Basin is saturated.
Hydraulic Conductivity
USG75 reports the results of several single well pumping tests in alluvial aquifers at the Nevada
Test Site. These wells are located outside of the area studied for the Yucca Mountain Project,
but the formations tested are broadly similar, and the results are generally applicable to alluvial
deposits within the immediate area of concern. These authors found the hydraulic conductivity
of the alluvial deposits to range from 0.020 to 2.84 m/d. Due to the discontinuous nature of
individual lenses or units within alluvial fill, hydraulic conductivity is expected to show wide
variations in magnitude.
Porosity
The sediments which comprise the alluvial fills are typically coarse grained and poorly sorted,
most of them having been deposited by flash flood conditions over many thousands of years.
Although sediments such as these characteristically have relatively large total porosities,
measured porosities tend to be highly variable due to their poorly sorted nature. USG75 reports
that the total interstitial porosity of 42 samples of valley fill range from 16 to 42 percent and
averaged 31 percent. Caliche, where present, would reduce porosity, perhaps significantly.
USG75, p. 37, reports that caliche is a common cementing material at all depths in a shaft sunk
in alluvium in the northwestern part of Yucca Flat to a depth of 550 feet.
7-95
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Effective Porosity
i
Poorly sorted sediments often have values of effective porosity that are substantially less than
their total porosity. Given the grain size and poorly sorted nature of the alluvium, effective
porosity values may range from a few percent to perhaps as much as 25 to 30 percent.
Storage Properties '
NDC63 estimated specific yield for the alluvial deposits in the Amargosa Basin using grain size
i
distribution methods. The estimated average specific yield for this basin is 17.34 percent; actual
values ranged from not less than 10 percent to not greater than 20 percent (NDC63).
Recharge and Discharge ;
There are several potential sources of recharge for the alluvial aquifers in the vicinity of Yucca
Mountain. One source is direct recharge from precipitation falling on the alluvial areas.
Recharge is also derived to some extent from infiltration of intermittent surface waters of the
Amargosa River and washes draining off the mountains (SAV94). A third source of recharge to
alluvial aquifers is infiltration or leakage from underlying bedrock aquifers. Human activity may
also provide a source of recharge to the aquifers, chiefly by return infiltration of irrigation and
percolation of sewage or wastewater. The primary method of estimating recharge in the alluvial
aquifers is to calculate discharge from the aquifer, most of which occurs as evapo-transpiration at
playas, and to assume inflows are equal to outflows. NDC63 and USG85e provide details of
calculation methods and estimates of recharge for|the Amargosa Basin; values are discussed in
Sections 7.1.2.3 and 7.1.2.4. ;
The nature and relative importance of potential recharge sources to the Amargosa Desert alluvial
aquifer is a matter of some debate. Perhaps the major source of recharge to the alluvial aquifer is
lateral flow into the alluvial deposits from the thirining volcanic aquifer to the north (USG86).
This is contradicted by USGSSf, which uses ground water geochemical data to argue that
"ground water in the west-central Amargosa Desert ....was recharged primarily by overland flow
of snowmelt in or near the present-day stream channels, rather than by subsurface flow from
highland recharge areas to the north," and that "much of the recharge in the area occurred during
Late Wisconsin time" (USG85f, p Fl). This conclusion fails to account for the eventual fate of
water in the volcanic units to the north and is probably too restrictive. ••
i
7-96
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The upward hydraulic gradients measured in the lower carbonate aquifer support the idea that
much of the outflow from the volcanic aquifer moves into the alluvial aquifer. Although this
outflow presumably occurs, somewhere between Yucca Mountain and Amargosa Valley, the
potentiometric surface, at the scale at which it is currently mapped, provides little indication as to
how or where this transition occurs. A recent study, using streamflow data and a modified
version of the HYMET model for the Amargosa River, suggests that the alluvial aquifer may also
be receiving recharge via upward flow from the carbonate aquifer (INY96).
USG91a shows water level altitude maps for 1950's (predevelopment) conditions in the
Amargosa Desert. Comparison of this map with more recent (1987) water level altitude maps
indicates that aquifer development may have had a significant impact on water levels and flow
directions. Pumping of the alluvial aquifer may have induced upward flow from the underlying
lower carbonate aquifer into the alluvial system. The extent to which areal recharge occurs via
infiltration of present-day precipitation falling directly onto the alluvial valleys is thought to be
minimal. This is because of the infrequent rainstorms and the shallow depths to which rainfall
soaks into the desert soil during such events. After a rainstorm, much of this water rapidly
evaporates back into the atmosphere (USG85f).
Several potential modes for natural discharge from alluvial basins exist, including interbasin flow
to other alluvial basins; leakage to the underlying units, either volcanic or carbonate; and
evapotranspiration (NDC63). Discharge from the alluvial aquifers also occurs in the form of
ground water withdrawals by pumping. In the Amargosa Valley alluvial basin, ground water is
pumped for domestic and irrigation purposes (USG91a). Quantitative estimates of recharge and
discharge from the Amargosa alluvial basin are discussed in more detail in Section 7.1.2.4.
Potentiometric and hydrochemical data indicate that the Alkali Flat (also known as the Franklin
Lake Playa), located in the southern end of the Amargosa Desert, is a major discharge area for
the alluvial aquifer system. Estimated discharge at Alkali Flat is about 10,000 acre-feet per year
(DOI63). Discharge at the playa occurs primarily through evapotranspiration, the principal
component of which is bare-soil evaporation (USG90b). Some ground water may flow beneath
the mountain at the south end of the playa and continue southward (USG96a). Regional water
table maps of the alluvial aquifer (see USG91a) also suggest that a portion of the flow in the
alluvial aquifer may be moving southwest through the abutting carbonate rocks of the Funeral
Mountains, and discharging into Death Valley. The extent to which this occurs is unknown.
7-97
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7.1.2.3 Regional Ground Water Flow and Hydrology
The nature of regional ground water flow in the Yucca Mountain area is governed by the
complex three-dimensional nature of the geological and structural units through which it flows.
As previously described, the geological setting in mis area involves a basement of Paleozoic 1
sedimentary rocks which have been complexly folded and faulted. The Paleozoic sequence is
overlain in many areas by a thick section of volcanic rocks and/or alluvial basin fill deposits.
The Paleozoic and volcanic sequences have been disrupted by faults which have juxtaposed
various units against one another and created the basin and range structure. The resulting
geological and stratigraphic complexity creates a correspondingly complex regional ground-
water flow system.
i
Key to understanding regional ground water flow in this area is the concept that the large-scale
flow system may comprise up to three coexisting ground water flow subsystems: local,
intermediate, and regional. These subsystems exist one on top of the other, as well as side by
side. This concept is illustrated in Figure 7-23. The coexistence of such subsystems means that
deep regional flow can pass beneath shallow local jareas of high permeability and that the
presence of hydraulic barriers or variations in permeability can cause appreciable discharge
upgradient from the hydraulic terminus of the system. Major flow systems in the Great Basin are
defined by the dominant flow system, whether it be local, intermediate or regional. Where
consolidated rocks are permeable enough to afford significant identifiable hydraulic continuity on
a regional scale, the local and intermediate types of systems are considered to be subsystems with
major regional flow systems. Boundaries betweeni systems are only generally defined; some may
represent physical barriers to flow, such as masses of intrusive rocks, while others represent
ground water divides or divisions where an area of parallel flow ultimately diverges
downgradient.
I
Regional Ground Water Flow Systems in the Yucca Mountain Area
The Great Basin is considered to consist of 39 "major flow systems" (USG93b). The study area
is located within the Death Valley Ground Water Flow System (DVGWS) which covers an area
of 15,800 square miles (40,100 km2) in Nevada and California (Figure 7-24). The boundaries of
the DVGWS are not precisely known; traditional lateral boundaries are topographic divides that
7-98
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Gaining reach, net gun from ground-water inflow although in localized areas stream may recharge wet meadows along flood plain.
Hydraulic continuity is maintained between stream and groundwatcr reservoir. Pumping can affect streamflow by inducing stream
recharge or by diverting groundwater inflow which would have contributed to streamflow.
Minor tributary streams, may be perennial in the mnnntam« but become losing ephemeral streams on the alluvial fans. Pumping will not
ilffectthe flow of theses streams became hydraulic continuity is not maintained between streams and the principal groundwater reservoir.
These streams are the only ones present in arid
JLosing reach, net loss in flow due to surface water diversions and seepage to groundwater. Local sections may lose or gain depending on
hydnuiHc gradient between stream and groundwater reservoir. Gradient may reverse during certain times of the year. Hydraulic con-
tinuity is maintained between stream and groundwater reservoir. Pumping can afiect streamflow by inducing recharge or by diverting
irrigation return flows.
Irrigated area, some return flow from irrigation water recharges groundwater.
Flood plain, hydrologic regimen of this area dominated by the river. Water cable fluctuates in response to charges in river stage and
diversions. Area commonly covered by phreatophytes (shown by random dot pattern).
Approximate point of maximum stream flow.
Figure 7-23. Schematic Illustration of Ground Water Flow System in the Great Basin
(USG76a)
7-99
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31' H
JS' K
^*.N%VV^
*\.. j •» ^ "
; —^ » -, r «^? •. - -.
i' •. f^> i'*'-£"'*$k: '"
-------
may be physical barriers to ground water flow or may coincide with ground water mounds
formed by local recharge. Rarely, however, are these boundaries true hydraulic barriers.
The DVGWS is further subdivided into a small number of hydrogeological subareas or basins.
Yucca Mountain is located within the Alkali Flats-Furnace Creek Ranch subbasin (Figure 7-25).
Definition of the hydrologic boundaries of the basins is greatly hindered by the complexity of the
geologic structure, the limited potentiometric data, and most critically, the interbasin movement
of ground water through the thick and aerially extensive lower carbonate aquifer (USG75). The
basin covers an area of about 2,800 mi2 and was named after the two major discharge areas near
its southern end (USG82c). The principal aquifers in the northern part of the subbasin are
volcanic aquifers; valley-fill and carbonate rock aquifers dominate in the southern part. The
subbasin receives water from recharge within its boundaries and probably also receives water as
underflow from adjoining subbasins. Ground water leaves the subbasin as evapotranspiration at
discharge areas or as interbasin outflow (USG96a). Alkali Flat is an area where ground water
discharge occurs almost entirely through evapotranspiration. The other major discharge is
thought to be from springs near Furnace Creek Ranch, near the headquarters of the Death Valley
National Monument. A 1984 study (USG84g) estimated discharge from the subbasin at about
15,600 acre-ft/yr; of this total, about 10,000 acre-ft/yr discharges at Alkali Flat and the remainder
discharges from springs and as evaporation near Furnace Creek Ranch in Death Valley. More
recent work (HST91) developed a conceptual model that excluded the Furnace Creek Ranch
discharge area from the shallow flow system that includes Yucca Mountain. HST91 reported
that a ground water divide could exist in the Greenwater and Funeral Ranges between the
southern Amargosa Desert and Death Valley. Such a divide, if it exists, could limit discharge
from the shallow flow system in the Amargosa Desert to the Furnace Creek Ranch area, although
it would not necessarily affect the deeper flow system that may also contribute discharge to the
Furnace Creek Ranch area.
Adjoining the Alkali Flats-Furnace Creek Ranch subbasin to the east is the Ash Meadows
subbasin. These subareas are separated by an irregular north-south line which runs east of Yucca
Mountain. In general, ground water flow in these basins is considered to originate from recharge
in the upland areas of the basin and to move in a southerly direction toward discharge points in
alluvial basins located in the southern parts of the basins. The southern portion of the boundary
between the Alkali Flat-Furnace Creek Ranch subbasin and the Ash Meadows sub-basin is
located along a line of springs (Ash Meadows) which coincides with the trace of a buried fault.
7-101
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J7-I5-N -
JVOO-N -
M'30'N
YUCCA MOUNT *IM
NffTEt NOBTKIW BCOOWY OF /tKAtl FLAT FURNACE SUBBJ»5N S
fLOW BOONOADY vnTM UMXRTLO* FROM PAXUTC «NO ««NCR UES*S
Figure 7-25. Alkali Flat-Furnace Creek Raiich Ground Water Subbasin (USG96a)
7-102
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This fault causes water to rise to the surface by juxtaposition of permeable and impermeable
units of the Paleozoic rocks. Subsurface outflow into the Alkali Flat-Furnace Creek Ranch
subbasin is probable, especially in the vicinity of the buried fault. Geochemical and
potentiometric data suggest leakage of water from the carbonate aquifer into the alluvial aquifer
east of the fault line (USG85f). The degree of connectedness of the two subbasihs may be more
significant than localized leakage across the bounding fault. USG96a suggests that "deep
hydraulic connection through the carbonate aquifer may connect the Ash Meadows area on the
east side of the Amargosa Desert to the Furnace Creek Ranch area of Death Valley. This
possible connection is consistent with the observation that the hydrochemistry of water from
springs that discharge at Furnace Creek Ranch is similar to the hydrochemistry of water
discharging at some springs in the Ash Meadows area. This similarity in hydrochemistry allows
the possibility of westward ground water flow through deep aquifers beneath the Amargosa
Desert, whereas flow through the shallower aquifers seems to be predominately southward"
(USG96a).
Ground Water Flow Directions and Potentiometric Surfaces
Within the DVGWFS, recharge from precipitation probably occurs at Timber Mountain, Pahute
Mesa, Ranier Mesa, Shoshone Mountain, and the Spring Mountains. In the vicinity of Yucca
Mountain, infiltration of runoff in Forty Mile Canyon and Forty Mile Wash probably contributes
to recharge. On a regional and subregional scale, ground water is generally considered to flow
from these recharge areas to discharge areas located at the southern end of the flow system
(USG75). Much of the ground water which travels beneath Yucca Mountain probably discharges
at Alkali Flat (Franklin Lake) in the southern Amargosa Desert and/or in the springs on the
eastern side of Death Valley. Death Valley is the ultimate ground water discharge area and is a
closed basin; no water leaves it as surface or subsurface flow (USG96a). Numerous workers
have constructed potentiometric surface maps for this area, including USG75, USG82c, USG84f,
USG91a, and USG94a. Availability and quality of potentiometric data for the subbasin are
highly variable. Wells are irregularly distributed throughout the subbasin; the greatest density of
wells is on Yucca Mountain itself and in the Amargosa Valley. Data are almost entirely lacking
in the mountainous recharge areas north of Yucca Mountain. In the immediate vicinity of Yucca
Mountain itself, numerous wells have been drilled to the saturated zone and the potentiometric
surface is well-characterized. The potentiometric surface in Amargosa Valley and in the vicinity
of Alkali Flat is also relatively well defined by numerous irrigation and monitoring wells. There
are almost no potentiometric data available in the Greenwater and Funeral Ranges, which bound
the Amargosa Desert on its southwestern side. Figure 7-26 shows the regional potentiometric
7-103
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surface for the DVGWFS. The following sections discuss in detail the nature of the
potentiometric surfaces in each of the three main aquifer types.
Volcanic Aquifer i
I
The lateral extent of the volcanic rocks that make [up Yucca Mountain is not well defined,
primarily because the volcanic units are buried beneath alluvial deposits in the topographically
low areas. South of Yucca Mountain, the volcanic section is believed to thin and pinch out
somewhere in the vicinity of Lathrop Wells (USG;85a, DOE94b). Where the volcanic unit is not
present, alluvial deposits presumably directly overlie Paleozoic sedimentary rocks. Where the
volcanic units thin south of Yucca Mountain, ground water flowing in the volcanic aquifer
discharges horizontally into the adjoining alluvial deposits and continues to flow in a southerly
direction beneath the Amargosa Desert.
At the scale of Yucca Mountain, there are significant variations from the regional flow pattern,
resulting in local ground water flow with a strong ^easterly component. The potentiometric
surface beneath Yucca Mountain has been relatively well-characterized. Potentiometric surface
maps are presented in USG95a, USG94a, and USG84f, among others. The potentiometric
surface can be divided into three regions: (1) a sm'all-gradient area (0.0001) to the southeast of
Yucca Mountain, (2) an area of moderate^gradient (of about 0.015) on the western side of Yucca
Mountain, where the water level altitude ranges from 775 to 780 m and appears to be impeded by
the Solitario Canyon Fault and a splay of that fault, and (3) a large-gradient area (0.15 or more) to
the north-northeast of Yucca Mountain, where water level altitudes range from 738 to 1,035 m
(USG94a). Numerous theories have been proposed to explain the presence of the three domains
and especially the cause of the large gradient area, where water levels decline by more than
900 feet over a distance of slightly greater than one mile. The position of the large gradient area
does not correlate well with any observed geologic feature in the upper 1,500 feet of the
mountain (FRI91). The area where the gradient has been defined is about 1.7 miles north of the
design repository. If the gradient is caused by a barrier to ground water flow, it could be of
particular importance to the design and performance of the repository; an increase in the
permeability of such a barrier could cause a substantial rise in water table altitude in the area of
the proposed repository. A rise in the water table would decrease the thickness of the unsaturated
zone beneath the repository and decrease ground water travel time from the repository to the
accessible environment (SIN89).
7-104
-------
JS* N
YUCCA «OO«TA*I
Of D£*fM
t!? 3ASM
Figure 7-26. Potentiometric Surface in the Death Valley Ground Water Flow System (USG96a)
7-105
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Possible causes of the large gradient other than the flow barrier include, but are not limited to: a
fault or fault zone; an intrusive dike; a change in lithologic facies or a pinch-out; a change in
fracture orientation, density, aperture, or fracture fillings; perched water zones; or some
combination of the above phenomena. Fridrich et1 al. (FRI94) have proposed two models for the
large gradient zone, integrating geologic, geophysical and geochemical evidence to support their
analysis. These and other authors interpret a northeast trending gravity low and drill hole data to
indicate the presence of a buried northeast striking graben (a downdropped block of rock
bounded on both sides by faults) immediately south of the water table decline. The large gradient
zone is coincident with the northern bounding fault of the proposed graben. The presence of the
i
northern bounding graben fault, which is not exposed at the surface and is not known to have
been encountered in any drill holes in Yucca Mountain, is central to both models proposed.
Briefly, the first conceptual model proposes that tljie buried fault zone provides a permeable
pathway through the volcanic section into the underlying deep carbonate aquifer. The second
model has the buried fault acting as the northern boundary for a much thicker and more
transmissive volcanic section south of the buried fault. These authors also suggest that rapid
draining of water in the large gradient zone may cause the low gradient area to the south and
southeast. In this model, the small gradient zone may result partly from a reduced ground water
flux in the volcanic rocks due to the capture of flow by the underlying deep carbonate aquifer.
Carbonate Aquifer j
The lower carbonate aquifer has a maximum thickness of about 8,000 m. Because the carbonate
aquifer in the study area is overlain by thick deposits of volcanic rocks or alluvium, flow
directions and gradients are not well-defined. Regional ground water flow through the lower
Paleozoic aquifer is considered to be generally-southward. Small-scale potentiometric surface
maps are presented in USG75. The lower carbonate aquifer is present below Yucca Mountain at
a depth of about 1,000 m and extends southward bjelow the Amargosa Desert into Death Valley.
There are a very limited number of holes that penetrate the lower carbonate aquifer beneath the
valley fill. Much of the physical knowledge of the system is based upon studies of the outcrop
areas, most of which are in the mountain ranges. The best interpretation of available geological
data indicates that the lower carbonate aquifer is continuous from beneath Yucca Mountain to
Death Valley and is a potential pathway for radionuclide transport in what appears to be the
ultimate discharge point for the aquifer in Death Valley.
7-10,6
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The extent of hydraulic communication between the volcanic and underlying Paleozoic sequence
is not well characterized. In the only well (UE-25p#l) at Yucca Mountain which penetrated into
the Paleozoic sequence, an upward hydraulic gradient (from Paleozoic to the Tertiary) was
measured. Analysis of earth-tide response of water levels in this well have been interpreted to
indicate that the carbonate aquifer is well-confined by an overlying low-permeability confining
layer and has a relatively high transmissivity (INY96). Additional evidence, including isotopic
composition and temperatures of ground water beneath Yucca Mountain, supports the concept
that ground water may be flowing from the Paleozoic aquifer into the volcanic aquifer (USG88c;
STU91).
Alluvial Aquifer
Significant amounts of ground water occur in the alluvial aquifer beneath the Amargosa Desert.
In the Amargosa Valley area, irrigation activity derives all of its water from wells completed in
the alluvial aquifer, some of which yield water at rates of several hundred gallons per minute.
Static water levels are less-man 55 m below the surface in some locations. Figure 7-27, taken
from USG91a, shows a map of the water table in the Amargosa Desert. USG91a also provides a
map of depth to water in the Amargosa Desert. Ground water flow in the alluvial aquifer is
generally perpendicular to the potentiometric contours. The potentiometric contours shown in
Figure 7-27 indicate that the predominant flow direction is to the south. The ground water flow
direction is also roughly parallel to the surface drainage direction. At the southern end of the
Amargosa Desert, low permeability playa and lake bed deposits create locally-confined
conditions. The potentiometric surface at Alkali Flat is in some locations above the ground
surface (USG90b).
The potentiometric surface shown in Figure 7-27 is drawn from 1987 data. Comparison of this
map with water level altitude maps for 1950's (predevelopment) conditions (USG91a) in the
Amargosa Desert indicates that irrigation pumping has had a significant impact on water levels
and local flow directions. Pumping of the alluvial aquifer may also have induced upward flow
from the underlying lower carbonate aquifer into the alluvial system.
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116
36M01
asrao- -
36*20-
36'IQ'
EXPLANATION
—680 Potentiometric contour—
Shows altitude at which water
level would have stood in tightly
cased wells. Dashed where
inferred. Contour interval, in
meters, is variable. Datum is
sea level
5 MILES
5 KILOMETERS
Potentiometric contours
modified from Claassen.
1985
Figure 7-27. Potentiometric Surface in the Amargosa Desert. Ground water flow is generally
south, perpendicular to contour lines. (USG90b)
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Ground Water Travel Times and Radionuclide Transport
The transport of radionuclides in the saturated zone away from a repository depends on a wide
variety of factors including, but not limited to, ground water and host rock geochemistry;
advective ground water velocities; radionuclide concentrations and retardation properties; flux
rates of radionuclides from the unsaturated zone; the presence of sorbing materials such as
zeolites and clays; rock fracture density; fracture-matrix interaction; future climate changes; and
anthropogenic influences. Knowledge of the transport properties in the site-scale and regional
flow systems would allow researchers to more completely address four of the most important
questions surrounding repository performance and regional ground water flow issues in the area
around Yucca Mountain:
1. What path would radionuclides from the repository follow?
2. How fast and how far would radionuclides travel in the saturated,zone?
3. Where would radionuclides become accessible to the biosphere?
4. What will the concentrations of radionuclides be when they become accessible to
the biosphere?
The answer to all of these questions is uncertain. The ability to know or predict the answers to
these questions depends on performing sufficient scientific investigations over the study area in
order to reduce the associated uncertainties to acceptable levels. Some level of uncertainty will
always remain, as it is not possible to completely characterize any underground system.
Recent testing activities conducted at the C-well complex have been designed to provide more
information regarding contaminant transport properties in the saturated zone (DOE96a,
DOE96b). Tracer testing at the C-wells complex has included the injection of both conservative
(non-sorbed/non-decaying) and nonconservative tracers (sorbed). All tracer tests were performed
by establishing a quasi-steady convergent flow field and hydraulic gradient by pumping from
borehole UE-25 c#3 for several days prior to the injection of tracer compounds. Test results are
collected by analyzing samples taken at regular intervals from the pumped well and preparing
"breakthrough curves" which plot the concentration of the tracer in the pumped well versus time.
After the first detection of tracer compound, breakthrough curves typically show an initial rapid
rise in tracer concentration, which then peaks and tails off gradually.
The first tracer test performed at the C-wells used sodium iodide, a conservative solute. Because
it is negatively charged, sodium iodide does not sorb to zeolites and clays, and has an average
7-109 :
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matrix retardation coefficient of 0.93. The retardation coefficient should be less than one
because of a process known as anion expulsion, wherein anions are repelled by negatively
charged grain surfaces and arrive at the recovery well prior to neutrally-charged tracers. Test
' conditions were negatively impacted by decreasing! pump discharge and the resulting nonsteady
hydraulic gradient and flow rates. Tracer recovery [data were analyzed to determine effective'
porosity and longitudinal dispersivity using an analytical solution. The analytical method
I
employed has a high uncertainty and calculated parameters do not represent a unique solution to
the breakthrough curve data. Test data were analyzed using several different sets of assumptions
including a single-porosity solution, a weakly dual-porosity solution, and a moderately dual
porosity solution. [
In a single-porosity solution, calculated fracture porosity was 0.036 and longitudinal dispersivity
was 17.00 ft. In a weakly dual-porosity solution, calculated matrix porosity was 0.032, fracture
porosity was 0.0068, and longitudinal dispersivity Svas 20.75 feet. In a moderately dual-porosity
solution, good matches were obtained using a matrix porosity of 0.0778, a fracture porosity of
0.0237 and a longitudinal dispersivity of 13.64 feet. It is important to recognize that parameters
used in analyzing tracer recovery data have a high degree of uncertainty and that because the
ground water flow field at the C-wells is anisotropie, the transport field is most likely anisotropic
as well. i
[•
;
Subsequent to performing the conservative tracer test, two additional pilot tracer tests were
performed. Both tests were conducted in the 100 meter thick isolated interval within the Bullfrog
member of Crater Flat Tuff. This interval has the largest hydraulic conductivity of any interval at
the C-holes. The objectives of these tests were to determine: (1) which injection well (c#l or
l
c#2) would result in a higher peak concentration of a conservative tracer, and thus be a better
injection well for a reactive tracer test, and (2) what minimum mass of lithium bromide would
have to be injected to conduct a successful reactive tracer test. Both pilot tests were successful in
that they clearly identified that Well c#2 is the preferred injection hole for a reactive tracer test
and that at least 80 kilograms (kg) of lithium bromide would be needed to ensure a successful
test. The analysis of these tracer tests and any subsequent tests for transport parameters is not
currently available. I
The current state of knowledge suggests that ground water beneath the proposed repository
moves laterally downgradient until the volcanic aquifer pinches out, at which point it discharges
laterally into the alluvial aquifer. Radionuclides dissolved in ground water would potentially
i
follow a similar path. Much of the ground water that enters the alluvial aquifer currently moves
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southward to the primary discharge location at Alkali Flat. Other actual or potential points of
discharge for the system include water wells in the Amargosa Desert and springs in the Furnace
Creek Ranch area of Death Valley.
Ground water travel times to any of these locations are not well known. Estimates of ground-
water travel times can be developed by simple calculations or by more sophisticated numerical
modeling. In either case, travel times calculations are based on hydraulic gradient, hydraulic
conductivity, and effective porosity of the formation through which the water is flowing. Of
these three parameters, hydraulic gradients are probably the best known and most easily
measured. A range of ground water travel times in the Tertiary volcanic aquifer has been
developed in support of DOE's Total System Performance Assessment conducted in 1993.
TSPA93 predicted a range in advective velocities between 5.5 and 12.5 m/yr. These velocities
represent average velocities in the Tertiary volcanic aquifer between the footprint of the potential
repository and a 5 km "accessible environment" located to the south and east of the potential
repository (DOE95f). Performance assessment parameters and results are more fully described in
Sections 7.3 and 7.4.
A more recent study on radionuclide transport in the saturated zone (DOE96c) co'ncluded that an
advective travel time of five m/yr is in the middle of the range of reasonable estimates. At this
velocity, unretarded radionuclides would take approximately 1,000 years to travel five km from
the repository and 5,000 years'to travel 25 km from the repository. This study also documents
the results of preliminary, highly simplified radionuclide transport modeling work performed
using advective velocities of five m/yr. The nature of downgradient breakthrough curves and
resulting peak dose calculations were highly dependent on assumed values of dispersivity. The
study also found that the breakthrough curves, travel times, and peak dose results were strongly
dependent on the retardation properties of individual radionuclides, the presence of sorbing
materials such as zeolites, and the possibility of fracture transport bypassing sorptive horizons
within the volcanic aquifer.
No reliable estimates of advective velocity in the alluvial aquifers have been made downgradient
of the potential repository.
An important unresolved issue is the extent of interaction between the volcanic aquifer and the
underlying carbonate aquifer. The possibility that radionuclides might enter the regional lower
carbonate aquifer, with its higher permeability, raises concerns that radionuclides could be
transported as far as Death Valley. Current evidence, such as hydraulic head measurements in
UE-25 p#l, isotopic data, and saturated zone temperature anomalies suggests that the lower
7-111 :
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carbonate aquifer has a higher hydraulic head thanjthe overlying units. This upward gradient
indicates that it is unlikely that radionuclide contaminants will be transported into the carbonate
aquifer in the vicinity of Yuqca Mountain. Velocities through the lower carbonate aquifer range
from an estimated 0.02 to 200 feet per day, depending upon geographic position within the flow
system (USG75). It should be noted that the figures given above are for an area of carbonate
rocks outside, and much larger, than the study area. No data are available regarding actual
ground water flow velocities in the study area. Cajrbonate rocks with solution-widened fractures,
cavities, and caves typically exhibit an extremely large variation in ground water velocities.
Ground water age dating (WIN76) using carbon-ik methods in the springs of Ash Meadows
suggested ages of ground water in the majority of jhe springs ranging from 19,000 to 28,000
years. INY96 describe more recent studies which indicate that water may move through the
lower carbonate aquifer in times less than 1000 to 2000 years.
7.1.2.4 Ground Water Resources and Utilization
Many of the studies performed in the Yucca Mountain characterization process have thus far
focused narrowly on the immediate area in and around the proposed repository. Few studies to
date have attempted to present a regional picture of ground water resources for the areas
downgradient from Yucca Mountain. This section presents a summary description of water
resources in the area downgradient (generally south) of Yucca Mountain.
I • ;
Water Quality
Volcanic Aquifer
i
The chemistry of water flowing through the volcanic aquifers exhibits complex dependency upon
rock composition, residence time in the aquifer, and position along a flow line (USG75). Ground
water chemistry in a volcanic rock is controlled by primary glass, pumice fragments, and the
diagenetic minerals (NAN89). Water samples fropn wells drilled in Yucca Mountain indicate
that the water is predominantly a sodium bicarborjate water containing small concentrations of
silica, calcium, magnesium, and sulfate (USG83).! Sodium levels are generally elevated in these
rock types due to the presence of volcanic glass, which is not stable in the presence of water and
contains appreciable sodium. Two water wells, J712 and J-13, currently supply water for site
characterization activities at Yucca Mountain and; have been pumped extensively for decades
with no signs of deteriorating water quality (USG83; USG94b). (Additional sources of
7-1)2
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information regarding ground water chemistry can be found in USG86, USG84d, USG91b,
USG91c, andUSG93a.)
With the exception of substances deliberately introduced into wells during drilling and testing,
such as drilling fluids (including diesel fuel at Well J-13 (USG83)) and radioactive tracers
(Iodine-131; USG93a), no anthropogenic effects on water quality are observed in the volcanic
rocks. This is attributed to the relatively low levels of human activity and the presence of a thick
unsaturated zone with long travel times for infiltration to reach the saturated volcanic rocks.
Alluvial Aquifer
The chemical quality of the ground water in the saturated alluvial deposits varies from place to
place. In general, ground water in wells closer to Yucca Mountain is of better quality than near
the ultimate discharge areas of the system, such as the southern Amargosa Desert and Death
Valley. Ground water near these latter areas contains higher concentrations of dissolved
constituents and is less suitable for most purposes (NDC63). NDC63 states that "although the
chemical quality of ground water in the Amargosa Desert may be suitable generally for irrigation,
water of median salinity is common and water of high salinity occurs locally." Ground water in
the alluvial aquifers in many cases contains excessive concentrations of fluoride; a dental
examination of school children in Beatty found that 19 out of 20 children who lived in Beatty
since birth were affected with dental fluorosis (NDC63). (See USG94b and USG91d for
additional ground water chemical quality data for the alluvial aquifer.)
Carbonate Aquifer
In general, water occurring in the carbonate rocks is a calcium and magnesium carbonate water.
Where water in the carbonate aquifer has moved through the overlying volcanic rocks, analyses
show increased levels of sodium and potassium (USG75). See USG84c for chemical analyses of
water from Well UE-25 p#l completed in the carbonate aquifer beneath Yucca Mountain.
7.1.3 Climate Considerations
For the purposes of this document, climate is defined as the ensemble of weather conditions over
time. Precipitation and temperature variability are the aspects of climate that are most significant
to the long-term performance of a high-level waste repository at Yucca Mountain. These
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parameters influence, directly and indirectly, water infiltration rates in the area of the proposed
repository.
"Variability" means the timing, rates of change, magnitude, and persistence of conditions.
Inferences about variability are based on studies of past conditions in the region, as recorded by
both geological and biological paleo-environmental indicators. Computer models of the
atmospheric circulation are used to simulate both past and future climatic regimes. Modelling
results are compared to paleoclimate data. The better their simulations of past climatic
conditions, the more confidence scientists and policy makers will have in the ability of models to
predict future climate. Thus, paleo-data are considered essential in assessing future climates.
The impact of human interference with naturally-occurring climate variations must also be
considered. Large-scale changes in atmospheric composition have occurred and are almost
i
certain to continue for the next several thousand years (HOU92). General circulation models
may be used to anticipate the consequences of such changes and to help chart the future course of
climate change. Since the concentration of greenhouse gases in the 21st century will likely
exceed anything the world has experienced for millions of years, the paleoclimate record may not
fully define the climate of the future. Unknown feedbacks or abrupt, rare changes in the climate
system may occur in the future. Nevertheless, the [paleo-record, combined with realistic
[
computer models of existing and future climate, provide the best set of tools currently available
to define the potential limits of climate variability (in the Yucca Mountain area.
7.1.3.1 Past Climate Conditions and Variations
Global climate has evolved over glacial to interglaeial time scales in response to changes in
orbital forcing (the relative position of the earth toithe sun, with consequent changes in the
geographical and seasonal distribution of incoming solar radiation). In simple terms, these
i
changes altered the Pole-Equator temperature gradients, which led to changes in atmospheric
I
circulation and the overall hydrological balance ofjthe earth. These changes caused ice sheets to
accumulate on the continents at high latitudes, the jsea level to fall, global temperatures to
decrease, and rainfall patterns in the tropics to shift.
I
i
Changes in incoming solar radiation alone were insufficient to bring these environmental
changes about; they were amplified by internal feedbacks of the climate system itself, most
probably through changes in atmospheric composition and the albedo (reflectivity) of the earth's
surface. Such feedbacks led to reduced levels of carbon dioxide and methane (both greenhouse
7-1 w
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gases); a higher overall albedo for the earth, due to more extensive snow and ice cover; and more
extensive deserts. However, at other times in the cycle of orbital changes, feedback mechanisms
brought about increases in greenhouse gases and other changes in the climate system, eventually
leading to rapid destruction of the ice sheets and abrupt deglaciations. The growth and decay of
ice sheets affected the atmospheric circulation, displacing jet streams equatorward and causing
massive increases in rainfall in previously dry areas.
Southern Nevada and the Great Basin experienced such dramatic changes, which, together with
lower temperatures, led to aquifer recharge and the filling of many closed basins with extensive
lakes. Such changes are evident in geologic features of the region. Variations in lake levels
extending back into the last glaciation are best known; they are generally well-dated and have
been studied in many areas of the western United States. Observed changes are well supported
by a variety of biological evidence, particularly that obtained from the analysis of packrat
middens, which contain discrete samples of local vegetation in the vicinity of the packrat nests
from particular time periods in the past. For example, when lake levels were high, vegetation
was generally more extensive; some areas that are arid today were forested. This can be seen
from the packrat middens, where vegetation can be related to past time periods.
Hydrological changes in the arid western United States do not coincide in detail with the record
of continental ice volume changes. However, it is clear that high lake levels were present when
the Laurentide Ice Sheet was extensive and that water levels fell in association with deglaciation.
As noted by Smith and Street-Perrott, "more than a hundred closed basins in the western United
States contained lakes during the Late Wisconsin [the last episode of the ice ages], 25,000 to
10,000 yr B.P. [before present], but only about 10 percent of the lakes are perennial and of
substantial size today...." Even in today's hyperarid Death Valley, there is evidence that an
extensive lake occupied the basin between 21,500 and 11,900 years ago (SMI83; HOO72).
The longer term record of hydrological variability is much harder to document, given the
problems of dating water levels and precipitation. In addition, it is possible that some paleo-
lakes may have been caused by slight tectonic changes or other geomorphological factors.
Furthermore, rapid changes in ice sheet size, as postulated from sedimentary records in the North
Atlantic and elsewhere, may have resulted in very abrupt changes in the hydrological regime in
the western United States.
If jet stream displacement, due to ice sheet growth and decay, is the principal factor in
hydrological change in the western United States, there is good reason to suspect that a quite
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variable hydrological regime has influenced the region over glacial-interglacial timescales.
Nevertheless, the more prolonged glacial episodes jwere dominated by cooler, wetter conditions,
associated with higher infiltration rates, more vegetation, and the presence of many freshwater
lakes in the Great Basin. Quantifying such changes is difficult, but Spaulding et al. estimate the
limit at the last glacial maximum as approximately 6°C colder, with precipitation levels double
those of today (SPA83).
7.1.3.2 Potential Future Climate Conditions
Orbital variations clearly have driven the broad-scale variations of global climate over the last
several million years, at least. These orbital variations are likely to be a dominant influence in
I
the future. Since the orbital variations are periodic and predictable, their occurrence in the past
and in the future can be calculated. Variations over the past million years have occurred within a
fairly limited envelope; predicted variations for thq future show that, for at least the next
250,000 years, the expected orbital changes will stay well within this envelope. How such
changes will affect climate can be assessed by using the solar radiation changes to force a global
climate model to simulate both past and potential climate variations in the future.
Most studies attempt to reconstruct past changes where the simulations can be verified by
observation, but a few attempts have been made over the past 25 years to forecast future changes,
at varying levels of sophistication. Figure 7-28 shows the results of these efforts, with the overall
parameter describing the output expressed (on the righthand side) in terms of global temperature.
Obviously, the sophistication of such calculations has increased over the years, but most studies
consistently predict that global climates over the next 60,000 years or so will gradually shift
towards a full glacial -mode, similar to that experienced 20,000 years ago during the most recent
glacial period. Indeed, the trend towards such a state began a few thousand years ago, in the mid-
Holocene Period.
The trend towards a glacial extreme is not
generally downward trend in temperature.
indication that conditions like those of today will
the future. It also appears that the "saw-tooth"
cold glacial conditions, followed by abrupt "
continue into the future.
7-116
monotonic, but involves minor oscillations on a
Following the temperature minimum, there is some
not return again until about 120,000 years into
nature of past climate variations—slow declines to
terminations" of glacial conditions—will also
-------
PAST ma BPJ o FUTURE (*o
STUQV 1
ANNUAL RATE
OF CHANGS IN
WIMTCB
tfttMOIATION
STVOY J
ICE
VOUIME
STUDY 3
ICE
SHEET
ICE
VOLUME
STUDY 5
4»0
C%>)
STUDY 8
ACUN t
WARM
COLO
WAHU
COLO
COLO
WAftU
COLO
COLO
WAflU
COLO
STUOTT
ICE
SHEET
VOLUME
COLO
Figure 7-28. Future Climates, Expressed in Terms of Overall Global Temperature Change
Future climates, expressed in terms of overall global temperature change, as
predicted by seven different models driven by changes in orbital forcing. The
boxes on each diagram delimit the last glacial and interglacial extremes. Dates
are in years x 103. (GOO92)
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In general, the present arid climate conditions are expected to be maintained in the future. The
Sierra Nevada Mountains, which lie to the west of Yucca Mountain, have a strong rain-shadow
effect on the Yucca Mountain Region. This effedt is expected to be maintained or enhanced in
the future because the Sierra Nevada range is still increasing in elevation (DeW93).
These are very broad conclusions that do not allow for the high-frequency oscillations,
superimposed on longer term trends, which have been seen in the Greenland ice cores and in
some marine sedimentary records from the North: Atlantic. High-frequency oscillations have
most recently been seen in the Santa Barbara Bas|n (BEH96). Such changes would be expected
to occur in any future glaciation, since they appear to be integrally linked to the dynamics of ice
growth and decay and their impact on ocean circulation (BRO94).
i '
What these models do not consider is the potential additional effects of greenhouse gas increases
on the radiative balance of the earth and, consequently, on the general atmospheric circulation. It
is generally believed that the small insolation changes brought about by orbital changes are
insufficient by themselves to bring about glaciation, or indeed to terminate'glaciations. The
critical issue is the feedbacks, which may amplify the small radiative signal, with the ice sheets
themselves playing a major role (via albedo effects, sea-level change, topographic influences on
atmospheric circulation, effects on ocean thermohaline circulation, etc.). What is not clear is
whether any near-term increase in greenhouse gases (in the next few decades to centuries) would
eventually be overwhelmed by the orbitally-induced shift toward future glaciation or if the
wanner climate would preclude such a development by minimizing the necessary feedback
mechanisms. Broecker (BRO75) termed this near-term warm episode a "super-interglacial"
because it may involve temperatures higher than ^n any recent interglacial period. As "such, it is
difficult to predict what the overall consequences of such a unique state might be for the future
evolution of climate.
;
One study of such a scenario used a 2.5D general computer model to assess both anthropogenic
effects and orbital forcing (BER91). The model Sassumes that the Greenland Ice Sheet will be
entirely consumed in the near term, but that the general direction of long-term climate change
towards glaciation is not changed. The peak timing of the next glaciation is delayed by about
5,000 years (Figure 7-29). However, this model !is still fairly crude and does not incorporate
many of the feedbacks that may be critical in the;evolution of future climate. More experiments
with transient climate simulations, using the next generation of coupled ocean-atmosphere
general circulation models, will be needed to obtain a more sophisticated answer to this question.
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FUTURE
10
20
30
40
50
-140-120-100-80 -60 -40 -20 0
•HME(ka)
NORTHERN
HEMISPHERE
l i 1 i I
0.00
050
1DO
20 40
60 80
AP
8
•33
I
g
3J
i
Figure 7-29. Model Simulations of Past and Future Climate Conditions i
Model simulations (solid line) of past and future climate conditions, expressed in
terms of changing ice volume on the continents, and including anthropogenic
greenhouse effects in the immediate future. Dashed line gives past global ice
volume changes as registered by oxygen isotope ratios in benthic foraminifera
from the oceans (BER91).
At this stage, there is no compelling evidence that the world of the next million years will not be
subjected to the same range of climate variations experienced over the last million years.
However, in the near term (from the next few decades to several thousand years), an enhanced
greenhouse effect will very probably bring about warmer conditions than have been experienced
for thousands, perhaps even hundreds of thousands of years. This was the general conclusion of
experts who were asked to assess the magnitude and direction of future climate change (Figure 3-
11 in DeW93). They estimate that the likely upper limit of a temperature increase in the mean
annual temperature of the Yucca Mountain Region would be about two to three degrees celsius.
Whether this effect will persist for hundreds or thousands of years depends greatly on
assumptions made about future energy consumption patterns and the overall availability of fossil
fuels. If society eventually limits fossil fuel consumption, this warmer episode may come to a
close, with the naturally-occurring trends then becoming dominant. Nevertheless, the possibility
7-119
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that a greenhouse gas-induced "super-interglaciar>may lead to unanticipated pathways in the
climate system and new climate states can not be entirely ruled out (BRO87).
The potential changes of greatest concern at Yucca Mountain are those associated with the
"glacial climate mode" rather than with an "interglacial mode." Past history indicates that wetter
conditions in the region have generally been associated with globally cooler climates, or with
transitions to such climates. Interglacial periods have been arid. Currently, no evidence suggests
that this basic pattern is likely to be different in the future. Hence, the immediate future climate
of Yucca Mountain, dominated by anthropogenic effects, is likely to be as dry or drier than the
present. Eventually, however, cooler and wetter conditions will dominate the area during
persistent glacial climate modes.
I
7.1.3.3 Summary Regarding Climate (
The climate in the Yucca Mountain region is currently warm and semi-arid, with a mean annual
[
average temperature of 16°C (61°F) and mean anriual precipitation of 170 rrim/yr (6.7 in/yr).
Precipitation varies throughout the year, averaging about 18 mm/month in the fall and. winter,
and about 9 mm/month in the spring and summer.
Physical evidence of past climates shows that climate conditions'previously cycled between cold
glacial climates and warm interglacial climates such as the present. Fluctuations averaged about
100,000 years in length. Present climate conditions have prevailed since the last glacial period
about 10,000 years ago. j
Infiltration, into Yucca Mountain, of water from precipitation is a factor of primary importance to
performance of a potential repository at the site. Projections of future climate conditions,
precipitation rates, and infiltration rates are therefore key factors in total system performance
assessments such as are discussed in Section 7.3.
The historical record of climate conditions and clitnate changes in the Yucca Mountain region
was interpreted quantitatively by DOE for modeling of future climate conditions in the Total
System Performance Assessment for the Viability Assessment (TSPA-VA; see Section 7.3.2).
For these performance evaluations, DOE assumed that there would be three characteristic climate
conditions in the future: the present-day dry climate, a long-term-average (LTA) climate, with
precipitation at levels twice the present, and a supisrpluvial climate, with precipitation three times
the current rates. The climate conditions were assumed to alternate in sequence, with average
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durations of 10,000, 90,000, and 10,000 years for the present-day, long-term-average, and
superpluvial conditions, respectively. For the base-case TSPA-VA evaluation of future
repository performance, the present day climate was assumed to continue for 5,000 years into the
future, and the first superpluvial climate period was assumed to occur about 300,pOO years in the
future. ]
For the TSPA-VA performance evaluations, the average annual precipitation rates were assumed
to be 170, 340, and 510 mm/yr, for the present-day, LTA and superpluvial climates respectively.
These precipitation rates were assumed to result in average infiltration rates of 7.7, 42, and 110
mm/yr. The three-fold increase in precipitation rate for the superpluvial climate, in comparison
with the present-day climate, was therefore assumed to result in a factor of 14 increase in water
infiltration into the mountain.
7.2 REPOSITORY CONCEPTS UNDER CONSIDERATION FOR YUCCA MOUNTAIN
7.2.1 Conceptual Repository Systems
Design concepts for a repository at Yucca Mountain have changed and evolved significantly
during the 20 years of site evaluation work to date. Changes have been made in response to
information from sources such as site characterization data, repository system performance
assessments, external technical .reviews, and evolution of a waste isolation strategy. Changes
have occurred in fundamental concepts as well as in design details. For example; the Site
Characterization Plan issued in 1988 (DOE88) envisioned vertical emplacement of waste
packages in individual boreholes in the floor of tunnels; current plans call for end-to-end
horizontal emplacement in long, excavated drifts. The 1988 waste package design concept was a
simple steel canister approximately two feet in diameter with an expected lifetime of 1,000 years
or less; the current design concept is a container about six feet in diameter with two-layer,
corrosion-resistant walls and a lifetime objective of more than 10,000 years. Other changes have
evolved as a result of acquisition of site and laboratory data and from consideration of the results
of total-system performance assessments.
In response to requirements of the Fiscal Year 1997 Energy and Water Appropriations Act (PL
104-782), the DOE performed a Viability Assessment (VA)24 for development of a repository for
The terms Total System Performance Assessment-Viability Assessment and Viability Assessment and the
acronyms TSPA-VA and VA are used interchangeably throughout this report.
7-121
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disposal of highly radioactive wastes at Yucca
policy makers with an estimate of the viability of a
time frame required for decision making.
Mountain. The purpose of the VA was to provide
repository at the Yucca Mountain site in the
The five-volume VA report was released by the DOE in December 1998 (DOE98). The
Department found "...that Yucca Mountain remains; a promising site for a geologic repository and
that work should proceed to support a decision in 2001 oh whether to recommend the site to the
President for development as a repository" (DOE98, Overview).
j
The design concepts used for the VA are described below. DOE considers the VA, and its
repository design features, to constitute a snapshot in time of an evolutionary process leading
potentially to a finding that the site is suitable for disposal and subsequently to a License
Application. Further development of the repository design features and performance evaluation
methodology will be needed for the Site Recommendation and for a License Application if the
site is found to be suitable for disposal. \
I
Design concepts used by the DOE in the Viability Assessment were as follows:
• Horizontal emplacement of waste packages in parallel excavated drifts.
An initial thermal loading on the surroundings corresponding to 85 MTU/acre.
• Emplacement of waste packages only between the Ghost Dance fault and the
Solitario Canyon fault. ;
Disposal of 63,000 MTU of commercial spent fuel and 7,000 MTU equivalent of
various types of defense wastes. A total of 10,500 waste packages would be
emplaced, consisting of 7,642 packages of commercial spent fuel and 2,858
packages of defense wastes. !
i
• Disposal in excavated drifts 5.5 m ih diameter, with a total of about 107 km of
tunnels and drifts in an emplacement area of 740 acres. Drifts would be spaced 28
m apart.
• Packages of commercial spent fuel would contain 21 PWR fuel rod assemblies or
44 BWR assemblies.
Waste package design features which include, for the commercial spent fuel
packages, dimensions of 2-m diameter and 6 m length, with an outer shell of A
516 carbon steel 10 cm thick and an inner shell of corrosion-resistant Alloy 22
that is 2 cm thick. j
i
7-122 '
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Temperature limits of 200°C for the drift walls and 350 °C for the commercial
spent fuel cladding. :
Waste types to be disposed would include uncanistered and canistered commercial spent fuel
assemblies; canisters of vitrified defense high-level wastes; navy spent fuel; other DOE-owned
spent fuel, such as from the Hanford N-reactor; and surplus plutonium from dismantled nuclear
weapons. Most of the commercial SNF is clad with zirconium alloys (Zircaloy-2 and Zircaloy-
4); about 1.15 percent is clad with stainless steel. In the VA, the DOE assumed that the Zircaloy
cladding would act as a significant barrier to radionuclide release. No credit was taken for
stainless steel cladding.
7.2.2 Design Concepts for Engineered Features of the VA Repository
7.2.2.1 Repository and Surface Facility Layouts
The VA reference design for excavation of tunnels and drifts for emplacement of wastes is
shown in Figure 7-30. The repository footprint, which covers about 740 acres, is offset from
both the Ghost Dance and Solitario Canyon faults. The footprint is about 1 km wide and 3 km
long. This layout resulted from consideration of factors such as potential for fault movement,
location of dominant fracture systems in the geologic formations, ease of access during
operations, and the heat emissions and temperature limits assumed as the basis for establishing
design parameters. The location of the repository within Yucca Mountain is shown in cross
section in Figure 7-31.
The VA plan for functions and layout of the North Portal facilities is shown in Figure 7-32.
Plans for South Portal operations and facilities are still under development and were not
addressed in the VA. :
Because of their initial high heat and radiation emissions, emplacement of the waste packages
will be done remotely. As previously noted, the VA design temperature limit for the drifts is
200°C; radiation field levels at the surface of the packages would be on the order of 35-
60 rem/hour.
7.2.2.2 Waste Package Design
7-123
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Waste package designs will be tailored to the characteristics of the waste type (commercial spent
PWR and BWR fuel; U.S. Navy spent fuel; other DjOE-originated spent fuel; vitrified high-level
waste; and immobilized surplus plutonium from nuclear weapons). The dominant types of waste
LEGEND
Avattable repository siting area
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SCAtf. 1:40.000
1cm - 4OOm
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Figure 7-30. Repository Layout for the VA Reference Design (DOE98)
7-124
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Elevation (m)
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packages in the repository will be those for commercial spent PWR and BWR fuel; in the VA
reference design, there would be about 7,600 commercial spent fuel packages, two-thirds of
which would contain PWR spent fuel and one-third BWR spent fuel. Most of the PWR packages
would contain 21 spent fuel assemblies; the BWR packages would contain 44 assemblies (the
BWR assemblies are about half the size of the PWR assemblies). Both types of waste packages
contain about 10 MTHM.
The reference waste package design used in the Viability Assessment for the 21-PWR container
is shown in Figure 7-33 (the BWR package is similar), and the design concept for the defense
high-level waste container is shown in Figure 7-34. A key feature of the designs is use of two
materials to form the walls of the package. The outer material, designated as a Corrosion
Allowance Material (CAM), is A 516 carbon steel. The inner material, designated as a Corrosion
Resistant Material (CRM), is a high-nickel alloy, Alloy 22, which is highly resistant to corrosion.
The CAM is intended principally to provide strength and radiation shielding for the package; the
CRM; is intended to serve as the principal barrier to contact of water with the waste form within
the package. ;
In the VA reference design, the waste packages were emplaced horizontally on concrete inverts
in excavated drifts that were 5.5 m in diameter and lined with concrete. A cross section diagram
of this reference design is shown in Figure 7-35. The drifts were spaced 28 m apart and the
waste packages were spaced about 19m apart in the drifts. Under this design concept, each
waste package acts as a point source of heat emissions for repository performance evaluation
purposes. An alternative design concept is to emplace the packages very close to each other end-
to-end, in which case the performance evaluations treat the packages as a line source of heat
emissions.
The VA also considered other engineered design concepts that were not included in the VA
reference design. These design options included use of drip shields to aid in delaying and
deflecting water from contact with the waste package, use of backfill, use of ceramic coatings on
the waste packages, and use of waste package designs with the CRM on the outside or with use
of two CRM materials. After the VA report was issued, the DOE began detailed evaluation of
alternative designs with the objective of selecting design features that would be used in the Site
Recommendation (SR) and the License Application (LA) if the Yucca Mountain site is found to
be a suitable location for disposal. The design that will potentially be used in the SR and the LA
is discussed in Section 7.2.2.5.
7-127
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Figure 7-35. Drift Cross-Section for the VA Reference Design (DOE98)
I
I
7.2.2.3 Thermal Management Strategy |
Thermal management strategy is concerned with using the heat emitted by decay of the
radioactive isotopes in the waste to control the temperature and the temperature gradients in and
around the repository, thereby controlling or affecting access of water to the repository, contact
of water with the waste packages, and the timing and rate of corrosion or degradation of the
waste packages and other components of the engineered barrier system.
The thermal management strategy used for the VA was to impose a high heat load on the rocks
surrounding the drifts so that water contained in trie pore spaces would boil and be driven away
from the drifts for as long as possible before the wjaste package heat emissions are too low to
sustain this phenomenon. The heat load selected fjbr the VA reference design was 85 MTU/acre,
7-110
-------
which was estimated to sustain temperatures at levels which would vaporize the percolation
water for about 2,000 years (DOE98, Vol. 3, Figure 3-14).
High thermal loading of the geohydrologic regime surrounding the drifts has potential to produce
a variety of effects on and within the regime, including opening or closure of fractures,
mineralization, and changes in the composition of solid and dissolved species in the percolation
water. The occurrence of such phenomena, and the impacts on long-term performance of the
repository, are highly uncertain and will be difficult to model reliably for repository performance
evaluations. These effects could lessen or improve repository performance. The geohydrologic
regime would undergo a temperature transient in which the temperatures near the drifts would
peak at about 150 °C a few tens of years after emplacement, and would not return to pre-disposal
ambient conditions for about 100,000 years. However, the temperature will have decayed to
levels where liquid water can impinge on the waste packages in no more than 2,000 years.
The Electric Power Research Institute has provided comprehensive analyses and discussions of
these complex issues and has developed models to characterize water/package contacts for
alternative engineered designs and geohydrologic regime characteristics (EPR96). Their analyses
demonstrate the wide range of conditions that can exist in the repository, and they also
demonstrate the dependence of performance on interactions between the heat transfer regime, the
hydrologic regime and repository thermal loading. 'They developed a five-dimensional matrix of
scenarios and packages-wetted fractions which "...provides a method for capturing the
correlations among heat transfer, water flow, waste package performance, and radionuclide
migration in a performance assessment model." DOE and EPRI performance assessment
methods and results are discussed in Section 7.3.
7.2.2.4 Data Sources \
Characterization of the Yucca Mountain site has spanned more than 20 years to date. Both
surface-based and underground investigations have been and are being performed to characterize
the natural features of a repository at the site.
Surface-based studies have included mapping of geological structures; monitoring of seismic
activity; use of gravitational, magnetic, and other non-invasive methods to infer geologic
characteristics at depth; monitoring of current weather and climate conditions; collection of data
to characterize past climates; heating of a large block of rock to determine the effects of heat on
hydrologic and geochemical properties; and drilling of numerous boreholes to obtain data on
7-131'
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geologic and hydrologic conditions at depth. Several hundred deep and shallow boreholes have
been drilled at the proposed repository site and within the region.
Underground data have been obtained from tunnels excavated specifically to obtain in-situ data
at the proposed repository horizon. The Exploratory Studies Facility (ESF), which is a north-
south tunnel 8 m in diameter and 7.9 km in length and parallels what would be the eastern
boundary of the repository and terminates at the North and South portals (see'Figure 7-30). The
Cross-Drift is an east-west tunnel which was excavated at a depth approximately 17m above the
i
proposed depth of the waste emplacement drifts anlJ at about the mid-point of the north-south
axis of the proposed repository. The surfaces of both of these tunnels have been mapped to
obtain data on-the geologic units, faults, and fractufes at the repository horizon.
l
Alcoves and niches have been constructed at various locations along these tunnels to serve as
facilities for a variety of experiments. Phenomena [and physical properties being characterized
include water flow characteristics in the unsaturatejj zone; drift-scale seepage; effects of high
precipitation rates on flow; effects of heating on roi± characteristics; fracture mineralization;
characteristics of small-scale fractures; and the presence and characteristics of fluid inclusions.
In addition to these site characterization activities at the repository horizon, other data acquisition
activities are in process. These include:
• Experiments are being performed in the tunnel facilities and at the Sundance fault
zone and the Drillhole Wash fault zone to extend the data base of "bomb-pulse"
Cl-36. This isotope can serve as a tracer to characterize the existence and
characteristics of potential "fast paths" for water and radionuclide transport
through the unsaturated zone.
• Pilot scale tests of backfill and drip 'shield performance are being conducted.
• The Nye County drilling program isj providing data on the geologic and hydrologic
characteristics of the alluvial deposits in the vicinity of Lathrop Wells. These data
will be used to refine or revise the saturated zone flow and transport models.
i
• A multi-phase, multi-purpose test program concerning radionuclide transport in
the unsaturated zone is being condubted at Busted Butte. Phases I and II are
currently underway; Phase III of the program would be conducted as part of the
performance confirmation program, i.e., after licensing if the site is approved for
disposal. ;
7-132
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The site data acquisition programs are augmented with laboratory programs to obtain other types
of data. An extensive program to obtain corrosion data for candidate waste package materials is
underway, involving a variety of corrosion environments and conditions expected potentially to
exist in the repository. Laboratory investigations also use rock samples to characterize chemical,
mechanical, and hydrologic properties of the geologic structures. Laboratory measurements also
characterize radionuclide solubilities and sorption properties using water with chemical
compositions expected to be characteristic of the repository.
These data acquisition activities have two broad purposes: to assure an adequate data base for
licensing reviews if the site is approved for disposal, and to reduce reliance on the results of
i
formal expert elicitations as a basis for performance models and performance parameter values.
To establish values for parameters used in the Viability Assessment, the DOE made extensive
use of recommendations produced from formal expert elicitations conducted in accordance with
guidelines established by the NRC. Process models subjected to expert elicitation included
unsaturated zone flow, near-field environment, waste package degradation, waste form alteration
and radionuclide mobilization, saturated zone flow and transport, probabilistic volcanic hazard
assessment, and probabilistic seismic hazard assessment (DOE98, Vol. 3, Table 2-1). Reviewers
of the VA, including the NRC, noted that the data base would have to be improved for a License
Application, so that there would be less reliance on expert opinion. Present activities are
intended to produce a data base that will be a sufficient foundation for performance models and
parameter values to be included in the License Application.
7.2.2.5 Alternative Repository Design Concepts Under Consideration '.
The DOE considered the repository design concept used in the Viability Assessment to be a
snapshot in time of the design evolution process. Within the VA documentation, the DOE
i
identified, and provided preliminary characterizations of, alternative design features not included
in the VA reference design. These included drip shields, backfill, alternative waste package wall
materials, ceramic coatings on the waste packages, alternative thermal loadings, and alternative
waste package emplacement configurations. The intent of these additional changes is to improve
the performance of the engineered barrier system or reduce uncertainties in assessing its
performance. Since issuance of the VA report in December 1998, the DOE has identified and
characterized six alternative engineered repository designs incorporating these options (DOE99).
As outlined below, one of these Enhanced Design Alternatives (EDA) has been selected to be the
reference design concept for the Site Recommendation. If considered necessary, further
evolution of the design may occur for the License Application if the site is approved for disposal.
7-133
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The ED As considered had common and variable features. Common features include use of drip
shields; use of carbon steel ground support, use of a steel invert with granular ballast, instead of
the concrete used in the VA reference design; use of a drift diameter of 5.5 m; use of pre-closure
'forced ventilation; and emplacement of 70,000 MT^HM of radioactive wastes.
Design features that varied for the ED As considered were the thermal loading and temperature
objectives; use of backfill; selection of waste package wall materials; use of thermal blending to
even out waste package heat emissions; drift spacing; waste package spacing; and repository
location within the characterized area. Constraints (imposed on the options were to maintain the
temperature of cladding on commercial spent nucle!ar fuel at less than 350 °C; allow personnel
access for off-normal events; and allow repository closure 50 or more years after start of waste
emplacement. The thermal goals for the EDA options, which influence many design features,
were: .
• EDA I: Maintain drift wall temperature below boiling
I
• EDA II: Keep centers of pillars between drifts below boiling
• EDA III: Cool waste package surface to 80 °C before relative humidity reaches 90
percent |
i,
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• EDA IV and V: Keep drifts dry for thousands of years
The design parameters for the EDAs considered are shown in Table 7-7. Note that EDA III
includes two options for the waste package wall materials.
Analyses of these options produced the results shown in Table 7-8. Comparison of these results
produced a recommendation by the M&O contractor to the DOE, which was accepted, that EDA
II was used as the initial, reference design for the Site Recommendation. Principal features of the
EDA II design are compared with those of the VA reference design in Table 7-9.
In comparison with the VA reference design, the EDA II design is expected to reduce
uncertainties that could be of concern during licensing reviews. Uncertainties that are expected
to be less significant as licensing issues are those concerning coupled thermal, hydrologic,
mechanical, and chemical processes; alteration of the natural system as a result of the heat load
on the geologic units surrounding the drifts; processes and phenomena that affect radionuclide
transport; and potential for localized corrosion of w|aste package wall materials. The EDA II
design is also expected to provide improved defense-in-depth and overall performance. One of
7-134 '
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the principal features of the design is that the time-temperature history of the waste packages is
expected to avoid conditions in which the Alloy 22 outer wall would be vulnerable to crevice
corrosion. ;
Repository performance assessment models and parameter values (see Section 7.3) were revised
from those used in the VA in accord with the EDA II design parameters and the information
emerging from the data acquisition program described in Section 7.2.2.4. The resulting
performance assessment, known as the TSPA for Site Recommendation (TSPA-SR) was issued
-
in late 2000. Principle differences between the TSPA-SR and the earlier VA assessment include
i
improved modeling of waste package performance for EDA II design conditions, which reduced
L
emphasis in juvenile waste package failures, and increased emphasis on disruptive events and
processes. The primary disruptive issues addressed in the TSPA-SR are igneous activity at the
site, and inadvertent human intrusion. ;
i
[
7.3 REPOSITORY SYSTEM PERFORMANCE ASSESSMENTS
The post-closure safety performance of a geologid repository for radioactive wastes is evaluated
using a Total System Performance Assessment (TSPA). A TSPA involves use of models of the
physical characteristics of the repository system, in a suite of linked computer codes, to forecast
the longterm performance of the system in terms of factors, such as-waste package degradation,
which lead to release of radiohuclides from the repository and their transport in the environment.
The TSPA tskes into consideration the features, processes, and events that can affect radionuclide
release and transport. ;
Features that affect performance include factors such as the corrosion rate of the waste package.
Processes that affect performance include factors such as the rate at which water seeps into the
drifts, and events important to performance include factors such as earthquakes, volcanic
eruptions, and intrusion of the repository by human action. A TSPA takes all of these factors
into account, consistent with the engineered and natural features of the repository system.
'
Evaluations of total system performance for potential repositories at Yucca Mountain have been
performed by DOE, EPRI, and the NRC. As discussed below, the DOE has performed a series of
TSPA evaluations, for purposes of helping to guide design evolution and site characterization
work. EPRI has also performed a series of independent evaluations, using models and methods
significantly different from those of the DOE. The NRC has performed evaluations to
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demonstrate their capability to perform licensing reviews of TSPA results that would be provided
by the DOE in a License Application.
DOE's historic TSPA efforts are discussed in Section 7.3.1 and Section 7.3.2. NRC's
performance assessments are discussed in Section 7.4, and EPRI's efforts are described in
Section 7.5. Results of recent assessments by DOE, NRC, and EPRI are compared in Section
7.6. The most recent DOE TSPA effort, TSPA-SR, is described in Section 7.3,10.
7.3.1 DOE's Historic Performance Assessments
DOE's TSPA process began with the PACE-90 project (DOE91). PACE-90 was not a total-
system evaluation; it focused on numerical modeling of the hydro logic regime and simulated
ground water flow and aqueous transport of radionuclides. Because data were sparse at the time,
models were simplistic and many performance factors were not considered. The PACE-90
analyses served to demonstrate the TSPA concept, and it laid the foundation for future TSPA
evaluations.
The DOE subsequently has conducted TSPA evaluations in 1991 (DOE92), 1993 (DOE94a,
DOE94b), 1995 (DOE95b), 1998 (DOE98) and, most recently, for the Site Recommendation
(TSPA-SR, TRWOOb). Each assessment built on the insights and results of prior assessments,
and on the evolving data base and design concepts. Each successive TSPA evaluation added
details and features to the models and parameter values in accord with progress enabled by the
evolving information base.
During the period of evolution of TSPA analyses to date, the regulatory basis for standards,
against which repository performance is to be evaluated, was revised. As discussed in Section
1.2 of this BID, the Energy Policy Act of 1992 directed the EPA to develop site-specific radiation
protection standards for Yucca Mountain, consistent with the findings and recommendations of
the National Academy of Sciences. Accordingly, the Agency has developed the 40 CFR Part 197
regulations supported by this BID. These standards establish dose limits as a basis for radiation
protection. The prior standards, contained in 40 CFR Part 191, also included individual
protection requirements (Section 191,15; see Section 1.4.4 of this BID) but established
cumulative release of radionuclides across an accessible environment boundary as the basis for
regulatory compliance.
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Because of the difference in the type of radiation' protection standards, the results of the TSPA-
VA analyses are expressed differently from those of prior analyses. Consistent with a dose-limit
standard, the TSPA-VA and TSPA-SR results are expressed as potential doses to receptors, for
time periods up to one million years. In contrast^ results for the TSPA 1991, 1993, and 1995
analyses were expressed in terms of a Complementary Cumulative Distribution Function
(CCDF), which is an appropriate representation of results for comparison with the cumulative
release standards established in the 40 CFR Part J191 regulations.
Key features of DOE's TSPA evaluations in 1991, 1993, and 1995 are summarized below.
I
TSPA-91 \
The TSPA-91 analyses were designed to develop the framework for probabilistic total-system
performance characterizations. They built upon the PACE-90 analyses by modeling nominal
conditions and disturbances from basaltic volcanism, human intrusion, and climate change. They
included the first set of stochastic analyses, in which hydrologic parameters were represented by
probability distribution functions based on site and analog data. Gaseous flow of C-14 was
modeled, the saturated zone was modeled for the; first time, and results were, for the first time,
obtained at the accessible environment boundary! as defined by EPA's 40 CFR Part 191
regulations. Future changes in climate were represented by a range of percolation flux values at
the repository horizon. j
i
•
TSPA-93 |
The TSPA-93 analyses were aimed at providing guidance for site characterization work and
engineered designs. In comparison with TSPA-9,1, the models of physical features and processes
were more sophisticated and the data base for selection of models and parameter values was
larger. Important features of the analyses included:
• A three-dimensional stratigraphy for the unsaturated zone which was based on site
data •
• A saturated zone model in which each geohydrologic unit was discretely modeled
Assessment of the effect of thermal loading (at levels of 57 and 114 kW/acre) on
performance \
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• Waste package failure models which included aqueous and dry oxidation
corrosion, and waste form degradation models which included dissolution and
oxidation
Consideration of two types of waste packages: the thin-walled, small-capacity
containers emplaced in boreholes, as envisioned in the Site Characterization Plan
(DOE88), and, for the first time, the large-capacity packages emplaced
horizontally in drifts
In anticipation of changes in regulations as a result of requirements of the Energy Policy Act of
1992, the TSPA-93 analyses included assessments of potential doses to humans as well as results
based on cumulative radionuclide releases from the repository, consistent with the 40 CFR Part
191 disposal standards. These results were illustrative, and were not intended in any way to
represent the actual potential performance of a repository at the Yucca Mountain site. At that
time the observation was made that more-representative models and data were needed to improve
the realism of the analyses.
TSPA-95 . '
As a result of studies of design options and guidance for site characterization work provided by
the results of the TSPA-93 analyses, the data basis for the TSAP-95 evaluations was significantly
improved over that which had previously been available. TSPA-95 sought to be as realistic as
possible on the basis of available information and the evolved repository and waste package
designs.
The focus of the TSPA-95 analyses was those components of the system that had been
determined by prior analyses to be most important to the waste isolation capability of the
repository. Emphasis was therefore placed on the engineered components and; the near-field
environment in which they would reside. In comparison with TSPA-93, the TSPA-95
evaluations used improved and more realistic models of the drift-scale thermal-hydro logic
environment and also of waste package degradation. Models describing the transport of water in
the near-field engineered barrier system were included, and flow in the unsaturated zone was
modeled. Disruptive events and gaseous release were not considered because they had been
shown in TSPA-93 not to be significant to overall performance.
Some of the models and parameter values used in TSPA-95 were based on judgments derived
from expert elicitations, because experimental data were limited or non-existent. Data
acquisition programs, such as corrosion testing and site characterization, are continuing and are
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expected in the future to enable replacement of expert elicitation judgments with experimental
data. i
The TSPA-95 analyses evaluated waste package lifetime, the peak BBS release rate, the
cumulative release at the boundary of the accessible environment, assumed to be 5 km from the
repository, and the peak dose rate, at 10,000 and one million years, to the maximally exposed
individual located at the boundary of the accessible environment. Evaluations were done using
alternative models and a range of alternative values for performance parameters, such as the
repository thermal loading, infiltration rate, and climate change. The DOE noted that, at the time
TSPA-95 was conducted, there were no documented models with substantiation adequate for use
with confidence in performance assessments. NeVer-the-less, TSPA-95 laid the foundation for
future TSPA evaluations using improved models and an expanded data base.
According to the DOE, the principal findings derived from the TSPA-95 analyses can be
summarized as follows: • '.
• Percolation flux at the repository horizon (and attendant seepage into the drifts) is
a dominant factor in repository system performance. This flux affects the
potential for water to drip into the drifts, the magnitude of radionuclide release
from a penetrated waste package, and the movement of radionuclides through the
unsaturated zone.
• Radionuclides that dominate dose potential for the 10,000-year time frame are Tc-
99 and 1-129. Long-term doses are> dominated by Np-237.
• Assumptions about dispersion and dilution in the UZ and SZ will have a strong
effect on peak dose rates.
• Excluding juvenile waste package failures from manufacturing defects, if waste
packages using the TSPA-95 design are not penetrated as a result of highly
aggressive corrosion conditions such as crevice corrosion, the EBS can by itself
provide complete containment of radionuclides for 10,000 years. Similarly, if the
percolation flux is low the natural-barriers system will provide complete isolation
for 10,000 years.
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7.3.2 DOE's TSPA for the Viability Assessment (TSPA-VA)
The TSPA-VA was part of the comprehensive assessment of the viability of the Yucca Mountain
project that was mandated by Congress in the Energy and Water Appropriations Act of 1997. In
comparison with prior TSPA efforts, the TSPA-VA was much more comprehensive and detailed.
Some previously used models were revised; models of repository features that affect performance
and had not been included in previous TSPA efforts were added to the com'puter code
configuration; waste package design features were revised; and data that had been developed
since TSPA-95 was prepared were used to provide details such as the spatial distribution of
infiltration rates.
The discussion in this section of the BID is specific for the VA repository design, the TSPA-VA
models and assumptions, and the data base used in the TSPA-VA. As noted by DOE in the VA
report, the VA data base, reference design, and TSPA results constitute a step in an evolutionary
process. Further design revisions and data additions have been conducted since the TSPA-VA,
leading to design features and TSPA methods and results for the Site Recommendation (TSPA-
SR), and eventually for a License Application if the site is found to be a suitable location for
disposal.
Comprehensive discussion of the TSPA-VA is included in this-BID because it is the most
recently available detailed information concerning DOE performance assessments for Yucca
Mountain. Although revisions to TSPA-VA methods and results are expected, only limited
information on future repository designs and TSPA methods is currently available.
Documentation of the first draft of the TSPA for the Site Recommendation is currently planned
to be available in July 2000; documentation of a revised TSPA-SR is currently planned for
February 2001.
7.3.2.1 Repository Design Features for the TSPA-VA
Repository design concepts have evolved significantly over the years of site evaluation. As
previously noted, for example, the design concept used in the Site Characterization Plan issued in
1988 was vertical emplacement of canisters with small capacities into the floors of the tunnels
and with expected lifetimes on the order of 300-1,000 years. The basic concept; used for the
TSPA-VA was to emplace large, highly robust waste packages with design lifetimes on the order
of tens of thousands of years horizontally in excavated drifts. This concept is similar to that used
in TSPA-95, but the waste package wall materials were different. '
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This section summarizes the engineered features 6f the VA repository that are of importance to
safety performance and TSPA results, hi general, these are design features that are specifically
selected to aid waste isolation by delaying and diminishing opportunities for water to enter the
drifts, to contact the waste form, leach out radionuclides, and transport the radioactivity to the
environment.
In the reference Engineered Barrier System (BBS) design that served as the basis for the TSPA-
VA analyses, the principal design features that contributed to waste isolation were use of high
waste package emplacement density so that repository temperatures would be high enough to boil
water in the rocks and drive it away from the repository for as long as possible; use of a drift liner
to help keep out seepage water for as long as the liner lasts; and use of a highly corrosion-
resistant waste-package wall material which would be expected not to be penetrated by corrosion
for very long periods of time. The TSPA-VA also characterized the potential performance of
supplemental engineered features (use of backfill! drip shields over the waste packages,-and
ceramic coatings on the packages), but these features were not included in the VA reference
design. ;
Assumptions That Provide the Basis for Design Parameter Values
i
I
Within the framework of the waste isolation strattegy outlined above, assumptions were necessary
as a basis for selecting design parameters. Key assumptions included the following:
The Nuclear Waste Policy Act of 1982 limits the repository to a total capacity of
70,000 metric tonnes of uranium (MTU) as spent fuel or equivalent. The
repository for the TSPA was assumed to contain 63,000 MTU of commercial
spent fuel and 7,000 MTU equivalent of defense wastes, including vitrified high-
level waste from defense productipn operations and spent fuel from naval
reactors. \
Spent nuclear fuel assemblies frorji pressurized-water reactors will be, on average,
25.9 years out-of-reactor, with a 3169 weight percent initial enrichment and a
bumup value of 39.56 gigawatt-days per MTU. Spent fuel assemblies from
boiling water reactors will be, on average, 27.2 years out-of-reactor, with 3.00
weight percent initial enrichment and a bumup value of 32.24 gigawatt-days per
MTU. i
• Commercial spent nuclear fuel (CSNF) will be emplaced in the repository in
packages containing 21, 12, or 24 PWR assemblies per package and 44 BWR.
assemblies per package each containing about 10 MTHM. There will be a total of
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7,642 CSNF packages in the repository. There will be a total of 2,858 packages of
defense wastes, for a repository total of 10,500 waste packages. ;
• The surface facilities, subsurface facilities, and waste package designs will be
based on a reference areal mass loading range of 80 to 100 MTU/acre.
The temperature of the drift walls will be limited to no more than 200°C (392°F).
The temperature of the CSNF fuel cladding will be limited to 350°C (662°F).
• The repository's western and eastern boundaries will be between the Solitario
Canyon fault and the Ghost Dance fault.
The reference repository and waste package designs that emerged from these and other
assumptions important to safety for handling and emplacement operations are summarized
below. '
Repository Footprint
The repository layout that resulted from the assumptions concerning standoff from the faults,
temperature limits, and the areal emplacement density is shown in Figure 7-30. The repository
east-west width is about 1 km and the north-south length is about 3 km. The repository would be
located at a depth about 300 m (1,000 feet) below the crest of the mountain and 300 m above the
water table. The main emplacement drifts would be 5.5 meters (18 feet) in diameter; 104 drifts,
totaling 107 km (67 miles) of length, would be excavated to emplace the 70,000 MTU of wastes.
The drifts would be spaced 28 meters (90 feet) apart, and the extraction ratio (fraction of the
volume excavated) for the emplacement region of the repository would be 19.6 percent.
Waste Package Emplacement Configuration
Given the assumptions about waste-package capacity, each package would be about 6 feet
(2 meters) in diameter and about 6 meters (18 feet) long to accommodate the dimensions of the
intact CSNF assemblies. Details of the package dimensions will vary because of variations in
assembly dimensions.
A cross-section diagram of a typical waste package emplaced in a drift is shown in Figure 7-35.
The package will be emplaced horizontally on steel V-shaped supports, which in turn are set on a
concrete invert and pier. The drift is lined with concrete. The invert completes a concrete ring
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around the perimeter of the drift and also provides a roadbed for construction and emplacement
operations. «
Waste Package Design
A perspective diagram of the waste package design for disposal of 21 PWR spent fuel assemblies
is shown in Figure 7-34. Packages for disposal of B WR spent fuel assemblies and for disposal of
defense wastes are conceptually similar in design. jAs previously indicated, the packages for
disposal of PWR and BWR spent fuel would be about 6 feet in diameter and 18 feet long.
Packages for disposal of defense wastes would be about 6 feet in diameter and 10 feet long.
The design features of most importance to the TSPA-VA are the materials selected for the waste
package walls, identified in Figure 7-34 as the inner and outer barriers. Each package has an
inner barrier of Alloy 22, which is a high-nickel, corrosion-resistant alloy intended in the design
I
to provide the principal barrier to penetration of water into the interior of the package. The outer
barrier, which in the reference design is a 516 steell is intended primarily to provide shielding
and package strength. The reference design thickness of the outer barrier is 100 mm (4 inches);
the inner barrier is 20 mm (0.7 inches) thick. j
'
Design Options
Many other possible design concepts and parameter values are identified and discussed in some
detail in the VA documentation (see, for example, Volume 2, Section 8 of DOE98). The options
include alternative design features, such as use of drip shields or ceramic coatings to defer the
time at which water can contact the waste package wall and begin to penetrate it, and alternative
design strategies. Although not part of the VA reference design, the effects of backfill, drip
shields, and ceramic coatings on repository performance were evaluated in the TSPA-VA.
Alternative strategies include use of a low emplaceiment density or long-term cooling before
emplacement, either of which would reduce the areal thermal loading and would be intended to
reduce performance issues and uncertainties arising from the high temperatures associated with
I
the VA reference design. DOE proceeded to characterize and evaluate some of the options, one
of which was chosen as the basis for the design forlthe Site Recommendation.
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7.3.2.2 TSPA Concepts and Methodology
This section presents an overview of TSPA concepts and methodologies that were the basis for
DOE's implementation of performance assessment in the TSPA-VA. As previously noted, the
TSPA-VA is a snapshot in time of performance evaluation for the VA reference design, data
base, and models that were available for the purpose. The TSPA has recently gone through
another iteration to become the TSPA for Site Recommendation (TSPA-SR). If the Yucca
Mountain project proceeds to the stage of preparing a License Application for a repository at
Yucca Mountain,'the details of the TSPA for the application would likely be different from those
of either the TSPA-VA or the TSPA-SR. Consequently, this section is intended to provide
general information on the basic concepts and methodology of TSPA, using TSPA-VA as an
example.
The basic TSPA principles used for the TSPA-VA have been adopted in radioactive waste
disposal programs throughout the world as the means for forecasting the post-disposal
performance of a repository. For any given repository natural setting and engineered design, the
process involves five basic steps:
• Develop and screen scenarios of conditions and factors important to performance.
Scenarios address features, processes, and events that can affect repository
performance, such as average annual precipitation rates and changes therein.
• Develop analytical models to represent the factors important to performance. The
models are usually implemented as computer codes.
• Assign values to performance parameters in the models. Some parameters will be
single-valued, such as the density of water at a given temperature; others will have
uncertainty ranges because of inherent variability or lack of certain knowledge of
the value.
• Implement the models by operating the computer codes. ,
• Interpret and apply the results for purposes such as identification of additional
data needs or assessment of compliance with regulatory standards^
For a proposed repository at Yucca Mountain with its particular geohydrologic setting, DOE
selected four basic performance strategy factors:
• Limit the potential for water to contact the waste packages ;
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• Design the waste package for a long lifetime
j
• Seek a low rate of release from breached waste packages
!
• Seek radionuclide concentration reduction during transport through the
environment to the location of the dose receptor
This strategy was implemented by identifying principal performance factors and components of
the TSPA modeling configuration as shown in Table 7-10. As indicated in this table, the model
components are aligned with the Key Technical Issues that NRC has identified as the basis for
review of DOE's assessments of repository performance. Parameter values and subsystem
models were developed for each of the 19 principal performance factors listed in Table 7-10.
Each of the performance factors listed in Table 7-16 can be characterized as a driver or an
inhibitor of radionuclide release and transport. For [example:
I
• Precipitation, infiltration, seepage, and dripping are drivers for radionuclide
release that bring water to the waste packages
• Waste package humidity, temperatuije, and chemistry drive the rate of attack on
the inner and outer waste package barriers
I
• The waste package wall is a principal inhibitor of radionuclide release; inhibition
of release is also accomplished by the integrity of the spent fuel cladding,
resistance to dissolution of the waste forms, and the limited solubility in water of
Np-237 |
• Radionuclide mobility during transit! from the repository to and through the
environment is aided if the radionuclides are attached to colloids but inhibited if
they become sorbed onto surfaces albng the flow path
i
• Transport of radionuclide-bearing water from breached packages brings the
radionuclides to the dose receptor location through pathways in the unsaturated
and saturated zones
I
• Dilution during transit and pumping jwill reduce the radionuclide concentrations in
water used by the dose receptor
• Biosphere transport will bring radionuclides into contact with the dose receptor in
accord with his/her life style and practices
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The specific characteristics of each of these drivers lor inhibitors of radionuclide release and
transport are represented in the parameters and models used in the TSPA.
7-148
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As noted in Section 7.2, one of the features of the repository design used in the TSPA-VA was an
initial high thermal loading, i.e., 85 MTU/acre, with a drift wall temperature of 200 degrees C.
The performance objective for this design concept is to drive the water in the geologic formations
around the repository away from the drifts for as loiig as possible, while radionuclides in the
wastes decay and heat emissions from the waste packages decrease. An adverse consequence of
the concept is that it produces high temperature levels and temperature gradients, which will
accelerate degradation processes and can change the characteristics of the geologic formations.
The thermal, chemical, hydrologic, and mechanical factors associated with the high temperatures
are coupled in highly complex ways that are difficult to model and characterize with reliable
parameter values. The modeling approach used in {the TSPA-VA uncoupled these factors,
thereby adding to the uncertainty of the TSPA-VA Results.
i
The computer codes and their configuration used in the TSPA-VA are shown in Figure 7-36. As
indicated in this diagram, thermal hydrology factors and UZ flow were modeled at both mountain
(large) and drift (small) scales. The Repository Integration Program (RIP) code receives input
from the codes for the individual performance factors and processes the inputs to calculate
radiation doses to the dose receptor(s). Many of the codes shown in Figure 7-36 were developed
or adapted specifically for use in the TSPA-VA; details are provided in the VA documentation
(DOE98) and supporting documents (DOE98a).
The codes used in the TSPA-VA include considerations of uncertainty and produce
. characterizations of uncertainty in the assessment results. Four types of uncertainty are
considered: parameter value uncertainty, conceptual model uncertainty, numerical model
uncertainty, and uncertainty in the occurrence of future events such as earthquakes or human
intrusion into the repository. For the TSPA-VA, there was considerable uncertainty in most of
the component models and in parameters that represent performance factors that are inherently
variable or had a sparse data base. Techniques such as Monte Carlo sampling are used to
characterize uncertainty in the results of the assessments; uncertainties in the peak dose rate
results of the TSPA-VA evaluations spanned four jo five orders of magnitude.
Nine radionuclides were considered in the TSPA-VA evaluations: C-14,1-129, Np-237, Pr-23l,
Pu-239, Pu-242, Se-79, Tc-99, and U-234. These are the nuclides that prior TSPA work has
shown to have the most potential to produce dose effects in the future because of their long half-
lives, their high dose consequences (e.g., Np and Pu), or their high mobility in the environment
(e.g., Tc-99, and 1-129). As discussed below, the highly mobile Tc-99 and 1-129 were found to
be the source for doses in the 10,000 year time period; Np-237 dominated doses in the period
7-15,0
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hydrology
INFIL
surface
infiltration
waste-package
degradation
ITOUGH2
unsatu rated
zone flow
calibration
TOUGH2
drift-scale
unsatu rated
zone flow
waste-form
degradation,
EBS transport
mountain-scale
unsaturated
zone flow
unsaturated
zone transport
SZ CONVOLUTE
FEHM
saturated
zone flow,
transport
saturated zone
transport
GENII-S
biosphere
dose
calculation
Run within RIP - ''
Biosphere
OUTPUT Parameters
r
KH
s,
X.
"a
°i
it
M,
c,
fs
Temperature
Relative humidity
Liquid saturation
Air mass fraction
Gas dux
Liquid flux
Infiltration flux
RadionuciWe mass flux
Radlonudide concentration
Fraction of WPs with seeps
Qs Seep Row rate
pH pH
LCOy7 Carbonate concentration
/ Ionic strength
tpj, Initiat-pit-penetration time
'pateft InHiat-patch-penetratlon time
Apfrf Perforated container area
Atufl Exposed fuel area
*SZJ Saturated zone transport time
BDCFf Biosphere dose conversion factor
EBS Engineered Barrier System
Legend
|l l| External Process Model '
| [ RIP Cells
'• : Rip Calls External Coda
_-_ Response Surface Between
Process Models
fr- Response Surface from
Process Model to RIP ;
_^~ Between RIP Cells and
External Coda >
Figure 7-36. Computer Code Configuration for the TSPA-VA (DOE98)
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tens-of-thousands to about 300,000 years; and Np-237 and Pu-242 were dominant in the period
from 300,000 to one million years. \
\
7.3.2.3 Key Features of the TSPA-VA Base Case Models
This section summarizes key features of the performance factors and computer codes that were
used to implement the TSPA-VA. The descriptiojns are based on information contained in
DOE98, Volume 3, Section 4. Highly detailed discussions of the performance factors were
provided in the chapters of the Technical Basis Document for the VA (DOE98a), and in topical
reports that were discussed as references, in the Technical Basis Document chapters.
Climate [
The TSPA-VA assumed there would be three characteristic climate regimes in the future at
i
Yucca Mountain, with periodic recurrence intervals: dry (current conditions), long-term average,
and superpluvial. Present conditions were assumed to prevail for the next 5,000 years. Long-
term average conditions were assumed to persist |for 90,000 years each time they occur, and
superpluvial periods were assumed to last for 10,000 years.
Average precipitation rates in the long-term average and superpluvial-periods were assumed to be
two and three times, respectively, higher than present rates, which average about 170 mm/yr.
Two superpluvial periods, in which glaciation is at a maximum and temperatures are a minimum,
were assumed to occur in the next million years: jone at about 300,000 years and the other at
700,000 years. Between the superpluvials, the 5,000-year dry periods and the 90,000-year long-
term average periods alternate. Under these assumptions, about 90 percent of the next million
years experiences the long-term average climate.
t
The water-table level was assumed to respond tojthe changes in precipitation, rising by 80 meters
from present levels during long-term average climates and 120 meters during the superpluvial
periods. One of the modeling consequences of the water-table rise is that the UZ flow path
length is shortened. j
Unsaturated Zone Flow and Infiltration
On the basis of site characterization data, the repository footprint was divided into six UZ flow
and infiltration zones. Three-dimensional steady-state flow models were developed for fracture
I
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and matrix flow under current climate conditions and were extrapolated to the wetter climate
conditions. Average infiltration rates for the present, long-term average and superpluvial climate
conditions were assumed to be 7.7,42, and 110 mm/yr, respectively. The infiltration rates were
therefore assumed to increase by factors of about 6 and 14 from the present rate, even though the
precipitation rate increases only by factors of 2 and 3.
Drift Scale Seepage
Characterization of seepage into the drifts was based on modeling of a three-dimensional,
heterogeneous fracture continuum surrounding the drifts. The seepage flow rate and fraction of
the packages that are affected by seeps were modeled in terms of percolation flux, i.e., the water
flux that arrives at the repository horizon after infiltration at the surface and flow through the UZ
above the repository. Percolation flux was characterized for each of the six regions of the
repository footprint and the three climate conditions, based on site data and the climate model.
The modeling showed that about 10 percent of the waste packages would be exposed to seeps
during the dry-climate period, 30 percent would be exposed to seeps during the long-term
average climate conditions, and 50 percent would be exposed during the superpluvial periods.
The estimates of the fraction of the packages exposed to seeps had a very high uncertainty range
in the TSPA-VA evaluations.
Thermal Hydrology
Thermal hydrology addresses the temporal and spatial impact of the spent fuel heat output on the
natural system geologic and hydrologic characteristics and on the performance of the engineered
features of the repository. Thermal hydrology models are used to calculate temperatures (waste
package surface, waste form, drift wall) and relative humidities in the drifts. Values for these
parameters provide information needed for other models such as the waste package degradation
model and the near-field geochemical environment models. Standard models of heat transfer,
and data concerning the physical properties of repository system materials, are used to
characterize the thermal parameters.
Near Field Geochemical Environment '
The near-field geochemical environment models calculate the time-dependent evolution of the
gas and water compositions that interact with the waste package, the waste form, and other
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materials in the drift. The evolution of changes inigas and water composition is modeled as a
sequence of steady-state conditions. The chemical;, thermal, hydrologic, and mechanical factors
important to the near field environment are in realty coupled, but an integrated model of the
coupling and its effects was not developed for the [TSPA-VA.
Five separate but interacting models were used in the TSPA-VA to characterize the near field
geochemical environment: j
Gas, water, and colloid compositions as they enter the drift
Composition of the in-drift gas pha'se
Chemistry of in-drift interactions of water with the solids and gases in the drift
• In-drift colloid compositions |
• In-drift microbial communities :
i
The near-field geochemical environment models ate connected to other component models (see
Figure 7-37). The near-field models receive input! from the UZ and thermal hydrology models
and from design parameters; they provide outputs jto the waste package corrosion model, the
waste form model, the UZ radionuclide transport model, and the nuclear criticality model.
Waste Package Degradation I
Modeling of waste package degradation was based on waste type contained in the package,
whether the packages were dripped on or not dripped on, and their location in the repository.
Seepage into the drifts is modeled as a function of the infiltration rate of water and the fracture
properties of the rock. With the expected percolation flux, only about one-third of the waste
packages are dripped for most of the one million year modeling period. If water seeps onto the
surface of a waste package, 100 percent of the surface is assumed to be wetted. Uncertainty in
the corrosion rate of the Alloy 22 corrosion-resistant barrier for the waste package wall was also
modeled. Corrosion of waste package materials was assumed to occur via pits and patches that
always encounter seeping water. Uncertainty in waste package manufacturing defects was also
addressed. The model used in the TSPA-VA assumed for the basic case that a single juvenile
waste package failure occurs 1,000 years after disposal.
i
Cladding Degradation \
Mechanisms included in models for degradation of fuel rod cladding on commercial spent
nuclear fuel included some pre-disposal failures, preep failure of zircaloy at high temperatures,
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total failure of rods clad with stainless steel, fuel rod fracture from falling rocks, and long-term
general corrosion failure. Breaching of cladding was assumed to expose all of the waste-form
surface in the rod to water that had entered the waste package.
Waste Form Degradation and Mobilization :
Dissolution of CSNF was modeled to be a function of pH, temperature, and total dissolved
carbonate; model parameters were based on experimental data. Dissolution of vitrified high-
level defense waste was modeled as a function of surface temperature and water pH, and a
dissolution rate constant for metals was used for degradation of the defense spent fuel from the
N-Reactor. Under the assumption that all spent fuel is exposed and wetted for rods with
breached cladding, the spent fuel would be totally dissolved in about 1,000 years. Dissolution of
uranium dioxide fuel is known to result in formation of secondary minerals which can trap
species such as Np-237 and reduce their release, but credit for this phenomenon was not taken in
the TSPA-VA modeling.
Engineered Barrier System Transport
Transport in the EBS was modeled as a series of connected mixing cells, with one cell combining
the waste form and waste package, and three pathway cells representing the invert, in order to
reduce numerical dispersion in model calculations. The models did not include factors that could
defer arid decrease radionuclide release after a waste-package wall is breached, such as low
seepage rates and partial seepage into the package interior, and in-package dilution. Sorption and
diffusional transport was assumed for radionuclide movement through the concrete invert.
Consistent with data which indicated rapid transport of plutonium from the Benham weapon test
location on the Nevada Test Site, a small fraction of the plutonium mobilized was assumed to be
attached to mobile colloids.
Unsaturated Zone Transport
The radionuclide transport model for the unsaturated zone was based on the flow model for that
zone. Three flow fields, corresponding to the three climate conditions, and a dual-permeability
geologic regime were assumed. Radionuclide movement was modeled using a three-dimensional
particle tracking model. Sorption was assumed to occur for Np-237, Pu-239, and Pu-242.
Matrix diffusion and dispersion were also assumed to occur. '
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Saturated Zone Flow and Transport \
i
Flow in the saturated zone was simulated using a coarsely discretized three-dimensional model
which establishes the general plume direction and flow path in the geologic media. Radionuclide
transport was assumed to occur in six one-dimensio|nal stream tubes corresponding to the six area
regions defined for the repository footprint. Based on the recommendations of the saturated zone
expert elicitation panel, the specific discharge in all stream tubes was assumed to be 0.6 m/yr,
and a dilution factor probability range, with a mean! value of 10, was assumed to apply to all of
the stream tubes. :
Biosphere Transport \
Water used by the dose receptor was assumed to be| drawn from a well 20 km (12 miles) down
gradient from the repository. Dilution was assumed not to occur during pumping, so the
radionuclide concentration in the water emerging ftiom the well is the same as the stream tube
concentration at the withdrawal location. The dose receptor was assumed to receive doses from
all biosphere pathways in accord with site-specific dose conversion factors and the water use and
life style habits assumed for the receptor. For the T^SPA-VA, DOE assumed the dose receptor is
a current-day average adult living in Amargosa Valley. A survey was conducted to obtain
lifestyle and dietary data for the dose evaluations. j
\
\
7.3.3 TSPA-VA Results
DOE produced the following categories of TSPA-VA results:
I
Deterministic results for the TSPA-VA base case
Results of uncertainty analyses using Monte Carlo techniques
Results of analyses to assess the sensitivity of performance to uncertainties in
parameter values
Assessments of the effect of disruptive events on performance
Assessment of the effect of design options on performance
Collectively, these assessment results address the expected performance of the repository, the
role of the various performance factors in producing the expected performance, factors that could
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alter expected performance, and the uncertainty in expected performance. The repository
performance forecasted for the base case is discussed in Section 7.3.3.1. Uncertainties in the
TSPA-VA result are discussed in Section 7.3.3.2.
7.3.3.1 Base Case Expected Repository Performance
The deterministic results for the TSPA-VA base case are responsive to the Congressional
mandate for assessment of "...the probable behavior of the repository in the Yucca Mountain
geological setting...". These results were a forecast of the dose rate to the average individual
located 20 km from the repository, for time periods up to one million years. Graphs showing
forecasts of peak doses throughout the million-year time period were produced, and specific
dose-rate values were identified and discussed for time periods of 10,000, 100,000 and one
million years.
DOE described the results for the deterministic evaluation in which values for all uncertain
parameters were set at their expected values as follows (DOE98, Volume 3, p. 4-21):
"1. Within the first 10,000 years, the only radionuclides to reach the biosphere
are the nonsorbing radionuclides with high inventories, technetium-99 arid
iodine-129, and the total peak dose rate is about 0.04 mrem/year. ;
2. Within the first 100,000 years, the weakly sorbing radionuclide neptunium-237
begins to dominate doses in the biosphere at about 50,000years, with ihe\total
dose rate reaching about 5 mrem/year.
3. Within the first million years, neptunium continues to be the major contributor
to peak dose rate, which reaches a maximum of about 300 mrem/year at about
300,000 years after closure of the repository, just following the first climatic
superpluvial period. The radionuclide plutonium- 242 is also important during
the one million-year time frame and has two peaks, at about 320,000 and 720,000
years, closely following the two superpluvial periods. There are regularly spaced
spikes in all the dose rate curves (more pronounced for nonsorbing radionuclides
such as Tc-99 and 1-129) corresponding to the assumed climate model for the
expected value base-case simulation...these spikes are a result of assumed, abrupt
changes in water table elevation and seepage through the packages. "
i
As shown in Figure 7-37, doses to the receptor 20 km from the repository, as a result of the
mobile Tc-99 and 1-129 radionuclides, first occur about 3,500 years after disposal. These fission
products are dominant because of substantial inventory in CSNF, high solubility in seepage
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water, relatively low decay rate relative to 10,000 years, and neglible sorption on tuff rocks. The
scenario presented in Figure 7-3 8 results from the |assumption that a single juvenile waste-
10,000-yr Dps© Rate
2,000
4,000 §,000
Time
Figure 7-37. TSPA-VA Base Case Dose Rates for Periods Up to 10,000 Years (DOE98).
3#Q$0ti-yir Dose ftate
ID'3
0 20,000. 40,000 i%000 80,000 100,000
Figure 7-38 TSPA-VA Base Case Dose Rates for Periods Up to 100,000 Years (DOE98)
I
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package failure occurs at 1,000 years; the "blip" in the curve at about 5,500 years is the result of
the change of climate conditions from dry to long-term average at 5,000 years, which causes a
major rise in the water table. During the 10,000-year period, 17 additional packages are modeled
to fail at various times, beginning at about 4,200 years. These failures contribute to the dose at
10,000 years in accord with the TSPA-VA model assumptions concerning package failure times
and conditions.
Dose rate histories for times up to 100,000 years are shown in Figure 7-38. Tc-99'continues to
dominate the dose rate up to about 50,000 years, after which the Np-237 dominates the dose rate
out to 100,000 years. There is a relatively large inventory of Np-237 in CSNF resulting from the
decay of Am-241. The Np-237 does not begin to appear at the dose location until .after about
30,000 years, because its release from the waste form is solubility limited and it exhibits some
sorption on the rock surfaces along the transport pathway. The Pu-239 does not begin to appear
at the dose location until more than 80,000 years have elapsed because it is more strongly sorbed
than the Np-237. A small fraction of the Pu-239 is assumed, however, to be attached to colloids
that are not sorbed onto the rock surfaces.
As with the 10,000-year results, the dose rate forecasts for periods to 100,000 years are
dominated by climate change assumptions and waste package failure history. The jagged
appearance of the Tc-99 curve is the result of individual package failures; each small peak
corresponds to a failure. This illustrates one of the key features of the TSPA-VA modeling
scheme: because features such as slow drip entry to the package interiors and in-p;ackage
dilution, which provide storage capacity along the transport path, were not included in the
models, the nonsorbing species such as Tc-99 directly track release behavior, and concentrations
are simply attenuated by dilution along the pathway. The sorbing and solubility-limited species,
such as Np-237 and Pu-239, have the capacity for storage along the transport path because of
these properties, but the effects would have been more exaggerated if factors such as in-package
dilution had been included in the TSPA models.
As shown in Figure 7-39, Np-237 continues to dominate the dose rate from 100,000 years all the
way to the end of the million-year dose evaluation period. At about 300,000 years, Pu-242
becomes the second most important contributor to dose and remains in this role, at a level about
a factor often less than that of the Np-237, to the end of the dose evaluation period. The
contribution of other radionuclides to dose during the long-range time frame is insignificant.
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The dose rate after about 300>000 years is seen iiii Figure 7-39 to be essentially constant. This is
because, in the TSPA-VA m'tfdeling scheme, the|repository as a source term for radionuclides
released to the environment-§bes into essentially'steady state. All of the packages that are
modeled to fail have failed; tile seepage fluxes into the repository and into the packages have
become virtually the sanie ahd constant, and the irate of change in exposure of waste form has
become constant.
t,uG0tOQO-yr Dose Kate
200,000 400,000 600,000 800,000 1,000,000
Time (years)
Figure 7-39. TSPA-VA Base Case Dose Rates tfor Periods Up to One Million Years (DOE98)
The dominant effect of wastei package failure history and climate conditions on dose rates
continues to the end of the million-year dose evaluation period. At about 200,000 years, cladding
degradation begins to contribute to the exposed waste form area, and at times greater than about
700,000 years, waste paekaj*e& that are never dripped on, which total about 55 percent of the
package inventory, begin" id fail as a result of low corrosion rates in a non-wetted condition over
a very long time frame.
i
The base case TSPA results1 for the VA repository show that the performance of the highly
complex and multi-element System is strongly doniinated by very few factors. In brief:
Performance^ dominated by assumptions concerning waste package failure
history ariid climate, and the effect
a consequent of the assumptions
climate ch'ahge'.
of these factors on predicted doses is primarily
concerning juvenile package failures and
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• Three nuclides dominate the forecast doses: Tc-99 and 1-129 in the shorter time
frames and Np-237 in the longer time frames. The dose levels associated with the
Np-237 are higher than those associated with the technetium and the iodine, in
large measure because the health consequences of a unit quantity of Np-237 are
much greater than those for the technetium and iodine.
• The fact that the dose results clearly reflect the occurrence of climate changes and
individual package failures shows that the TSPA-VA modeling system is
fundamentally simple. Factors in performance that would serve to smooth and
smear the consequences of phenomena that change system conditions were
omitted from the models. I
7.3.3.2 Uncertainty in the TSPA-VA Results
The Monte Carlo type of analyses that were done to assess the uncertainty in the TSPA-VA
deterministic base-case results showed an uncertainty range spanning about four to five orders of
magnitude throughout the million-year period, as shown in Figure 7-40. These results were
obtained by using statistical methods to select values from the distributions for the uncertain
parameters used in the TSPA-VA models. For each of the three time frames (i.e.:, 10,000,
100,000, and one million years) one hundred such runs were done, and a few 1,000-run studies
were done to demonstrate that the uncertainty ranges found for the 100-run studies were
representative.
The large uncertainty range, i.e., spanning four to five orders of magnitude, is in part due to the
many uncertain parameters in the TSPA-VA computer codes. The RIP code alone, for example,
contains 177 uncertain parameters, and there are many more in the codes that have inputs to RIP.
Another possible cause of the wide uncertainty range is that many of the uncertain parameters
themselves have wide uncertainty ranges, either as a result of use of a broad range of possible
values because the actual value of the parameter is poorly known, or because the parameter is
inherently highly variable. It would be difficult, if not impossible, to sort out the sources and
principal causes of the uncertainty range. The uncertainty range for the TSPA-VA results is
therefore a consequence of the specific way uncertainty was used in assigning numerical value
distributions to parameters in the TSPA-VA models and codes.
Another source of uncertainty, not reflected in the results of the TSPA-VA studies, is the
possibility that some of the models used in the codes may not be correct, e.g., because of a sparse
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10
-8
0
2000
4000 6000
Time (years)
8000
10000
o
200000 400000 ! 600000 800000 1000000
Figure 7-40. Uncertainties in the TSPA-VA Base Case Results (DOE98)
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data base, or, as in the case of modeling of the near-field geochemical environment, because
coupled phenomena were uncoupled to simplify modeling. This type of uncertainty should be
regarded as uncertainty in the conceptual models for the waste containment and isolation
systems. In translating conceptual models into calculational models, conservative assumptions
are typically made about processes which should be included and how the processes would
operate. This is done, in part, for modeling convenience, and, in part, because the level of
process complexity cannot be handled manageably. These assumptions can have significant
implications for interpreting the results of performance assessments, and should be understood
when interpreting the results. (See Section 7.3.3.5 for additional discussion of conservatism in
the TSPA-VA modeling.)
In evaluating the status of knowledge and uncertainty as a prelude to selecting further work to
improve the TSPA methodology for a License Application (DOE98, Volume 4), DOE often
noted that the models used in the TSPA-VA might not adequately capture the full range of
possibilities. If this is indeed the case, and the uncertainty in parameters or models has to be
expanded in order to embrace the full range of possibilities (as opposed to simply revising the
model in response to better information), the uncertainty ranges for future TSPA results might
actually be broader.
DOE used a technique known as Stepwise Regression Analysis to determine which of the
performance factors were most important to the uncertainty results. These evaluations showed,
for the 10,000-year time period, that the fraction of the packages contacted by seepage, the mean
Alloy 22 corrosion rate, the number of juvenile failures, and the saturated zone dilution factor are
the most important performance parameters. For the 100,000-year period, the most important
parameters were the seepage fraction, the mean Alloy 22 corrosion rate, and the Variability in the
Alloy 22 corrosion rate. For one million years, the most important factors were found to be the
seepage fraction, the saturated zone dilution factor, the mean Alloy 22 corrosion rate, and the
biosphere dose conversion factors. The fraction of waste packages contacted by seepage water
was the dominant performance factor for all three time periods. It is the dominant factor for
TSPA modeling of repository system performance because it has a direct effect oh the number of
waste packages that fail, and it has a very large uncertainty.
!
Additional sensitivity studies were done to determine the performance factors of secondary
importance to the TSPA-VA results. In these analyses, the performance factor of primary
importance were held constant, and Monte Carlo runs were done for the other uncertain
parameters. The performance factors that were held constant at their baseline values were the
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infiltration rate and mountain-scale saturated zone flow rates, the fraction of the waste packages
contacted by seepage, the seepage flow rate, the Alloy 22 mean corrosion rate, and the Alloy 22
corrosion rate variability. !
I
With the above parameters held constant, the parameters of principal secondary importance at
10,000 years were found to be the saturated zone dilution factor, the biosphere dose conversion
factors, the solubility of technetium, and the fraction of seepage contacting a package that enters
a failed package. The factors that were important for 10,000 years were found to be also
important for 100,000 years, except that the solubility of neptunium replaced the solubility of
technetium as an important factor, and the fractipn of saturated zone flow in alluvium was added
to the list. At one million years, the most important factors were the saturated zone dilution
factor, the cladding failures by corrosion and by mechanical disruption, the biosphere dose
conversion factors, the saturated zone longitudinal dispersivity, and the saturated zone alluvium
fraction. In all time frames, the most important pf these secondary factors was the saturated zone
dilution factor. ;
All of these sensitivity findings reflect the fundamentals of repository system performance: the
potential doses depend primarily on the fraction of waste packages intercepted by seepage, the
amount of waste form available to be a source of radionuclides, the amount of water available to
pick up the radionuclides and to transport them to the environment, the amount of water available
to dilute radionuclide concentrations, and the extent and means of interaction of the dose receptor
with the contaminated water. |
1
7.3.3.3 Effects of Disruptive Events on Performance
The TSPA-VA evaluated the effects of four type's of disruptive events on repository performance:
basaltic igneous activity, seismic activity, nuclear criticality, and inadvertent human intrusion.
The basis for inclusion of evaluations of the effepts of disruptive events on repository
performance includes the probability of occurrence of the event, the consequences of occurrence,
and any regulatory requirements that mandate or exclude consideration of disturbances.
The igneous activity evaluations considered events in which molten igneous material is cooled
within the earth or on the surface. In the case wljtere magma reaches the surface, explosive
releases may carry radioactive materials directly into the atmosphere. Cooling of magma within
the earth may involve destruction of waste packages so that radionuclides in the waste form are
more accessible for release and transport. !
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Results of the direct-release igneous activity evaluations showed that the maximum dose rate
from this volcanism would be about two million times less than for the base case ground water
contamination scenario. The underground-cooling scenarios showed that dose rate peaks would
occur tens of thousands of years after the actual magma intrusion event.
The seismic activity studies considered phenomena such as rockfall onto waste packages as a
result of earthquakes, and the effects of seismicity on the hydrologic regime in the near field and
in the saturated zone. These studies showed that rockfalls could not contribute significantly to
waste package degradation until after at least 100,000 years and that changes to the hydrologic
regimes would be negligible. Overall results of the analyses showed that seismic events would
have almost no effect on repository performance over one million years.
The potential for nuclear criticality within waste packages and external to the packages after
transport of fissionable material from the package was investigated within the TSPA-VA. The
evaluations were done assuming that criticality occurs 15,000 years after emplacement, which is
when the commercial spent fuel is most reactive. The analyses determined that criticalities
external to the waste packages are not a credible event, and that criticality within a package is
extremely unlikely and would have insignificant consequences. Criticality within a waste
package is extremely unlikely because only 8 percent of the commercial fuel waste packages
contain sufficient fissile material to acheive a critical mass and only 10 percent of the waste
packages are expected to be breached in 40,000 years. Breached waste packages must retain
sufficient water to act as a moderator for the nuclear chain reaction to be sustained and DOE has
estimated that only 25 percent of the breached waste packages will hold water for a period
sufficient to flush out boron which is included in the waste package as a neutron absorber. Even
if criticality did occur within the waste package, the incremental radioactivity is less than the
normal radioactivity from most waste.
In keeping with the recommendations of the NAS panel that developed the technical basis for the
Yucca Mountain standards, a stylized human intrusion scenario was characterized and evaluated.
Intrusion of a waste package by an 8-inch drill bit, as a result of search for water, was assumed.
The bit was assumed to penetrate the package and the mountain stratigraphy to the water table,
with large quantities of pulverized fuel being transported to the bottom of the bore hole, which
was never sealed. Water would then dissolve the fuel inventory at the bottom of the bore hole
and transport radioactive material to the dose receptor location. The intrusion was assumed to
occur at 10,000 years, which is the first time at which it is estimated the drill bit could penetrate
the package wall. The NAS panel did not feel it would be useful to assess hazards to drillers or
7-165 \
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I
to the public from radioactive materials transported directly to the surface since these risks would
be the same for all geologic repositories. |
|
The total amount of fuel deposited at the bottom of the bore hole was assumed to range between
550 and 2,700 kilograms (1,200 and 6,000 pounds), which corresponds to about 5 to 22 percent
of the total spent fuel inventory in the package The actual mass of fuel that would be intercepted
by the 8-inch drill would be about 160 kg, so the analyses assumed that large quantities of fuel
would be entrained by the bit as it passed through the package.
r
i
The analyses for this intrusion scenario showed that the consequent radionuclide releases for the
2,700 kg release would produce a blip in the dose rate curve, in comparison with the base case,
that starts at about 11,000 years, peaks at 12,000 years, at levels about 145 times higher than the
base case dose rate at that time (i.e., 1 mrem/yf), and returns to base case levels at about 14,000
years. The 550-kg spent fuel release from intrusion produces a dose rate at 12,000 years that is
3.7 times the base case dose rate. All effects of the intrusion on dose rate are gone by 150,000
years. The TSPA-VA observed that the effects of the intrusion on dose rates are significant only
for times near the occurrence of the intrusion, and that the maximum resulting dose is 1 mrem/yr.
7.3.3.4 Effects of Design Options on Performahce
I
The TSPA-VA included evaluation of the effect, on repository performance, of design features
that were not included in the VA reference design. The three features considered were emplaced
drift backfill, drip shields, and ceramic coating of the disposal containers, with backfill. The
objective for use of these design options would be to reduce and defer liquid water contact with
the waste package: !
7.3.3.4.1
Effects of Backfill
The backfill was assumed to be crushed tuff, emplaced 100 years after the end of emplacement
operations. The backfill will initially perform £is a thermal blanket for the waste packages, and
cause a temperature spike of as much as 80-90°C. The temperature spike might cause a slight
increase in the waste package corrosion rate, but it would also delay the rate of increase of
relative humidity as the heat emissions from the waste packages decrease and the repository
system cools. A potentially major effect of backfill would be to change the potential for, and
patterns of, seepage water contacting the waste packages. This effect was not modeled in the
TSPA-VA analyses. I
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The analyses for the assumed backfill effects showed that the use of backfill would defer
corrosion of the Corrosion Allowance Material, but corrosion of the Corrosion Resistant Material
would be virtually unaffected based on the modeling assumption that corrosion of this material is
driven by whether or not dripping occurs and the same dripping conditions are assumed for the
case of backfill and no backfill. Use of backfill would therefore have little effect on repository
performance if the backfill does not reduce or defer contact of seepage water with the waste
packages. The backfill might actually have effects such as diverting the seepage water around
the waste packages or reducing the amount of seepage that gets to the package as a result of
evaporation, but a basis for modeling such effects was not available for the TSPA-VA.
7.3.3.4.2
Effects of Drip Shields
The drip shields were assumed to be made of Alloy 22 and to be 2 cm (0.8 in.) thick. The shields
would be shaped like a Quonset hut, shrouding the waste packages but not touching them. The
dripshields would be covered with backfill, emplaced 100 years after emplacement of the waste
packages was completed. The shields upper surfaces were assumed to be totally wet in dripping
regions of the repository, and they were assumed to fail only by general corrosion. After drip
shield failure, 10 percent of the waste package area under the failed shield was assumed to be
wetted (in contrast, the base case analyses assumed 100 percent of the package surface area
would be wetted) because only a small fraction of the drip shield surface area was modeled to
fail. ;
TSPA-VA results based on the above assumptions showed that the drip shields enhanced the
overall waste package lifetime by more than 100,000 years. Dose rates for the first 300,000 years
are reduced by one to two orders of magnitude in comparison with base case results. After
500,000 years, the drip shield dose projections become the same as those for the base case. The
results were interpreted to indicate that the life span if the drip shield is the key determinant of
improved performance.
As a result of these findings, drip shields are included as a design feature for theirepository
design expected to be selected as the reference design for the Site Recommendation and the
License Application (see Section 7.2.2.5).
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7.3.3.4.3 Use of Ceramic Coating of Disposal Containers with Backfill
This design option involves coating the waste packages with a ceramic material in order to delay
corrosion of the outer wall of the packages (in the VA design, A 516 carbon steel). Backfill is
added to the repository to protect the ceramic coatings.
Performance of this design concept was modeled assuming that the ceramic coating functions as
a barrier to oxygen transport to the carbon steel package wall. For the assumed conditions, the
analyses determined that the ceramic coatings would not be breached for more than 300,000
years. Dose rates would not begin until about 500,000 years, and at one million years the dose
rates would be nearly two orders of magnitude less than those for the TSPA-VA base case.
If ceramic coatings perform as modeled for the JTSPA-VA, they would have a profound effect on
repository system performance. At this time, however, there are uncertainties and concerns
associated with potential for defects and flaws in the coatings, differential thermal expansion
between the coating and the substrate that could result in cracks in the coating, and dissolution of
the coating over long time periods. Analysis ofjthese effects is needed before the potential
I
benefits of use of ceramic coatings can be verified.
7.3.3.5 Conservatism In The TSPA-VA Base Case Results
j
The TSPA-VA base-case results (an expected (average) value dose rate of 0.04 mrem/yr 10,000
years after disposal, to a reference person 20 kni downstream) are a consequence of choices that
were made concerning performance parameter values, performance models, and assumptions.
This section discusses conservatism that was exercised in making the TSPA-VA choices, and the
effects of conservatism on the base case results.
Similar discussions are provided for the NRC
performance assessments (Section 7.3.5.3) and the EPRI assessments (Section 7.3.6.4).
Performance Parameters \
The TSPA-VA base-case evaluations used expected values of performance parameters, based on
available information. Expected values for some of the parameters, such as the dilution factor
for the saturated zone and corrosion rates of Alloy 22, were based primarily on results of expert
elicitations because of limited availability of data at the time that the TSPA-VA analyses were
performed. The parameter values developed by the expert elicitations may be conservative
because the experts are, themselves, working with limited information. Expected values of
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parameters, and the uncertainty ranges for parameters that are inherently variable, may change in
the future as a result of data additions, but the TSPA-VA analyses sought to be as realistic as
possible, rather than conservative, in their choices of performance parameter yalues.
Performance Models
Conservatism in the suite of performance models and computer codes used for the TSPA-VA
analyses was introduced by using simplified models and by omitting from the suite of models
some performance factors that could have significant impact on predicted doses. Examples of
this type of conservatism include:
Dilution and transport delay for radionuclides released from the waste form but in
water still within the failed package were not considered. Under realistic package
failure conditions during the first 10,000 years, when disruptive failure scenarios
are insignificant, water will fill the package interior very slowly from a
penetration in the top. By the time that radionuclide release and in-package
transport occurs, temperature gradients will be too low to drive advective
transport processes, and temperature levels will be too low for inside-to-outside
corrosion of the Alloy 22 to occur and create an exit at the bottom of the package.
Radionuclide transport rates within the package will therefore 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 provide dilution. Radionuclide releases to
the exit of the package may therefore be greatly delayed, and concentrations at the
package exit would be much lower than for the no-dilution assumption.
• Release of radionuclides from a breached waste package was assumed in the VA
models to begin immediately after the waste package was breached, i.e., an exit
hole in the metal container was assumed to be created as soon as the container
wall was breached by corrosion. In reality there would be a time delay before an
exit hole at another location on the container was developed. This time delay
could be relatively short if exterior corrosion was taking place concurrently at
opposite sides of the container, or it could be very long if, as indicated above, the
exit pathway had to develop from inside the container. By delaying the exit of
radionuclides the actual containment time of the waste containers would be
significantly increased and doses during the regulatory time frame would be
consequently decreased.
0 Dilution of radionuclide concentrations during transit of the unsaturated zone
from the repository to the water table was not considered. When few packages are
failed and releasing radionuclides (in the TSPA-VA, only 18 of 10,000 packages
are failed at 10,000 years), uncontaminated percolation water adjacent to
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contaminated streams emanating from the failed waste packages could provide
extensive dilution as a result of mixing of contaminated and uncontaminated
water in the fracture and matrix flow paths. This mixing would lower the
radionuclide concentrations at the start of saturated zone transport and result in
lower predicted doses to the receptor.
[
I
• A simple, one-dimensional model of radionuclide transport along the saturated
zone flow paths from the repository to the dose receptor location was used, and
dilution of initial SZ radionuclide concentrations under the repository was
assumed to occur at the end of the path, in accord with dilution factors
recommended by experts (for thet base case, a dilution factor of 10 was used).
Processes that could delay and disperse radionuclide transport along the pathway,
and therefore would reduce the predicted dose rates to the receptor, were not
included in the modeling. '
• Dilution during well pumping by the dose receptor was assumed not to occur.
This expected dilution process, which is included in NRC modeling of repository
performance, would reduce predicted doses to the receptor.
These processes and phenomena were omitted from TSPA-VA modeling of repository
performance because at the time the data base for characterizing the relevant performance
parameters and their uncertainties was limited or non-existent. Also, the magnitude of these
effects is difficult to quantify with high confidence even with site characterization and laboratory
work focused on them. However, these processes would be expected to function in the actual
repository environment, and reasonable but cautious estimates could be made to support
assessments, through a combination of data collection and expert judgment.
i
Rather than choosing to incorporate models for these processes in the TSPA-VA assessments,
with estimated values of the parameters used in the calculations, they were omitted from the suite
of TSPA-VA models. This approach had the consequence of producing a spectrum of
performance results that are an assessment of a potentially very conservative performance
scenario, incorporating some unrealistic modeling assumptions. Omission of these modeling
features introduces a significant level of conservatism in the assessment results whereas better
performance would reasonably be expected. j
Additional data (e.g., additional characterizatiori of the SZ geology and hydrology), may enable
inclusion of at least some of these performance factors in the TSPA for the LA. Their omission
introduces conservatism to the TSPA results, but also avoids licensing issues that may be
difficult to resolve unless a data base adequate t6 support their use is available.
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Conservative Assumptions
The TSPA-VA evaluations included conservative assumptions for some of the key performance
factors, as follows:
>
In the base case, early failure of a waste package was assumed to occur at 1,000
years as a result of an imperfection such as a poor weld. Performance parameters
selected in association with this assumption (e.g., the size of the hole on the
package wall) were such that nuclide releases from this single package were a
dominant factor in the predicted base case dose rate at 10,000 years.
• The Corrosion Resistant Material for the waste package wall, Alloy 22, was
assumed to be penetrated rapidly by crevice corrosion as a result of being under
carbon steel in the VA waste package design. This assumption was derived from
the waste package expert elicitation, which conservatively interpreted the highly
limited data base for the corrosion performance of Alloy 22.
Li characterizing corrosion processes, the TSPA-VA assumed that all ground
water seeping into the emplacement drifts contacts the waste packages, even
though the package width is only one-third the width of the drift, thereby
overstating the amount of water available to cause corrosion. In addition, the
entire surface of a waste package wetted by seepage water dripping onto the
package was assumed to be wetted, and all seepage water contacting the package
was assumed to enter the package wall penetration(s) when they occur. The
TSPA-VA support analyses (DOE98a) recognized that only a small fraction of the
waste package surface would be wetted (the total amount of water contacting the
package each year is estimated to be on the order of 20 liters), and that only a
fraction of the seepage water contacting the package would enter; the wall
penetration (e.g., because corrosion products would block entry). Because of
uncertainties in placing values on the relevant performance parameters, these
factors, which could greatly defer and diminish radionuclide release from the
waste form, were omitted from the TSPA-VA evaluations and the bounding
conservative assumptions were used.
The TSPA-VA assumed that 0.1 percent of the Zircaloy-clad commercial spent
fuel rods emplaced in the repository will be "failed" at the time of emplacement,
that the spent fuel contents of each penetrated waste package will include 1.15
percent stainless-steel-clad fuel rods, all of which fail completely and immediately
when the package wall is penetrated, and that all waste form area in failed fuel
rods is exposed and contacted by water that enters the package. Overall,
therefore, 1.25 percent of the waste form area in a failed package was assumed to
be exposed and wetted, hi the context of the TSPA-VA evaluations this was
considered by some (i.e., NRC staff and the NWTRB) to constitute cladding
"credit" because only a small fraction of the waste form in a failed package was
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assumed to be exposed and wetted. The TSPA-VA assumptions may in fact
greatly overstate the extent of exposed waste form area. An extensive data base
shows that "failures" of spent fuel Jcladding are predominantly hairline cracks
which would expose only a small waste form area. In addition, the Zircaloy
cladding is not susceptible to significant degradation after disposal, and there are
only about 2,100 stainless-steel-clad subassemblies, which could be packaged
together in less than 100 of the 10,000 waste packages. These segregated
packages could be made more failure resistant by using some of the design
options assessed in the VA, such as drip shields. With a greatly prolonged waste
package lifetime the level of assuined cladding "failure" at emplacement would be
lowered by an order of magnitude with consequent lowering of the dose to the
receptor, hi summary, if only the penetrations of Zircaloy cladding that exist at
emplacement allow water to contact the waste form, and if extreme assumptions
concerning stainless-steel-clad spent fuel are avoided, the DOE assumptions could
overstate the waste form area available for radionuclide release by as much as
three orders of magnitude.
7.3.4 Reviews of the TSPA-VA
Formal reviews of the DOE Viability Assessment and the TSPA-VA were documented by the
Nuclear Regulatory Commission, the TSPA-VA Peer Review Panel, and the Nuclear Waste
i
Technical Review Board. Their comments are summarized below.
7.3.4.1 NRC Review of the TSPA-VA
In a March 1999 letter to the NRC Commissioners, the NRC Staff provided comments on the
TSPA-VA (NRC99c). hi addition, the NRC provided some informal feedback to DOE during
the May 25-27,1999 DOE/NRC Technical Exchange (NRC99b). The NRC's feedback was
based primarily on a comparison of the TSPA-VA with NRC's TPA 3.2 performance assessment.
Details of TPA 3.2 are presented in Section 7.3.5.
substantive differences in the models and paramet
As discussed in that section, there are
;ers used by the two agencies. The purpose of
this section is not describe the differences between the TSPA-VA and TPA 3.2 but rather to
summarize some of the key NRC comments on the TSPA-VA.
The NRC Staff review covered: (1) the preliminary design concept for the critical elements of the
repository and the waste packages; (2) the TSPA based on this design concept and data available
as of June 1998; and (3) the license application (LiA) plan. The Staff did not review the DOE
cost estimates to construct and operate the Yucca Mountain repository. The review focused on
those issues that needed to be addressed before the LA is issued (scheduled for 2002) to insure
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that the application will be complete and minimize the need for a protracted license review. The
NRC agreed with DOE's position that work should proceed toward a decision on recommending
the Yucca Mountain site as a repository for high-level waste. ,
There were a number of areas where the NRC Staff did not have major comments at the time of
its review based on general agreement with DOE on the particular issues. These included:
mechanical disruption of the waste packages; radionuclide release rates and solubility limits;
spatial and temporal distribution of flow in the unsaturated zone (UZ); distribution of mass flux
between fractures and matrix in the unsaturated zone; retardation in the UZ fractures; retardation
in the water-production zones and alluvium; dilution of radionuclides in the ground water from
well pumping; airborne transport of radionuclides; dilution of radionuclides in the soil; and
location and lifestyle of the critical group. This is not to say that these processes are
insignificant; rather, there were no significant issues in these areas at the time of the reviews.
Areas where the Staff had significant comments included:
• Repository design
Waste package corrosion
• Quantity and chemistry of water contacting waste packages and waste forms
• Saturated zone flow and transport ;
• Volcanic disruption of the waste packages
• Quality assurance
With regard to repository design, NRC expressed concern as to whether adequate time was
available before the LA is scheduled for submittal to address all the design options under
consideration, select a reference design, develop data and models, and conduct the analyses
required to produce an LA which is complete and of high quality.
Doses received by down gradient receptors are highly sensitive to the corrosion performance of
the waste packages. The DOE is exploring several alternatives to the waste package design used
in the TSPA-VA, which was a 10-cm outer layer of carbon steel corrosion allowance material
and a 2-cm inner layer of Alloy 22 corrosion resistant material. It was not clear to the NRC that
the DOE would be able to gather adequate long-term corrosion data in time to definitively
support the LA. The TSPA-VA relied heavily on expert elicitation rather than long-term test data
and this is a significant weakness. ;
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The amount and chemistry of the water which contacts the waste packages is of critical
importance not only to waste package lifetime bijit also to release of radionuclides once the waste
package is breached. The NRC concluded that "u.the range of activities outlined in the LA Plan
are unlikely to provide an adequate licensing basis for assessing the quantity and chemistry of
water contacting waste packages and waste forms Additional data and analysis of seepage
under both isothermal and thermal conditions wijl be required for a complete LA."
The NRC was not satisfied that flow and transport in the saturated zone from beneath the
repository to a receptor 20 km down gradient had been adequately characterized. Additionally,
the NRC did not concur with the DOE's view that saturated zone uncertainties were a
"moderate" contributor to receptor dose uncertainties. This descriptor was inappropriately
optimistic based on sensitivity studies conducted by both organizations. The Staff expressed
concerns that the location where ground water enters the alluvium (which delays radionuclide
migration) was not well documented. High permeability features between the repository and the
receptor could alter the flow direction away from the alluvium and confine the flow to the
fractured tuffs. ;
Based on Staff review, the NRC concluded that the consequences of volcanism were understated
in the TSPA-VA. The DOE assumptions on physical conditions were not representative of
basaltic volcanism at Yucca Mountain. In addition, the DOE's models did not consider the
impact of the dynamic forces produced by the vojlcanism on waste packages in a volcanic
conduit.
i
i
Implementation of an appropriate Quality Assurance (QA) program has been an on-going
problem. The NRC has reviewed and accepted the DOE's QA program on procedural basis.
However, audits and surveillances have identified deficiencies in implementing the program.
Some data in the technical data bases are not traceable. The NRC is concerned that the LA Plan
did not recognize these implementation deficiencies and provide for remedies.
The NRC staff provided some additional reactions to the TSPA-VA in the May 1999 Technical
Exchange (NRC99b). The TSPA-VA documentation included several features which facilitated
the NRC's understanding of the DOE performance assessment. These included extensive use of
1
plots of intermediate outputs such as time-dependent Tc-99 release from a waste package. Plots
of the performance of sub-systems such as the number of waste packages which failed as a
function of time were also valuable as were dosejrate plots which showed the mean, median, 5th,
and 95th percentiles over tune. The DOE's presentation of the results of sensitivity analyses and
7-1174
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the dose rates expected with alternative conceptual models also enhanced the NRC's
understanding of the TSPA-VA. On the other hand, the NRC felt that there were areas where
transparency and traceability could be improved. The NRC staff noted that the flow of key
information between the REP computer code and external process models was difficult to trace.
The NRC also concluded that there was inadequate sampling of parameters potentially important
to repository performance and they could not determine whether correlations between sampled
parameter had been properly addressed. The Staff suggested that a table listing all important
parameters and their assigned distributions would significantly facilitate review.
The I'JRC felt that both agencies needed to have a better technical basis for establishing the initial
waste package failure levels. Improved linkage was required between initial defects and waste
package failure rates. This would involve consideration of the detectability of initial defects and
consideration of the expected performance of the defective waste packages. Further, with regard
to long-lived waste packages, the NRC averred that there were potential failure processes such as
stress corrosion, microbial activity and exposure to alternating wet/dry cycles which could
accelerate failure. These processes were not considered by either organization. '
The NRC concluded that there were no major performance-affecting differences in the
approaches taken by the two organizations with regard to ground water infiltration and deep
percolation. However, the modeling approaches taken for unsaturated zone flow and transport
differed markedly.
In past near-field modeling the DOE did not consider that penetration of the boiling isotherm in
the drift wall could occur by water flowing down a fracture. The NRC concluded that the DOE's
assumption that water will not contact a waste package until the waste temperature drops below
the boiling point was not conservative. In the TSPA-VA, the drift seepage model was based on
ambient conditions and was not coupled to a thermal model. DOE assumed the first waste
package fails after 1,000 years and is under a drip. '.
The NRC observed that the TSPA-VA methods for calculating biosphere dose conversion factors
(DCF) were consistent with the NRC approach, but the Commission raised some questions as to
whether the procedures used for sampling the DCF distributions created modeling
inconsistencies. The NRC also felt that the documentation on dose parameters used in the
TSPA.-VA needed to be improved.
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The NRC concluded that the model for igneous activity used in the TSPA-VA was inadequate.
Additional work would be required to develop acceptable models. However, based on
discussions between the DOE and the NRC subsequent to publication of the TSPA-VA, the NRC
was of the opinion that acceptable modeling approaches can be developed before the License
Application is submitted.
t
7.3.4.2 Review by the TSPA Peer Review Panel
i
DOE created the TSPA-VA Peer Review Panel to provide the Civilian Radioactive Waste
System Management and Operating Contractor with a formal, independent review and critique of
the TSPA-VA (PRP99). In its review of the Viability Assessment, the Panel was charged with
considering both the analytical approach used and its traceability and transparency in assessing
the probable behavior of the repository. Factors evaluated in assessing the analytical approach
[
included: i
\
• Physical events and processes included in the assessment
Use of appropriate and relevant data
• Assumptions made j
• Abstraction of process models used in total system models
• Application of accepted analytical'methods
• Treatment of uncertainties
The Panel concluded that, due to the complexity pf the system and the nature of the current or
reasonably obtainable data, it may be impossible for any technical team to develop the analytical
capabilities to prepare a credible assessment of the probable future behavior of the repository.
The long time scales which must be considered, coupled with the complexity of the geologic
setting, compound the analytical problems. The Panel suggested that dealing with these complex
coupled processes can best be handled through bounding analyses or by incorporation of
engineered features which minimize the effects of these processes.
In the Panel's words, a credible assessment "would have needed to include:
i
• Component subsystem models thai capture important and relevant phenomena;
• Adequate databases;
• Proper coupling between the subsystem models; and
• Tests of modeled behavior". ]
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Although the TSPA-VA offers many examples of partial, even substantial, success in each of
these four areas, the Panel has also observed examples of important deficiencies in each.
• Concerning subsystem models, the final dose estimates within the TSPA-VA rest
in large part on potentially optimistic, or at least undemonstrated, assumptions
about the behavior of certain barriers in the system (for example, performance of
the cladding and the waste package).
• Concerning databases, some of the important analyses are not supported by an
adequate database, (for example, databases for corrosion of spent fuel alteration
products and the saturated zone analysis).
• Concerning coupled processes (that is, thermohydrological, thermomechanical,
and therniochemical effects) and the data and models that support them, the Panel
believes that it may be beyond the capabilities of current analytical methodologies
to analyze systems of such scale and complexity. For this reason, the effects of
coupled processes can probably best be dealt with through a combination of
bounding analyses and engineered features designed to minimize the effects of
such processes. >
• Concerning tests of modeled behavior, the TSPA-VA does not contain the
convincing direct measurements or confirmation of the modeled behavior of
components or subsystems for which testing is feasible. This testing should be
part of the analyses of such a complicated system."
The Panel concluded that the sensitivity analyses in the TSPA-VA did not provide sufficient
insights to overcome these deficiencies and uncertainties.
The Panel expressed concern over the lack of data relating to the performance of the waste
packages and reliance on instead on expert elicitation. The Panel stated that DOE must define
the environmental extremes to which the Alloy C-22 corrosion resistant liner will be exposed and
establish experimentally the critical temperature for crevice corrosion in these aggressive
environments. The need to obtain more and better data to enhance performance assessment
credibility was a repeated theme throughout the Panel's report.
The behavior of the waste packages is strongly dependent on the extent to which: contact with
infiltration water seeping into the drifts is minimized. The Panel was not convinced that the
TSPA.-VA base case correctly captured seepage into the drifts over long periods of time. The
Panel concluded that "Better characterization of the hydrologic properties near the drifts,
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improved modeling, consideration of coupled effects, and additional experimentation at the drift
scale would add confidence to the approach taken."
The Panel reviewed the impacts of five potentially disruptive processes on the Yucca Mountain
repository. The Panel concurred with DOE findings in the TSPA-VA that impacts of
earthquakes would be minor as would the impacts of volcanism on offsite groups. The Panel
also agreed with DOE's analysis that nuclear criticality was highly improbable and, if it occurred,
only modest increases in offsite doses would be expected. However, the Panel was not satisfied
with DOE's analysis of human intrusion. They stated that the scenario in which the waste
generated from an intruding borehole was driven downward into the SZ was not realistic and
analytical treatment of transport within the saturated zone was potentially non-conservative. The
particular concern with the transport model wasithe assumption that radioactive material was
distributed over a wide area at the top of the SZ.: This would not be the case with the selected
drilling intrusion scenario. The Panel noted that a regulatory basis for analyzing human intrusion
had not been established by either NRC or EPAiat the time when the TSPA-VA calculations
were made. The approach taken on the climate change in the TSPA-VA was judged to be
reasonable, in-so-far as temporal variations in precipitation are concerned. The Panel noted that
the U.S. Geological Survey disputed the manner in which the variation in precipitation was
translated into infiltration rates into the repositoiy but the Panel took no position on that issue.
I
Two potentially non-conservative approaches used in the TSPA-VA were identified by the Peer
Review Panel, namely:
• Long-term performance of Zircaloy cladding on spent fuel
• Buildup of radionuclides in soil irrigated with contaminated groundwater
With regard to cladding performance, the Panel ptated that additional failure mechanisms
including (1) pitting and crevice corrosion, (2) hydride-induced embrittlement and cracking, and
(3) unzipping of the cladding due to secondary phase formation when the UO2 fuel is converted
to various alteration products in a moist, oxidizing environment all need to be experimentally
stigated. Until such work is completed and [the expected cladding longevity can be
invest
substantiated, the TSPA-VA assumptions about [the ability of the cladding to act as a significant
barrier are not defensible. !
The Panel observed that irrigation water was assumed in the TSPA-VA to be deposited on the
soil for only one year prior to intake by the receptor via various soil-related pathways. In reality
irrigation can continue for thousands of years and an equilibrium concentration for each nuclide
i
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will be established in the soil which is higher than that based on only a one-year exposure period.
In addition, the assumption that iodine is rapidly washed through a soil column is not supported
by field observations which show considerable holdup in the surface layers.
The Panel also identified three factors which were believed to treated with significant
conservatism in the TSPA-VA including: ;
Transport through penetrations in the waste package ;
Retention of radionuclides in spent fuel alteration products
Potential sorption of technetium and iodine in the UZ and SZ
The Panel felt that the.modeling of the transport of radionuclides from failed waste packages
through pits, cracks or crevices was not realistic since no significant retardation was included.
Since this assumption is not consistent with expected physical reality, better methods are
required to analyze the movement of radionuclides within and from the failed waste packages.
Any UO2 in spent fuel packages which is exposed to moist air is expected to be converted to
secondary uranium minerals such as schoepite within a few hundred years after waste package
and cladding failure. It has been experimentally established that neptunium would be
incorporated into the alteration products and, consequently, Np release would be controlled by
the dissolution rate of these alteration products. While this process was not included in the
TSPA-VA base case, it was cursorily examined in a sensitivity analysis (DOE98, Volume 3,
Section 5.5.3). No impact was shown over the first 10,000 years or after about 700,000 years
because releases are dominated by other nuclides for those time periods. However, at 100,000
years, the dose rate is reduced by about a factor of 10 when solubility of Np from the alteration
products is considered. '
No sorption of technicium or iodine (the major contributors to dose over the first 10,000 years)
on geologic materials was considered in the TSPA-VA. However, the Peer Review Panel cited
field observations, such as those of Straume et al. (STR96), taken near the site |of the Chernobyl
nuclear power plant accident suggesting that radioiodine may be retarded in soil surface layers.
The Panel did not cite any instances where technetium was retarded but suggested that the issue
should be reviewed on the basis, for example, of measurements near the Chernobyl site.
In addition to these general conclusions, the Panel provided detailed comments on all of the
component models used in the TSPA-VA including the UZ flow, thermohydrology, near-field
geochemical environment, waste package degradation, fuel cladding as a barrier, waste form
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degradation, radionuclide mobilization, UZ transport, SZ flow and transport, biosphere, and
disruptive events. Recurring themes were the need for additional data and improved models to
produce a credible and defensible LA. i
i
7.3.4.3 Review by the U.S. Nuclear Waste Technical Review Board
i
The Nuclear Waste Technical Review Board (NV^TRB; see Section 4.4 of this BID) also
critiqued the TSPA-VA (TRB99). The Board stated that they had identified no features or
processes which would disqualify the Yucca Mountain site but felt that DOE should give serious
attention to replacing the high-temperature design! evaluated in the TSPA-VA with a ventilated
low-temperature design where waste package surface temperatures were maintained below the
boiling point of water. Such a change should significantly reduce the uncertainties involved in
attempting to analyze complex coupled thermal-hydraulic and thermal-mechanical, and thenmal-
geochemical interactions within the repository. |
-j
The NWTRB also expressed concerns as to whether the amount of work required to support a
technically defensible decision on Yucca Mountain could be completed on DOE's proposed
schedule, which calls for a site recommendation decision by 2001. This is a matter of particular
concern, since the Board stated that expert elicitation should not be used as substitute for data
gathering at the site or in the laboratory. Areas where additional factual input is required include
waste package performance (e.g., resistance to stress-corrosion cracking), and the magnitude and
distribution of seepage into the repository.
The Board also stressed the need for long-term scientific studies assuming the site is ultimately
found to be suitable and construction is approved.1 These scientific studies should include
selected aspects of both natural and engineered barriers.
In summary, the Board agreed with DOE "that Yucca Mountain continues to merit study as the
candidate site for a permanent geologic repository and that work should proceed to support a
decision on whether to recommend the site to the President for development. ... The Board
supports continuing focused studies of both natural and engineered barriers at Yucca Mountain to
attain a defense-in-depth repository design and to increase confidence in predictions of repository
performance."
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7.3.5 NRC Total System Performance Assessments
7.3.5.1 Background
To support it licensing responsibilities, the NRC is developing the capability to review DOE's
TSPA in support of a License Application, if the Yucca Mountain site is found to be suitable for
disposal. The Commission staff, like DOE, is iteratively developing TSPA modeling capability
based on evolving information and insights concerning factors that affect repository system
performance. Development of the TSPA methodology is independent of DOE's effort, and the
DOE and NRC TSPA models and codes differ in detail. :
The NRC's strategic planning calls for early identification and resolution, at the staff level, of
TSPA issues before receipt of an LA, if the Yucca Mountain site is found to be suitable for
disposal. The principal means for achieving this goal is on-going, informal, pre-licens'ing
consultation in which performance issues are identified and discussed, and issue resolution is
sought. Resolution of issues is sought at the staff level before formal licensing reviews, but
issues may be raised and considered again in the licensing process. •
To implement its goals, the NRC has focused its pre-licensing work on issues most critical to the
post-closure performance of the proposed repository; these have been designated as Key
Technical Issues (KTI). To facilitate dialog with DOE concerning resolution of the KTIs, the
NRC has established Issue Resolution Status Reports (IRSR) to serve as the primary mechanism
through which feedback to DOE concerning KTIs and KTI subissues will be expressed and
documented. The IRSRs address acceptance criteria for issue resolution and the status of
resolution. Updating revisions of the IRSRs will be issued periodically as progress is made in
resolution of the KTIs and their subissues.
One of the Key Technical Issues identified and discussed in an IRSR is Total System
Performance Assessment and Integration (TSPAI). The NRC has, to date, issued the original
version of the IRSR on this topic in April 1998 and Revision 1 in November 1998 (NRC98). As
basis for its review of the DOE TSPA and development of its own TSPA methodology, the staff
has adopted the hierarchical structure of performance assessment factors shown in Figure 7-41.
This performance factor structure was used to develop the NRC TSPA code structure (e.g., TSP
3.x.y) illustrated in Figure 7-42. This code structure can be compared to DOE's TSPA-VA code
structure shown in Figure 7-36.
7-181
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7-183
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• Transparency and Traceabilitv of the Analysis. This subissue emphasizes staff
expectation of the contents of DOE's TSPA to support an LA. Specifically, it
focuses on those aspects of the TSPA that will allow for an independent analysis
of the results."
Details of acceptance criteria and review methods for the subissues related to demonstration of
overall performance and demonstration of multiple barriers will be provided in the next revision
of the IRSR for TSPAI. Details of criteria and review methods for model abstraction, scenario
analysis, and transparency are included in NRC98.
7.3.5.2 NRC Development and Use of TSPA Models
The content and characteristics of NRC's TSPA models have, like DOE's, evolved over time as
information and insights as basis for the models have developed. Current models, also like
DOE's, are considered to be a snapshot in time from an on-going model-development process.
Under its Iterative Performance Assessment (EPA) program, NRC has adopted a phased approach
to its TSPA modeling capability. Phase 1 used relatively simplistic models and was designed
primarily to demonstrate capability to perform TSPA reviews as part of the licensing reviews.
Phase 2 used significantly enhanced modeling methods to identify and assess factors of primary
importance to repository system performance. Phase 3, which is still underway, uses more
i
general and versatile computer codes to perform TSPA evaluations analogous to those performed
by DOE.
Three versions of the Total-system Performance Assessment (TPA) code have been developed in
Phase 3 of the IPA program. TPA 3.1.3 has been used to calculate mean doses'for alternate
conceptual models, and TPA 3.1.4 has been used for system-level sensitivity and uncertainty
studies. The most recent version of the TPA code, 3.2, was used to provide feedback to DOE on
the results of NRC's review of the TSPA (see Sections 7.3.2. and 7.3.3). !
The most recently documented description of the NRC TPA codes is provided in NUREG 1668,
which describes the characteristics and use of the 3.1.3 and 3.1.4 codes to perform sensitivity and
uncertainty analyses for a proposed repository at Yucca Mountain (NRC99a). Characteristics of
the TPA 3.2 code have not yet been documented, but results of its use were presented and
discussed at the May 1999 DOE/NRC Technical Exchange (NRC99b) in which NRC staff
provided feedback to DOE concerning results of their review of the TSPA-VA.
7-185
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The TPA 3.1 and 3.2 codes have capability and flexibility comparable to those of the DOE codes
for the TSPA-VA. As previously noted, the DOE and NRC codes differ significantly in detail,
but both have capability to evaluate performance for alternative repository design features,
natural system features, and disruptive scenarios, at a level of detail and characterization of
uncertainty commensurate with the available information base. At present, the principal
difference between the NRC and DOE performance assessment codes is that the NRC codes give
considerable attention to disruptive events associated with seismicity and volcanism, while the
DOE approach considers these phenomena to be unlikely to occur in ways that could affect
repository performance. These differences are expected to be resolved as part of the issue
resolution process. |
Principal features of the NRC's Phase 3 performance assessment codes include the following:
Water infiltration into the subsurface. Calculation of percolation flux takes into
account the time history of climate change, variation of shallow infiltration with
climate change, and the areal-average percolation flux at the repository horizon.
Near-field environment. The near-field environment, which affects the waste
package corrosion rate, is characterized in terms of drift wall and waste package
surface temperatures, relative humidity, water chemistry, and water reflux during
the thermal pulse phase.
|
Waste package degradation and BBS release. Waste package failures depend on
near-field conditions, corrosion mechanisms and rates, and mechanical effects
such as rockfall. Radionuclide release from the BBS is calculated in terms of rate
of release from the waste form, solubility limits, and transport mechanisms out of
the BBS. No credit is taken in the base case for cladding performance as a barrier.
Transport in the UZ and SZ. Time-dependent flow velocities in the UZ are
calculated using the hydrologic properties of the major hydrostratigraphic units.
Matrix and fracture flow are modeled. Radionuclide retardation on fracture
surfaces is assumed not to occur, but sorption in the rock matrix is modeled. The
conceptual hydrologic model for flow in the SZ assumes fracture flow in the tuff
aquifer and matrix flow in the alluvial aquifer.
Airborne transport for direct releases. NRC performance assessments include
consideration of airborne releases from low-probability intrusive igneous events
which cause direct release of waste package materials into the air. Factors
considered include number of packages failed and quantities of radionuclides
released, ash deposition patterns, and degradation of deposited, contaminated ash.
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• Biosphere dose exposure scenarios. Dose evaluations are done for the average
person in a designated receptor group. Two types of groups are considered: a
farming community 20 km downgradient from the repository, and a residential
community: The farming community is assumed to use contaminated ground
water for drinking and agriculture; the residential community uses it only for
drinking. Dilution of radionuclide concentrations in the ground water as a result
of pumping is considered.
NUREG-1668 (NRC99a) reports the results of dose evaluations in which the base case TSP 3.1.4
model and 11 alternative conceptual models (such as including cladding credit) were used to
calculate doses at 10,000 and 50,000 years for a receptor 20 km from the repository. The
repository system conceptual design was similar to that used by the DOE in the TSPA-VA, but
the corrosion-resistant inner package barrier was assumed to be Alloy 625. The annual base case
mean peak total effective dose equivalent (TEDE) was projected to be 2.3 mrem at 10,000 years.
Annual results for the alternative conceptual models ranged from a low of 0.012 mrem when
cladding credit was taken to a high of 12.5 mrem when no radionuclide retardation was assumed.
The range of results is shown as a bar chart in Figure 7-43.
As previously noted, the NRC presented its more recent TSP 3.2 results evaluations at the
DOE/NRC Technical Exchange in May 1999 (NRC99b). Results presented for the ground water
dose using the NRC's mean-values data set are shown in Figure 7-44, for 10,000 and 100,000-
year dose rates. As can be seen, the 10,000-year dose rate is forecasted to be about 0.002
mrern/yr, and the 100,000-year dose is about 0.2 mrem/yr. These results can be1 compared to
DOE's TSPA-VA results, which indicated a 10,000-year dose rate of 0.04 mrem/yr and a
100,000-year dose rate of about 5 mrem/yr (see Figures 7-37 and 7-38). Reasons for differences
in the NRC and DOE results are not readily apparent because parameter values ;and modeling
approaches used by the two agencies differed markedly. For example, the DOE assumed
cladding credit while the NRC did not; the NRC assumed an average of 32 juvenile waste
package failures while the DOE assumed one; the DOE used three-dimensional modeling of UZ
below the repository which suggested significant lateral diversion while the NRC used one
dimensional modeling with seven stream tubes and no lateral diversion, hi addition, the NRC
assumed dilution during pumping of contaminated ground water by the dose receptor, while DOE
assumed this dilution did not occur.
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Ctf
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7.3.5.3 Conservatism In The NRC Performance Assessments '
As noted in Section 7.3.5.2, the NRC staff are independently developing performance assessment
capability in order to be able to be able to perform comprehensive reviews of DOE's TSPA in the
License Application. The NRC performance assessment capabilities and methods are, like
DOE's, continuing to evolve. Documentation of NRC's parameter values, models and
assumptions are not yet as comprehensive as DOE's; the most recent description of the NRC
models and the results of their use was provided in the NRC/DOE Technical Exchange of May
27-29, 1999 (NRC99b). As reported during the Exchange, NRC's base-case performance
evaluations using VA design parameters projected a 10,000-year dose rate of about 0.003
mrem/yr; DOE's base-case 10,000-year dose rate projection was 0.04 mrem/yr. :Conservatisms
in NRC's performance parameters, models, and assumptions, as indicated by information
provided at the Technical Exchange, are summarized below.
Performance Parameters
NRC presentations at the May 1999 Technical Exchange indicated that "mean values" of the
performance parameters were used for the base case performance assessments. Values of some
of the parameters were presented, but comparisons with DOE are difficult because of differences
in modeling approaches and parameters used. In general, NRC's use of "mean values" appears to
correspond in concept to DOE's use of "expected values." Values of parameters used by NRC
for precipitation and infiltration were, for example, similar to those used by DOE.
!
Performance Models
Key features of NRC's performance assessment modeling approach that are indicative of
conservatism include the following:
• Impacts of igneous events, seismic rock falls, and fault displacements on waste
packages were included in the models. Seismicity impacts were included in the
base case evaluations; volcanism and faulting impacts were treated separately.
• No credit was taken for spent fuel cladding as an engineered barrier. Half of the
spent fuel in a failed waste package was assumed to be exposed, wetted, and a
source for release of radionuclides.
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Transport of radionuclides in the unsaturated zone from the repository to the water
table was assumed to occur vertically, with no effect of matrix diffusion or
sorption on fracture surfaces. This assumption is similar to that made by DOE in
the TSPA-VA. ;
• Radionuclide transport in the saturated zone was assumed to occur via four
pathways through fractured tuff and alluvium. Transport in the tuff occurred only
via fractures, with flow rates between 50 and 500 m/yr. Flow velocities in the
alluvium were assumed to be between 3 and 5 m/yr, and radionuclide retardation
was assumed to occur.
• Dilution of radionuclide concentrations in ground water as a result of pumping by
the dose receptor was assumed to occur (the dilution factor was not stated). This
is a non-conservative modeling feature in contrast with DOE's assumption that
such dilution does not occur.
Conservative Assumptions
Conservative assumptions in the NRC performance assessments described at the May 1999
Technical Exchange (NRC99b) included the following:
Thirty-two waste packages were assumed to be defective at the time of
emplacement. Rates and mechanisms of degradation and radionuclide release for
these and other packages that fail were not described, however. ',
The mean value of the localized corrosion rate for the Alloy 22 corrosion resistant
material in the waste package was stated to be 2.5 E-4 m/yr. This is a factor of
100 higher than experimental values cited in EPRI's EVIARC-4 report (EPR98)
and in DOE's VA Technical Support Document (DOE98a). ;
Detailed comparison of NRC and DOE performance assessment conservatisms is not possible
\
because the modeling approaches and parameters used differ significantly. In general, it appears
that, in comparison with DOE, NRC's approach produces a larger radionuclide source term (e.g.,
as a result of assuming no cladding credit), but compensates for it by assuming that dilution
occurs during pumping. The net result is that the results of NRC's performance assessments
reported at the May 1999 NRC/DOE Technical Exchange agree with DOE's TSPA-VA results
within an order of magnitude.
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7.3.6 BPRI Total System Performance Assessments
7.3.6.1 Background 1
i
The nuclear power utilities have for many years maintained oversight of the OCRWM program
in DOE because of their contracts with the Department concerning its responsibilities for receipt
and disposal of commercial spent fuel. Technical contributions to the oversight are provided by
EPRI in programs that are selected and guided by the utilities. EPRI maintains peer capability to
review and comment on DOE's program activities and to independently perform performance
assessments and other analyses of the type done by the Department within the OCRWM
program.
i
EPRI has performed independent total system performance assessments in parallel with DOE's
efforts. A report on EPRI's TSPA concepts and methods was first issued in 1990 (EPR90), and
TSPA reports were subsequently issued in 1992, 1996, and 1998 (EPR92, 96, 98). A more
recent EPRI report (EPROO) represents more of a critique of the supporting documentation for
the TSPA-SR than an independent assessment. The EPRI studies have kept pace with the DOE
efforts, making use of the evolving repository design concepts, data bases, and modeling
methods. The EPRI Phase 4 report, issued in November 1998 (EPR98) parallels the DOE's
TSPA-VA report (DOE98) and uses the VA design. The EPRI Phase 5 report, issued in
November 2000, provides a critique of the models and assumptions in the DOE's TSPA-SR.
The overall goal of the EPRI assessments is to provide an "...independent assessment of the
performance of the potential repository site, identifying fatal flaws in the site itself, in the
engineering design, or in the licensing program, so that the decision makers in the utility industry
can judge the likelihood of potential outcomes of the licensing process and take appropriate
action" (EPR96). i
I
\
7.3.6.2 EPRI's TSPA Technical Approach [
i
EPRI uses a logic tree approach to performance assessment modeling. The EPRI TSPA code is
termed the Integrated Multiple Assumptions and Release Calculations code (IMARC). The logic
tree approach, illustrated in Figure 7-45, represents uncertain inputs to the TSPA calculations as
nodes in a tree, with branches from a node indicating alternative models or parameter values for
that input and the weight associated with that model or parameter value. In contrast, the DOE
TSPA code structure (Section 7.3.2.2) and the NRG approach (Section 7.3.4) use a central
7-192
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processor (e.g., the RIP code for DOE), which is fed information from codes for the various
repository performance factors.
CLIMATE
SCENARIO
FOCUSED SOLUBILITY ;
INFILTRATION FLOW AND ALTERATION RETARDATION
FACTOR TIME
Figure 7-45. EP-RI's MARC Logic Tree (EPR98)
All TSPA methods include models for essentially the same performance factors, ,e.g., climate,
infiltration, waste package performance, etc. They differ, however, in the details of how they
model the performance factors and in their assignment of values for uncertain performance
parameters. For example, the DOE assumed three climate conditions for the TSPA-VA, with
precipitation spanning the range 170 to 540 mm/yr; in contrast, the EPRI interpreted the historic
climate data to indicate two future climate conditions, with precipitation spanning the range 150
to 220 mm/yr.
Other key features of the EPRI Phase 4 TSPA modeling approach are outlined below. As for
DOE, details of models and parameter characterization have evolved in accord with evolution of
the data bases for performance assessment. Because the modeling approaches used in IMARC-4
were similar to those used in IMARC-3, the IMARC-4 report (EPR98) did not repeat technical
details of modeling that were discussed in EPR96.
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Climate
I
i
As indicated above, EPRI's interpretation of available data concerning past and possible future
climate conditions led to an estimate that the long-term average precipitation should be between
ISO and 220 mm/yr, a much narrower range than used by DOE in the TSPA-VA. EPRI believes
the DOE precipitation values are based on ostracode species assemblages found in Minnesota
and Washington, rather than on specific plant taxa calibrated near Yucca Mountain.
Infiltration \
\
The basic MARC net infiltration model is a one-dimensional finite difference code that
incorporates source and sink terms for surface infiltration, uptake of water by plants, and
drainage from the root zone to the deep subsurface (which is net infiltration). For Phase 4, the
runoff features of the model were revised as a result of recent data. As a result, the net
infiltration for current climate conditions increased from the Phase 3 (1996) value of 1.2 mm/yr
to 7.2 mm/yr. The full glacial climate value increased from 2.9 to 19.6 mm/yr. (DOE's TSPA-
VA values showed similar increases in comparison with TSPA-95 values.) The TSPA-VA
results are higher than the Phase 4 results because the DOE assumed a precipitation rate of 300
mm/yr as compared to EPRJ'sassumption of 195 mm/yr for a full glacial climate.
|
Near Field Conditions \
\
\
For IMARC-4, EPRI developed a model and analytic solution which describes heat transfer and
fluid flow in the near field in terms of a uniform djisk-shaped heat source located in a moist,
unsaturated, porous medium. Large-scale convectiive gas flow and countercurrent flow of water
and vapor were assumed to occur. Heterogeneity of the repository's geohydrologic regime was
represented by what was termed "focused flow, and "hot" and "cool" zones of the repository
were characterized. The objective of the modeling was to estimate that fraction of the waste
package inventory that is wetted; results indicated that the maximum fraction of the waste
packages that are wetted is 0.24. In contrast, DOE's expected values in the TSPA-VA for waste
packages with seeps were about 0.5 during superptavial conditions and about 0.33 during the
extended periods associated with long-term average climate (DOE98, Volume 3, Figure 4-3).
7-1^4
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Waste Package Performance
The waste package performance model used in IMARC-4 differed significantly from that used in
EVIARC-3 because of improved understanding of the repository environment and corrosion
processes, and because the reference corrosion resistant material (CRM) was changed from Alloy
825 to Alloy 22. The basis for characterizing corrosion rates was changed from Weibull
distributions25 to recently-obtained corrosion data and the results of DOE's expert elicitation on
waste package performance. Corrosion rates were characterized for various environmental
conditions, e.g., humid air or water dripping onto the package, and for various corrosion
mechanisms, including crevice corrosion of the Alloy 22, which is anticipated to represent the
mechainism fdr most-rapid penetration of the CRM. Results for the VA waste package design
(see Section 7.2) show that, in the absence of drips onto the package, penetration would not occur
for more than one million years. When drips do contact the packages, penetration by general
corrosion is predicted not to occur for about 30,000 years. Under adverse conditions, the carbon
steel outer wall could be penetrated in only 300 years, and the Alloy 22 inner wall could be
penetrated by crevice corrosion, which is conservatively assumed to occur during the time period
during which the waste package temperatures are greater than about 80°C. The EPRI estimates
that "hot" waste packages would remain above the 80°C threshold for crevice corrosion for about
3,000 years. For "cold" waste packages this period would be reduced to about 200 years. The
EPRI notes in IMARC-4, as did the DOE in the TSPA-VA, that the data base foriestimating
Alloy 22 corrosion rates is currently quite limited.
Source Term Parameters
Source term parameters discussed in IMARC-4 include radionuclide sorption, solubility, release
from the waste form, and waste form alteration. Values for these parameters were changed in
IMARC-4 in comparison with IMARC-3 because of recent data additions. The computer code
COMPASS, Version 2.0, which is a compartment model for predicting radionuclide release rates
from the engineered barrier system (EBS) into the near-field rock, was used in IMARC-4. The
i
Compass 2.0 code models EBS features, such as waste form, canister corrosion products,
backfill, and rock fractures, as compartments. It accounts for time-dependent cladding
degradation, modes of water contact with the waste package, and modes of water'transport
through the waste package interior (overflow or through-flow).
25
A Weibull distribution is a function used to describe the fraction of waste packages which have failed as
a function of time based on mean container lifetime, threshold failure time and failure rate at the mean lifetime.
7-195
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I
Discussions of source term parameters in IMARC-4 addressed the following:
• New values of sorption coefficients for sorption of radionuclides on corrosion
products (principally iron oxides) were presented for cases where recent data
differ from results of a prior expert elicitation by more than a factor of five.
Median values for the actinides are in the range 5-10 m3/kg; the median value for
Np is 0.1 m3/kg. \
I
Extensive discussion was presented on the validity of the two-orders-of-
magnitude reduction in the solubility of Np in the TSPA-VA in comparison with
TSPA-95. The EPRI analyses basically concurred with the action, which was
based on re-assessment of prior data and additions to the data base for solubility
values. The solubility of neptuniun) is important to prediction of doses after
10,000 years, when neptunium is thp principal contributor to dose.
l
• Extensive discussion was provided concerning thin films surrounding spent fuel
undergoing dissolution. The EPRI concluded that the TSPA-VA approach was a
"sensible, but non-unique first step in attempting to derive more realistic
radioelement solubility constraints from laboratory tests." The EPRI
recommended additional modeling and laboratory tests to establish lower, more
realistic solubility constraints. !
[
Flow and Transport in the Unsaturated and Saturated Zones
The flow and transport models used in EVIARC-4 were the same as those used in Phase 3.
Values used for parameters were revised, however! as a result of recent insights concerning
conceptual modeling of the UZ and SZ and continuing integration of field and theoretical studies.
'
The IMARC-4 UZ hydrology model accounts for transient, variably-saturated flow and
advective-dispersive transport in a coupled dual-porosity-dual-permeability regime, from the
base of the repository to the water table. Radionuclide sorption can occur both in the fractures
and in the rock matrix. In the SZ, the model takes [into account three-dimensional advective-
dispersive transport of the radionuclides during do^vn-gradient migration. The SZ model can
handle matrix diffusion, radionuclide sorption and daughter-product ingrowth.
The repository footprint can be divided into subregions, each of which constitutes the top of a UZ
hydrologic column. Input variables such as infiltration rates can therefore be varied over the area
of the repository. The model assumes that there is'no lateral coupling between the columns and
that the system is isothermal, so that no coupling to the energy equations is needed.
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Once the radionuclides reach the water table, they can advect, disperse, sorb, diffuse into or out
of the matrix, and decay within the three-dimensional SZ. Ground water flow in the SZ is
assumed to be representative of long-term steady-state conditions. The bulk hydraulic
' conductivity of the fractured rock mass is assumed to be representative of an equivalent porous
medium, which may be anisotropic.
IMARC-4 discusses the impact of recent determinations that the net infiltration rate is much
higher than originally believed, and the discovery of bomb-pulse Cl-36 at repository depths, on
conceptual modeling of the UZ. It also discusses the impact of current lack of data for the SZ on
uncertainty in the flow paths and dilution factors for the SZ. It notes that EvIARQ-3 asserted that
overall dilution for the SZ was about a factor often, and that this value is retained in EVIARC-4
and corresponds to the base case value used by DOE in the TSPA-VA. It also discusses dilution
for a small radionuclide plume, such as would result from a single package failure, and asserts
that the dilution factor for this situation would be on the order of 100,000.
Biosphere
The EPRFs IMARC analyses use a probabilistic model to estimate radiation doses. The model
has three basic parts: probabilistic modeling of releases from the repository, characterization of
dose conversion factors for the biosphere pathways 'and the nuclides of interest, and
characterization of the dose receptor. In IMARC-4, EPRI used a farming critical group and the
water-only pathway for their base case. Other possible dose circumstances (e.g., all pathways)
were also evaluated. The critical group was assumed to be located 5 km from the boundary of
the repository, i.e., at the boundary for release to the accessible environment as defined by 40
CFR Part 191.
The hypothetical critical group was assumed to extract ground water from the point of highest
contamination in the contaminant plume, and to use this contaminated water for all of their food
and water needs for their entire lifetime. Dose conversion factors were based on ICRP
definitions of dose established in 1991 and on IAEA recommendations for metabolism of the
elements established in 1994.
7.3.6.3 Results of IMARC-4 Dose Evaluations :
The EPRFs IMARC-4 analyses produced base case results for conditions and assumptions
outlined above, and also produced results for a wide range of sensitivity analyses. The EPRI
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[
base case results are shown in Figure 7-46. These results were obtained assuming that 0.01
percent of the waste packages had failed at emplacement (i.e., one package) and that 0.1 percent
failed soon after emplacement (i.e., 1,000 yr). Thbse early failures may be caused by
manufacturing defects, construction errors, or emplacement mishandling. The EPRI modeling
assumes no corrosion failures during the initial 10,000 years while the DOE modeling assumes
that 17 waste packages will fail by corrosion during this period. Thus, the EPRI assumption for
total waste package failures (juvenile plus corrosibn) is 11 while the equivalent DOE assumption
is 18. |
I
i
The dose receptor was assumed to be an average member of a farming community located 5 km
from the repository, and the doses are the result of exposure only via the ground water pathway.
When all exposure pathways were included, the dose rate variations as a function of time were
similar to those shown in Figure 7-46, but about a factor often higher. This indicates that, for
the EPRI modeling approach for the critical group, the drinking water contribution to dose is
minor in comparison with the agricultural and other pathways.
Comparison of Figure 7-46 with the results of the; DOE TSPA-VA analyses, Figure 7-39, shows
that the dose rates at various times are generally similar (e.g., DOE projects a dose rate at 10,000
years of 0.04 mrem/yr; EPRI projects 0.08 mrem/yr), and the sources of dose are similar, i.e., Tc-
99 and 1-129 are dominant in the near term and Np-237 is dominant in the long term. In the
I
EPRI results, Figure 7-46, the decrease in dose rate over the interval 60,000 to 100,000 years is
the result of depletion of the Tc-99 and 1-129 inventories for release from the repository.
EPRI IMARC-4 results are compared to DOE's TSPA-VA results and NRC's TSP 3.2 results in
Section 7.3.7. >
7.3.6.4 Conservatism In The EPRI Performance Assessments
As indicated in Section 7.3.6.2, the EPRI approach to total system performance assessments
differs markedly from those used by DOE and NRC. As a result, direct comparison of EPRI
conservatism with that of DOE and NRC is neither possible nor appropriate. In general, the
IMARC-4 report (EPR98) suggests that EPRI seeks to be as realistic as possible in all aspects of
its assessment efforts. For example, EPR98 criticizes the DOE interpretation of data concerning
past climates as being too conservative, observes that the assumption of an early package failure
is arbitrary, and notes that the EPRI and TSPA-VA approaches to modeling of fracture /matrix
interactions in the saturated zone differ markedly!
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Figure 7-46. Results of EPRI's IMARC-4 Dose Evaluations (EPR98)
In contrast to DOE's adoption of expert opinion as the basis for waste package material corrosion
rates, EPR98 includes a comprehensive effort to develop parametric models of corrosion
behavior on the basis of available data. Like NRC, the EPRI EMARC-4 analyses; take no credit
for spent fuel cladding as a barrier. However, in contrast to NRC's bathtub model, EPRI uses a
flow-through model for water entry to and exit from the interior of a failed waste! package. This
is similar in concept to DOE's approach, which assumed that radionuclides are instantaneously
released to the EBS from the wetted waste form.
The EVIARC-4 report, EPR98, includes a discussion which compares the IMARC-4 and TSPA-
VA results. The report states:
"We observe that the magnitude of the doses estimated by IMARC Phase 4 are in
general agreement with those in the TSPA-VA (within an order of magnitude for
all time periods). This agreement can be considered quite close, given that the
models, level of abstraction, and input parameters for particular FEPs [features,
events, and processes] are considerably different between the two analyses.
Whether this is simply fortuitous or speaks to the robustness of the combined
7-199
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analyses is not altogether clear. It may be ^hat one particular combination of
conservatisms (andpotential non-conservatisms) in one TSPA effort were, on the
whole, balanced by a different combination ofconservatisms/nonconservatisms in
the other TSPA analysis. There is certainly some evidence for this.
\
In the end, this independent comparison of TSPA approaches for the proposed
Yucca Mountain repository provides further confidence that the major FEPs
controlling the overall safety ofthe facility\have been identified. "
7.3.7 Comparison of DOE, NRC. and EPRI TSPl\ Results for the VA Repository
Although the TSPA models, assumptions, and parameter values used by DOE, NRC and EPRI
differed greatly, each of the TSPA evaluations discussed above (DOE's TSPA-V A, NRC's TSP
3.2, and EPRI's IMARC-4) has as its basis the VA repository design concept, key features of
which are the waste package design (an outer wall of carbon steel and an inner wall of Alloy 22),
and an areal heat loading of 85 MTU/acre. Despite widely different modeling concepts, and with
only the principal design features of the repository land the existing data base as the basis for
commonality of the analyses, the results of the three TSPA efforts are quite similar, as shown in
Figure 7-47. j
In Figure 7-47 the EPRI results are decreased by a factor often in comparison with the actual
results because the EPRI dose receptor was assumed to be located only 5 km from the repository.
This location, in comparison with the 20 km distance assumed by DOE and NRC, would not
have achieved the SZ radionuclide concentration reduction as a result of dilution that was
assumed for the DOE and NRC analyses. Decreasing the EPRI results by a factor of 10 therefore
puts all results on essentially the same basis with rkspect to the SZ dilution factor.
The similarity of the three sets of TSPA results may be the fortuitous consequence of offsetting
assumptions. For example, DOE's TSPA-VA took credit for cladding performance as a barrier
but took no credit for dilution during pumping; NRC's assumptions were the opposite of these.
Conversely, the similarity may be due to the dominant influence on results of performance
factors for which the three analyses made similar assumptions, e.g., those concerning future
climate conditions and early waste package failure^. For all analyses, the dose rate results at
10,000 years are dominated by radionuclide releases from packages that were assumed to fail
7-20,0
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relatively soon after repository closure, and by the highly mobile Tc-99 and 1-129 isotopes whose
arrival at the dose receptor location is not significantly affected by assumptions concerning
phenomena along the UZ and SZ pathway. !
i
After EPA and NRC post-closure radiation protection standards for a possible repository at
Yucca Mountain are established, opportunities for differences in assumptions concerning the
dose receptor and biosphere pathways will be narrowed. Similarly, the need for assumptions
concerning performance parameter values will be reduced by future additions to the data base.
However, alternative TSPA modeling approaches; can and will be maintained.
r
i-
7.3.8 Performance Assessments in the Yucca Mountain DEIS
DOE issued its Draft Environmental Impact Statement (DEIS) for disposal of highly radioactive
wastes at Yucca Mountain in August 1999 (DOE99a). The DEIS used, as the basis for the
Proposed Action, the VA repository design and basic TSPA methodology. Details of TSPA
methodologies used for the DEIS and the VA differ, however, because of the differences in the
scopes and purposes of the DEIS and the VA. [
The VA and the TSPA-VA focused on a single radiation-dose receptor location and a reference
design for engineered features of the repository; variations in performance were evaluated as a
result of variations in the values of performance parameters for this single, fixed system. In
contrast, because of the scope of options considered in the DEIS, the TSPA-DEIS considered
alternatives for dose receptor locations, waste quantities and types, and repository designs.
Consideration of these options made it necessary ;to modify some of the details of the TSPA-VA
[
methodology for the TSPA-DEIS evaluations. \
7.3.8.1 Comparison of Bases for the DEIS and v TSPA Evaluations
A principal cause of the differences in TSPA methodologies for the VA and the DEIS is the
difference in waste quantities and types considered. As mandated by the Nuclear Waste Policy
Act of 1982, the VA considered 70,000 MTU of emplaced wastes, assumed to comprise 63,000
MTU of commercial spent fuel and 7,000 MTU-equivalent of DOE spent fuel and high-level
wastes from defense production operations. The DEIS used these waste quantities as the basis
for the Proposed Action, but also considered othejr amounts and types of wastes.
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The DEIS alternatives to the quantities and types of wastes considered in the VA and the
Proposed Action were designated as Modules 1 and 2. Module 1 increases the quantity of
commercial spent fuel from 63,000 to 105,000 MTU. The latter quantity is based on the Energy
Information Administration's assumption that all currently operating reactors would extend their
licenses for a period of 10 years; the quantity represents the expected maximum discharge of
commercial spent fuel through 2046. Module 1 also includes all DOE spent fuel and high-level
wastes not included in the Proposed Action. This addition increases the mass of such wastes in
the repository from 7,000 MTU-equivalent, for the VA reference design and the DEIS Proposed
Action, to about 14,000 MTU-equivalent for the Module 1 option.
Module 2 includes the waste inventories of Module 1 plus inventories of other types of wastes
that are candidates for disposal at Yucca Mountain. These are designated as Greater-Than-Class-
C (GTCC) wastes and DOE Special-Performance-Assessment-Required wastes. The quantities
of these wastes are characterized in terms of their volume, which is estimated in the DEIS to total
about 6,000 cubic meters. This would correspond to 8 percent of the total volume of Module 2
wastes. The incremental effect of the additional wastes under Module 2 is small in terms of
repository radionuclide inventories and their long-term releases from the repository.
Consideration of Module 1 and Module 2 waste inventories and alternative thermal loadings of
85, 60, and 25 MTU/acre produces great variation in the potential size of the repository footprint.
Figure 7-48 shows the emplacement blocks identified to accommodate the DEIS disposal
inventory options, with the emplacement area for the VA reference design (63,000 MTU
commercial spent fuel; 7,000 MTU-equivalent DOE wastes; 85 MTU/acre) shown as the shaded
area in the so-called Upper Block. For emplacement of all wastes considered in the DEIS
(Modules 1 and 2), at the lowest thermal loading considered (25 MTU/acre), all of the area of all
of the blocks shown in Figure 7-48 would be used for waste emplacement. All blocks shown in
Figure 7-48 are in essentially the same stratigraphic horizon, i.e., that used for the TSPA-VA.
The repository emplacement areas associated with the options considered are summarized in
Table 7--11, reproduced from Table 1-25 in Appendix I of the DEIS.
Table 7-11. DEIS Estimates of Waste Emplacement Areas :
j MTV/acre
85
60
1 25
Drift Spacing,
meters
28
40
38
Emplacement Area,
740
1,050
2 520
Emplacement Area,
1,240
1,750
7-202
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Base Case Inventory
Hijh Thcnrnl Lxwdrnp
Figure 7-48. Emplacement Block Layout for DEIS Disposal Option
7-204
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As indicated by Table 7-11, under the options considered by the DEIS, the repository
emplacement area could be, at the highest inventory and lowest thermal loading, six times greater
than for the VA reference design (i.e., 4,200 acres vs. 740 acres). This potential range of
repository conditions necessitated modifications to the TSPA models that were used in the
TSPA-VA for the reference VA repository.
The performance assessment codes and their configuration used for the TSPA-DEIS assessments
were basically the same as those used for the TSPA-VA (see Section 7.3.2 and Figure 7-36 of
this BID). Details of some of the codes were, however, modified for the DEIS evaluations.
Basic performance assumptions (e.g., those concerning juvenile failure and wetting of waste
packages, amount of spent fuel failed at emplacement, etc.) were the same for both sets of
evaluations. •
To accommodate the DEIS options for waste inventories, thermal loading, and dose receptor
location, the TSPA-VA codes were modified in four areas: thermal hydrology, waste package
degradation, waste form dissolution, and elements of the RIP code. Modifications to the RIP
code were concerned with the repository environment, the FEHM model, stream tubes, and
radioriuclide transport paths. Details of the modifications are provided in the DEIS support
document, "EIS Performance-Assessment Analyses for Disposal of Commercial and DOE Waste
Inventories at Yucca Mountain" (DOE99b).
!
7.3.8.2 Results of the TSPA-DEIS Evaluations ;
The TSPA-DEIS calculated mean dose rates at 10,000 and one million years for the "average
individual" as defined in the VA (Section 7.3.2 of this BID). Biosphere dose conversion factors
used in the TSPA-DEIS evaluations were the same as those used in the TSPA-VA. The
biosphere dose conversion factors were assumed to have the same value at all dose receptor
locations considered (5, 20, 30, and 80 km) even though the 5 km distance, which is not suitable
for irrigation or farming, would be a drinking-water-only pathway, and the 80 km distance is a
lake playa, where discharge and evaporation of contaminated water would produce deposits of
contaminated dust. The assumption of constant dose conversion factors for all distances is stated
in the DEIS to be conservative because development of location-specific factors for the 5- and
80-km distances would result in values for the factors that are lower than those actually used.
Results of the TSPA-DEIS dose evaluations analogous to those obtained for the TSPA-VA
evaluations are presented in Chapter 5 of the DEIS in tables showing mean peak dose rates for
7-205
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the Proposed Action inventory and the alternative [dose receptor distances and areal mass
loadings considered. Results for the Proposed Action inventory and 10,000-year doses, which
are presented in Tables 5-4, 5-8, and 5-12 of the DEIS, are summarized in Table 7-12.
Table 7-12. Peak Dose Rates at 10,000 Ypars for the Proposed Action Inventory
and Alternative Distances and Thermal Loads
Dose Receptor
Distance, km
5
20
30
80
Peak Dose Rates at 10,000 Years, mreni/yr
85 MTU/acre
0.32 j
0.22* j
0.12 !
!
0.03 1
60 MTU/acre | 25 MTU/acre
0.14
0.13
0.05
0.003
0.13
0.06
0.04
0.0005
* Corresponds to TSPA-VA base case distance arid thermal loading conditions
As noted in the Table 7-12 footnote, the result for the 20-km distance and the thermal loading of
85 MTU/acre (0.22 mrem/yr) corresponds to the base case evaluation in the TSPA-VA, which
showed a mean 10,000-year dose rate of 0.1 mrem/yr for 100 probabilistic realizations and a
10,000-year mean dose rate of 0.04 mrem/yr for the deterministic base case evaluation in which
expected values of all parameters were used. t
l
The time variations of peak dose rates for times up to 10,000 years in the TSPA-DEIS
evaluations are shown, and compared to the TSPA-VA results, in Figure 7-49. Since all basic
modeling conditions and assumptions were the same for the two sets of evaluations, differences
in the details of the curves can be ascribed to the modifications made to the TSPA-VA codes for
I
the TSPA-DEIS evaluations. The similarity of the curves supports the DOE contention that the
model modifications for the DEIS evaluations hacl only a minor impact on results.
The TSPA-DEIS results for time history of peak doses over a period of one million years are
shown in Figure 7-50, which was included in the DEIS documentation as Figure 4.1-3 of the
support document, DOE99b.The curves for distances of 5, 20, and 30 km, which are virtually
indistinguishable in Figure 7-50, are also virtually identical to the results obtained in the TSPA-
VA evaluations, presented in the VA documentation as Figure 4-13 of Volume 3 of DOE98. The
DEIS support document, DOE99b, speculates that the low dose rates for the 80 km distance are
the result of greater dispersion and radionuclide holdup in the alluvium which, in the
performance assessment models, becomes the flow and transport medium at distances beyond
about 25 km. , !
7-206
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0.1-
0.01 - —
0.001-
,§ O.O001 -
O.O0001 -
DEIS Mean. High Thermal Load
DEIS Mean, Low Thermal Load
VA Base Case (Fig. 4-28. VA)
VA Expected-Value (Fig. 4-13, VA)
1.000 2.000 3.000
4.0OO 5.000 6.000
Poslclosure lime (years)
7.000 8.000 9.000 ; 10.00O
Figure 7-49. Time History of Projected Dose to 10,000 Years; VA and
DEIS Evaluations (DOE99a) •
i • 1 i ... i ....
i , | ' 1 • ' -'• — ' — '
20km
• 30km
S km
— • flOkm
• 1
I
0 200,000 400,000 600,000 800,000 1,000,000
Time (years)
Figure 7-50. DEIS Dose Rate Time Histories for Periods Up to One Million Years (COE99a)
7-207
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As shown by Figures 7-49 and 7-50, the TSPA-V^. and TSPA-DEIS dose rate results are
consistent with each other because the performance assessment models, parameters, and
assumptions used in the two sets of evaluations were essentially the same. Both sets of
evaluations therefore used the conservative assumptions discussed in Section 7.3.3.5 of this BID,
and both therefore substantially understate the expected performance of a repository with the
design features (i.e., the VA design) that were the basis for the evaluations and the results
obtained. If the performance evaluations for the Fpal EIS are revised to reflect the design
features expected to be the basis for the Site Recoinmendation, i.e., the EDA II design discussed
in Section 7.2.2.5 of this BID, the projected performance of the repository would be greatly
improved in comparison with the results obtained for the DEIS. Preliminary performance
assessment results for the EDA II repository are discussed in Section 7.3.9, and these evolved
into the full TSPA for Site Recommendation (TSPA-SR), discussed in Section 7.3.10. The
results of the TSPA-SR clearly show dramatic improvement in performance in the 10,000-year
time period.
7.3.8.3 DEIS Evaluations of Radionuclide Concentrations in Ground Water
[
•
The DEIS used the VA design and modeling methods to calculate ground water concentrations of
radionuclides released from the repository. Results were obtained for the various waste
inventory, thermal loading, and dose-receptor distance options within the DEIS scope.
The results of the DEIS concentration evaluations for the radionuclides released during periods
up to 10,000 years and transported to locations 5, |20, and 30 km downstream from the repository
are summarized and compared to the current Maximum Concentration Limits (established for the
Safe Drinking Water-Act in 1976) in Table 7-13. [The predicted ground water concentration
values are strongly influenced by the assumed waste package failure at 1,000 years and by
assumptions of limited dilution during transport. [AS a result of the assumptions that maximize
the release from the repository and minimize dilution during transport, the radionuclide
concentrations shown in Table 7-13 are much higher than would be predicted for more realistic
performance assumptions. [
7-208
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Table 7-13. Comparison of DEIS Ground Water Concentrations With MCLs -
All Concentrations in picoCuries/liter.
Radionuclide
Contributors to
lOK-YrDose
Tc-99
1-129
C-14
Current MCL,
inpCi/L
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
95 Percentile
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
1.9
14
6.3
0.40
0.15
0.0
5.6
5.9
0.21
The apparent inversions of concentrations with distance are a consequence of the modeling methods used for the
DEIS performance evaluations. ;
As can be seen in Table 7-13, the concentrations reported in the DEIS for the VA repository are
well below the current MCL values despite the conservative assumptions that are the basis for the
concentration evaluations. These assessments (both the TSPA-VA and DEIS) have, subsequently
been shown to be conservative compared to current design and TSPA-SR analyses.
Consequently, the primary conclusion that the results are below the MCLs remains current and
appropriate. TSPA-SR analyses have shown that there are no anticipated releases that can be
compared with the MCLs during 10,000 years (see Section 7.3.10.)
7.3.9 Preliminary TSPA Results for the EDA II Design
The current design concept for a repository at Yucca Mountain, EDA II, is discussed in Section
7.2.2.5. DOE intended that the engineered design features used for the VA were a step in
repository design evolution, which would culminate in the design that would be used for the
License Application if the Yucca Mountain site is approved for disposal. During 1999, the
Department therefore defined and assessed alternative improved designs and selected the EDA II
concept, discussed in Section 7.2.2.5, as the concept to be used for the Site Recommendation.
This design concept was subsequently used as the basis for the TSPA-SR, described in Section
7.3.10. ;
7-209
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7.3.9.1 Performance Factors Basis for the EDA II Design
I
The basis for selection of the EDA II design was, in large measure, the findings and
recommendations that emerged from reviews of thi VA design concept by parties such as the
Nuclear Waste Technical Review Board (TRB99), the NRC staff (NRC99c), and the Total
System Performance Assessment Peer Review Panel (PRP99). The reviewers determined that
some of the engineered features of the VA repository contributed significantly to uncertainty in
the TSPA-VA results and would raise technical issues that would be difficult to resolve during
licensing reviews. '
Major design features of the VA repository that contributed to TSPA-VA performance
uncertainty included: [
\
The high thermal loading, 85 MTUJ'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
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 dnfts 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. !
I
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 corrosibn of the Alloy 22, thereby increasing the rate
of penetration of the Alloy 22 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, mobilize the radionuclides, and transport the
radionuclides to the environment. 1
As a result of these concerns, DOE adopted the EDA II design described in Section 7.2.2.5, i.e., a
design with major features consisting of use of drip shields, use of Alloy 22 as the outer wall of
the waste packages, increased spacing between drifts, and a thermal loading reduced from 85
MTU/acre to 60 MTU/acre. A comparison of principal design features of the VA and EDA II
7-210
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designs, and the impact of the EDA II design on the uncertainties associated with the VA design,
is shown in Table 7-14. As can be seen from this table, the design changes are responsive to the
concerns cited above.
Table 7-14. Impact of EDA II Design Features on VA Performance Uncertainties
Identified by Reviewers of the TSPA-VA
Design Feature
Areal thermal loading
Drift Liner and
Invert Material
Drift Spacing
Waste Package Materials
Peaik Waste Package
Power
Drip Shield
Backfill
VA Repository
85 MTU/acre
Concrete
28 meters
10 cm. Carbon steel over
2 cm. Alloy 22
95% above average
None
None
EDA n Repository
60 MTU/acre
Steel
81 meters
2 cm. Alloy 22 over
5 cm. 3 1 6L stainless
20% above average by
blending assemblies
2 cm. Titanium 7
Yes
EDA H Impact
Reduce thermal coupling
issues
Eliminate effect of
concrete materials on
water chemistry; reduce
corrosion rates and
radionuclide release rates;
increase package lifetime
No temperature rise
above boiling point in
rock between drifts;
reduces overall
performance uncertainty
Eliminate crevice
corrosion potential;
reduce Alloy 22
penetration rate by a
factor of 25 or more;
increase 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 '
7.3.9.2 Evolution of the Repository Safety Strategy ;
In conjunction with adoption of the EDA II design, DOE has also revised its repository safety
strategy. As stated in the Rev 4 Repository Safety Strategy document (TRVVOO), the strategy
7-211
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continues to rely on multiple natural and engineered barriers to achieve safety performance, but
reflects the EDA II enhanced engineered design and improved understanding of the principal
performance factors. The revised strategy also is rpsponsive to modifications in the regulatory
1 framework for the potential repository system. It reflects the NRC's change from use of
prescriptive subsystem performance objectives to use of a risk-informed, performance based
approach, with a total system performance assessrnent built on defense-in-depth and safety
margins. i
•
Implementation of the revised strategy has been accomplished in part by refining and increasing
the number of performance factors considered in modeling and assessment of performance. In
contrast to the 19 performance factors considered in the TSPA-VA, the Rev 3 strategy considers
27 performance factors, with seven of them identified as factors of principal importance, as
shown in Table 7-15. The principal factors are: seepage into drifts; performance of the drip
shield; performance of the waste package barriers; solubility limits of dissolved radionuclides;
retardation of radionuclide migration in the unsaturated zone; retardation of radionuclide
migration in the saturated zone; and dilution of radionuclide concentrations during migration.
A technique defined as "neutralization analysis" was used to help identify the principal factors.
In using this technique, specific performance factors are removed from the performance
modeling system to determine its impact on overall'performance. An example of the use of
neutralization analysis, reproduced from TRWOO,|is shown in Figure 7-51.
7.3.9.3 Results of early TSPA Evaluations for the JED A II Design
Interim results of some TSPA evaluations for the EDA II repository were published leading up to
the completion of the TSPA for Site Recommendation (TSPA-SR). These interim results clearly
illustrate some of the advantages of the EDA II de|sign, and are reproduced in this section for the
sake of completeness and clarity. However, some of the issues raised in these interim analyses
were justifiably eliminated in the subsequent TSPA-SR, and the reader is referred to Section
7.3.10 for the latest information on TSPA. '•
I
Figures 7-52 and 7-53 show results presented at trie NWTRB meeting of June 1999 (TRB99a).
Results presented in the TSPA-VA (DOE98) have been added to the figures to show the
comparison with TSPA-VA results. From the similarity of the shape of the curves in Figure 7-
52, it is evident that the basis for the TSPA-VA and TSPA-EDA II analyses was at least similar,
7-212
-------
if not the same, e.g., assumption of juvenile waste package failure at 1,000 years, accompanied
by the other performance assumptions detailed in the TSPA-VA (DOE98, Vol.3).
In contrast to the results shown in Figures 7-52 and 7-53, The Rev 3 repository safety strategy
document, TRWOO, states that the base case TSPA analysis for the EDA repository shows that
radionuclide releases are not expected for more than 100,000 years. These results; take into
account the expected performance of the titanium drip shields and the waste packages with Alloy
22 as the corrosion resistant wall material on the outside.
A scenario which assumed common-mode failure of a drip shield and the underlying waste
package at 9,000 years produced the radionuclide release and dose results shown in Figure 7-54,
which is reproduced from Figure 2-2 of TRWOO. This dose/time-history curve is the same as the
base case for the neutralization analysis shown in Figure 7-51.
The early-failure scenario represented by the dose/time curve shown in Figure 7-54 is extremely
improbable. Its probability of occurrence is on the order of 10"9, estimated as follows:
I
One drip shield out of 10,000 has to fail by being penetrated by water.
For the common-mode failure to occur, the failed drip shield has to be over a
waste package that is vulnerable to early failure as a result of a phenomenon such
as a poor weld. About 0.1 percent, or one out of 1,000 packages, might have such
vulnerability.
The failure in the drip shield has to be over the vulnerable part of the waste
package surface, e.g., the weld bead. If the hole in the drip shield is on the order
of 2 cm (one inch) in diameter, and the waste package is on the order of 600 cm
(six meters) in length, the chance of the drip shield hole being over the vulnerable
part of the package surface is on the order of one in 100.
Combination of these factors produces the result that the probability of the early common-mode
failure leading to nuclide releases after about 10,000 years, as shown in Figure 7-54, is
(10"4)(10"3)(10"2) = (10~9). This is the same order of likelihood of occurrence as that for volcanic
intrusion of the repository. This result, in combination with the base case results showing no
radionuclide releases for 100,000 years, suggests that the basic paradigm for TSPA methodology
and assumptions for EDA II repository performance evaluations should be significantly different
from that used for the TSPA-VA analyses.
7-213
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Table 7-15. Factors Potentially Important to Postqlosure Safety (TRWOO)
Key Attributes of the
Limited Water
Contacting Waste
I Package
.
0
Long Waste Package
j Lifetime
J Low Rate of
I Radionuclide Release
from the EBS
Delay and Dilution of
11 Radionuclide
| Concentrations During
Transport Away from
the EBS
Principal Factors in the Viability Assessment
Precipitation and Infiltration into the Mountain
Percolation to Depth
Seepage into Drifts
Effects of Heat and Excavation on Flovy
N/A
N/A
Dripping onto the Waste Package
-lumidity and Temperature at the Wasti:
^ackage |
Chemistry on the Waste Package
Integrity of Outer Waste Package Barrier
Integrity of Inner Waste Package Barrier
Seepage into Waste Package
Integrity of Commercial Spent Nuclear Fuel
(CSNF) Cladding
Dissolution of UO, and Glass Waste Fjjrms
Solubility of Neptunium-237
Formation of Radionuclide Bearing Colloids
Transport within and out of the Waste Package
EBS Radionuclide Migration — Transport
Through Invert
Transport through Unsaturated Zone ( JZ)
Transport in the Saturated Zone (SZ)
Dilution from Pumping
Climate
nfiltration
Unsaturated Zone (UZ) Flow above Repository
Seepage into Drifts*
Coupled Processes — Effects on UZ Flow
Couples Processes — Effects on Seepage
Environments on Drip Shield
Performance of Drip Shield*
Environments on Waste Package
Performance of Waste Package Barriers*
Environments within Waste Package
Commercial Spent Nuclear Fuel (CSNF) Waste
Form Performance
DOE-Owned Spent Nuclear Fuel (DSNF), Navy 11
Fuel, and Plutonium Disposition Waste Form II
Performance
Defense High-Level Waste (DHLW) Waste Form II
Performance
Solubility Limits of Dissolved Radionuclides* ||
Colloid Associated Radionuclide Concentrations II
In-Package Radionuclide Transport
Transport through invert
Advection Pathways in the UZ
Retardation of Radionuclide Migration in the UZ*
Colloid Facilitated Transport in the UZ ||
Coupled Processes — Effects on UZ Transport ||
Advection Pathways in the Saturated Zone (SZ)
Retardation of Radionuclide Migration in the SZ*
Colloid Facilitated Transport in the SZ
Dilution of Radionuclide Concentrations during
Migration*
Biosphere Transport and Uptake
*Principal factors of the postclosure safety case.
7-214
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1.E+3
1-E+2
CB1.E+1
o>
.>>
fl.E+0
e>
in
o
Q
I.E-I
•J.E-2
ra
3
cl-E-3
1.E-4
1.E-S
1,000
Proposed Annual Dose Limit
Neutralize transport barrier
function of saturated zone
Neutralize transport barrier
function of unsaturated zone
Neutralize flow barrier
function of overlying rock
10,000
Time (year after closure)
100,000
Figure 7-51. Barriers Importance Analysis to Assess Natural Barriers of the
Repository System-Early Waste Package Failure Scenario (TRWOO)
104
103
102
All Pathways, 20 km
10"
j - 1 - r
EDA-II, backfill
EDA-ll, no backfill
VA (Fig. 4-12, p. 4-23, Vol.3
0
2,000 4,000 6,000 8,000 10,000
Time (years)
Figure 7-52. Comparison of VA and EDA 10,000-Year Doses (TRB99a)
7-215
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All Pathways, 20 km
VA (Fig. 4-21, p. 4-51, Vol. 3)
Preliminary Draft
EDA-II, backfill
EDA-ll, no backfill
0 200,000 400,000 600,000 800,000 1,000,000
Time (years)
Figure 7-53. Comparison of VA and EDA Million-Year Doses (TRB99a)
The "Performance Factor" results presented in Table 7-8 of this BID and the TSPA-VA results
presented in DOE98 Vol. 3 can be combined to produce the comparison of major radiation dose
milestones for the VA and EDA H repositories shown in Table 7-16. As can be seen from the
values in this table, the expected performance of the EDA II repository is significantly better than
that of the VA repository. This is the result of the; design features selected specifically to
improve expected performance and to reduce uncertainties in expected performance and in the
results of TSPA evaluations. As previously noted (TRWOO), the EDA II repository would not be
expected to release radionuclides and to cause radiation doses for more than 100,000 years.
The peak dose values in Table 7-16 are the result of use of highly conservative assumptions used
in the TSPA-VA and TSPA-EDA II evaluations, such as the assumption that all of the area of the
waste form in a commercial fuel rod with breached cladding is exposed to, and contacted by,
water that enters the interior of a waste package whose wall is penetrated by water (see Section
7.3.3.5 of this BID). Use of more realistic assumptions would reduce the estimated peak dose
rates by several orders of magnitude.
7-216
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Table 7-16. Comparison of Major Dose Milestones for the VA and EDA II Repositories
Dose Parameter
Dose at 10,000 Years, mrem/yr
Time to 15 mrem/yr, years
Time to 25 mrem/yr, years
Peak Dose During One Million Years
Time of Peak Dose, years
VA Repository
0.04
70,000
90,000
350 mrem/yr
320,000
EDA II Repository
0.001
290,000
310,000
85 mrem/yr
650,000
7.3.10 Performance Evaluation for the Site Recommendation
Since selection in 1999 of the EDA II design (Section 7.2.2.5) to be the basis for the Site
Recommendation (SR), and publication of Revision 3 of the Repository Safety Strategy in
January 2000 (TRWOO), DOE continued to evolve and refine the design and strategy concepts for
use in performance evaluation for the SR. This evolution has led to the recent publication of Rev
4 of the Repository Safety Strategy in November 2000, and the publication of the Total System
Performance Assessment for Site Recommendation (TSPA-SR). This last document has now
replaced the TSPA-VA as the current iteration of the TSPA for Yucca Mountain. In this section,
the key features of the Repository Safety Strategy (Section 7.3.10.1) and the TSPA-SR (Section
7.3.10.2) are described. Very recently, the EIS for Yucca Mountain has been supplemented
7.3.10.1
Evolution of the Repository Safety Strategy
Key features of Revision 3 of the Repository Safety Strategy (TRWOO) are summarized in
Section 7.3.9.2 above. That version of the strategy was developed to be responsive to comments
on the VA design and TSPA models, and to conform to anticipated regulatory requirements. The
Rev 3 strategy is reflected in the EDA II design, described in Section 7.2.2.5.
Revision 4 of the Repository Safety Strategy was described by DOE in the June 6-7 Technical
Exchange and compared to the Revision 3 strategy (DOEOOa). The comparison of important
factors in the Rev 3 and Rev 4 strategies is summarized in Table 7-17. This comparison
indicates that Revision 4, in comparison with Revision 3, is based on more extensive analyses,
improved performance assessment models, a broader data base, and an integrated and extended
evaluation of factors important to performance.
7-217
-------
Table 7-17. Comparison of Rev 3 and
Rev 4 Repository Safety Strategies
Strategy Element
Revision 3
Revision 4
Principal Factors
Subjective judgments about
performance factors
Judgments supported by barrier
neutralization analyses
No consideration of disruptive
factors
Based on extensive TSPA analyses
and barrier importance analyses
TSPA includes both nominal and
igneous activity scenarios
Performance Assessment
Use of VA models
Updated, fully-documented models
Probabilistic analyses to address
uncertainties
Measures to Increase
Assurance of Safety
Preliminary consideration of safety
margins and defense in depth
Initial plans for safety assurance,
performance confirmation
Full evaluation of safety margins
and defense in depth
Full documentation of potentially
disruptive features, events, and
processes
Rev 1 of Performance Confirmation
Plan •
Rev 4 of the Repository Safety Strategy revises the Principal Performance Factors established in
Rev 3, which are discussed in Section 7.3.9.2 and shown in Table 7-15. Rev 4 retains all of the
Rev 3 principal factors except wellhead dilution, which is no longer considered to be a principal
factor because dilution factors are expected to be determined by regulation. Rev 4 adds four new
principal factors to the remaining Rev 3 list, which includes waste package performance,
seepage, drip shield performance, dissolved radionuclide concentrations, and UZ and SZ travel
times. The principal performance factors added bjy Rev 4 are:
I
I
Colloid-associated radionuclide concentrations, added for defense-in-depth.
•
• Probability of igneous activity. Radionuclide release is possible in less than
10,000 years; risk depends on probability of occurrence.
Effects of igneous activity on the repository. Risk depends on damage to waste
packages and drip shields. [
Biosphere dose conversion factors* Radionuclide release is possible in less than
10,000 years; risk depends in part pn biosphere transport and uptake.
7-2^8
-------
Overall, the barriers seen to be potentially most important to waste isolation include the
overlying rock, drip shield and waste package performance, and the UZ and SZ radionuclide
transport barriers. Other important barriers are the commercial spent fuel cladding, HLW
canisters within the HLW waste packages, the drift invert, and the inner waste package barrier,
which is planned to be 100 mm of stainless steel (DOEOOb). ;
The DOEOOa presentation also identified and addressed potential performance assessment
vulnerabilities under the Rev 4 strategy. These vulnerabilities and the envisioned means to
mitigate them are shown in Table 7-18.
Table 7-18. Potential Performance Assessment Vulnerabilities and Mitigation Measures
Potential Vulnerabilities
Adequacy of treatment of model uncertainty
Over-conservatism in some models
Thermal loading issues
Potential for igneous activity
Reliability of complex metal barriers
Possibility of peak dose rate exceeding 100 mrem/yr
Mitigation Measures
Mitigate through defense in-depth, including analysis
of rockfall effects
Studies to assess appropriatenes's of less conservatism
in key models
Improve basis for thermal design for LA;
Use flexible design that can be modified after
performance confirmation testing; maintain options
until final selection
Demonstrate low probability, low risk, jand high margin
Use defense-in-depth and alternative EBS concepts
Reduce conservatism in key models, i.e., solubilities of
Np and Pu, UZ and SZ flow and transport
DOE completed documentation of Rev 4 of the Repository Safety Strategy in November 2000.
7.3.10.2
TSPA-SR
The methodology used for the SR (TSPA-SR) was described by TRWOOb. The methodology is a
step in Iterative development of TSPA methodology, in that it includes improvements in
comparison with VA models and methodology that are responsive to comments on the VA, and
that traceability has been improved in comparison with the VA. The TSPA approach has also
been modified for the TSPA-SR to meet anticipated EPA and NRC regulatory requirements.
Documentation of the TSPA-SR methodology, models, and bases is included in Analysis/Model
Reports (AMRs) and Process Model Reports (PMRs) that are derived from the AMRs. There are
more than 100 AMRs and 9 PMRs.
7-219
-------
The reference repository design for the TSPA-SR is basically the EDA II design discussed in
Section 7.2.2.5 of this BED. Key design features include:
i
Average thermal load of 62 MTHM/acre
• 50 years of ventilation !
Pre-emplacement blending of fuel to level the thermal load
Titanium drip shields over the waste packages
No backfill |
End-to-end loading of waste packages, with close spacing
All waste packages have 20 mm ouj:er layer of Alloy 22 and 100 mm inner layer
of stainless steel; weld stresses mitigated by laser peening
I
An important feature of the TSPA-SR in comparison with the TSPA-VA is that the TSPA-SR
starts with a comprehensive identification and screening of features, events, and processes
(FEPs) that can affect repository system performance (the VA used assumed conditions and
omitted some FEPs). Scenarios for the retained FEPs are constructed and screened, and
implementation of the scenarios in the TSPA is specified. The TSPA-SR includes scenarios for
expected ("nominal") conditions, and scenarios foi: low-probability disruptive events: volcanism,
and human intrusion. [
t
TSPA dose projection results were produced for the nominal conditions, the volcanic scenarios,
the human intrusion scenario, and ground water concentrations in comparison with the EPA
ground water protection standards. Unlike the earlier TSPA-VA, the nominal and igneous
disruptive event are combined via their probabilities into a single probability-weighted dose
consequence curve that is compared with performance objectives. Evaluation results include
characterization of uncertainties and their significance.
i
The TSPA-SR performance assessment model is an upgraded version of the TSPA-VA model.
The core dose calculation model, which receives input from models of performance factors such
as UZ and SZ flow and transport, is now termed GoldSim and is an improved version of the RD?
code used for the TSPA-VA. The basic code configuration for the TSPA-SR evaluations is
similar to that for the TSPA-VA, which is shown in Figure 7-36.
The TSPA-SR methodology differs significantly from the TSPA-VA methodology in that it
includes consideration of disruptive events retained after screening of all FEPs. As described in
the DOE/NRC June 6-7, 2000 Technical Exchange (DOEOOc), Igneous activity and seismic
damage to cladding were screened into the performance scenarios to be considered. Seismic
effects on cladding were included in the evaluations for nominal performance of the repository.
7-220
-------
All other potentially disruptive events were screened out, i.e., faulting, ground motion damage to
the BBS, rockfall, seismically-induced water table rise, and nuclear criticality were eliminated
from consideration.
DOE developed the human intrusion scenario for the TSPA-SR to be consistent with existing
guidance in the draft 40 CFR 197 (EPA99), the proposed version of 10 CFR 63 (>!lRC99), and
the projposed version of 10 CFR 963 (DOE99a). The implementation of the regulatory
requirements was conducted in the TSPA-SR as shown in Table 7-19 (TRWOOb). 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 more reasonably be degraded
enough to be unrecognizable as an intrusion event, was treated as a sensitivity case study.
Table 7-19 Implementation of regulatory requirements in the TSPA-SR for regulatory
requirements (Table excerpted from TRWOOb).
Assumed intrusion is a drilling
event.
Assumed intrusion ins acute and
inadvertent.
Inadvertent drilling event
Drilling result is a single, nearly
vertical borehole that penetrates
a waste package and extends
down to the SZ.
Borehole penetrates a degraded
waste package.
Single vertical borehole from
surface through a single waste
package to the SZ.
Intrusion occurs 100 years after
closure
Intrusion time should take into
account the earliest time after
disposal that a waste package
could degrade sufficiently that
current drilling techniques could
lea to waste package penetration
without recognition by the
drillers.
Intrusion occurs at 100 years (a
10,000 year intrusion time is
examined in a sensitivity '•
simulation).
Borehole properties (diameter,
drilling fluids) are based on
current practices for resource
exploration.
Borehole results from
exploratory drilling for
groundwater.
Borehole diameter consistent
with an exploration groundwater
well.
7-221
-------
NRC Base Assumptions (from
Proposed 10 CERPart 63)
EPA Additional and/or *
i«- _ " »l, vj.%
Conflicting Assumptions (from
Proposed 40 CFR Part 197) ^
Conceptualization for TSPA-
• ^ SR
3orehole is not adequately
sealed to prevent infiltrating
water.
Natural degradation processes
gradually modify the borehole,
the result is no more severe than
the creation of a groundwater
flow path from the crest of
Yucca Mountain through the
[
potential repository and to the
water table. t
Infiltraton and transport through
the borehole assumes a degraded,
uncased borehole, with
properties similar to a fault
pathway.
Hazards to the drillers or to the
public from material brought to
the surface by the assumed
intrusion should not be
considered.
Only consider releases through
the borehole to the SZ; consider
releases occur gradually through
air and water pathways, not
suddenly as with direct removal.
Groundwater is only pathway
considered.
A separate consequence analysis
is required, identical to the
performance assessment, except
for the occurrence of the
specified human intrusion
scenario.
Unlikely natural processes and
events are not included, but
analysis could include}
disturbances by other processes
or events that are likel^ to occur.
Intrusion borehole is applied to
nominal case; effects of
volcanism are not included.
Peak dose is not to exceed 25
mrem/yr. in the first 10,000
years,
Peak dose is not to exceed 15
mrem/yr. In the first 10,000
years.
Does not affect simulations.
The approached used in TSPA-SR for evaluating these conditions are shown in Table 7-20. 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.
7-222
-------
Table 7-20 Technical Assumptions Implemented in the Human Intrusion Scenario in TSPA-SR
(Table excerpted from TRWOOa).
Issue
** a "^
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
sroperties
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
-TSPA-SR Implementation ?
Typical water well borehole has a diameter of 20.3 cm (8
in.)
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-13
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 hte potential repository.
Uncertainty in location is captured in infiltration rate and
location that radionuclides enter the SZ. II
Borehole length from the potential repository to SZ II
conservatively assumes water level consistent with glacial
transition climate.
Assume SZ flow and transport properties i identical to
nominal scenario.
A.ssume exposure pathways and receptor characteristics
identical to nominal scenario ||
7-22
-------
In the igneous disruption scenario, a dike is assumed to intersect the repository as a result of
volcanic activity. Probabilities of an intrusion event resulting in a dike intersecting the
repository, and of volcanic eruption as a result of dike intrusion, are established by expert
elicitation. Repository response to the dike intrusion, and dose consequences of volcanic
eruption, were evaluated on the basis of repositoryj design features, site data, and use of data from
the 1995 Cerro Negro eruption as an analog.
i
i
The eruption was assumed to be a violent strombolian type eruption. 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 groundwater from the disrupted waste
packages. Biosphere dose conversion factors were modified to account for the effect of ash
discharge and fallout on dose potential as a result of the eruption on biosphere pathways.
i
i
The TSPA-SR methodology also differs significantly from that of the TSPA-VA in its treatment
of saturated zone flow and radionuclide transport. | The presentation at the June 6-7, 2000
DOE/NRG Technical Exchange (DOEOOd) described the changes as follows:
i
The 3-D SZ site-scale flow and transport model is used to simulate radionuclide
transport in the TSPA-SR (vs. the streamtube approach in the TSPA-VA)
I
Radionuclide concentrations are calculated in the water supply of the hypothetical
farming community in TSPA-SR (ys. concentration in the SZ, as in TSPA-VA)
Matrix diffusion is explicitly
(vs. use of the effective porosity
TSPA-VA)
simulated in the SZ site-scale model for TSPA-SR
approach for transport in fractured media used in
Particle tracking method used for
for TSPA-SR (vs. finite element transport
VA)
7-224
radionuclide transport in the SZ site-scale model
method used in streamtubes for TSPA-
-------
Minor sorption of Tc and I in alluvium for TSPA-SR (vs. no sorption in TSPA-
VA)
The SZ site-scale flow and transport model was calibrated using data from ongoing DOE and
Nye County drilling and measurement programs. At present, an "Alluvial Uncertainty Zone" has
been defined as a result of limited data. The northern boundary of the alluvium varies across the
entire uncertainty zone; the western boundary of the alluvium varies approximately from the
Fortymile Wash channel to the tuff outcrops in the west; and the flow path length in the alluvium
(to the 20-km radius) varies from about 1 up to 9 km..
The TSPA-SR approach to biosphere modeling is basically the same as that which was used in
the TSPA-VA. The definition of dose receptors was based on the definitions in the draft EPA
and NRC regulations, e.g., the EPA's Reasonably Maximally Exposed Individual 'and the NRC's
Critical Group. An analysis of uncertainty in ground water usage by a hypothetical farming
community was performed on the basis of current water usage and demographic data. An
analysis of radionuclide buildup in soils from long-term irrigation with contaminated ground
water was also performed for the TSPA-SR.
The TSPA-SR methodologies and information outlined above were included in the AMRs and
the PMRs, which provide detailed documentation of technical approaches and justifications used
in the TSPA-SR.
Results for the TSPA-SR are shown in Figure 7-54 for the igneous and nominal scenarios. There
are no doses during the first 10,000 years from the nominal scenario. This is a substantial change
for the TSPA-VA results, in which the design led to a significant potential for juvenile releases.
These juvenile releases have been eliminated by the EDA II design, with the results shown in the
figure. Doses during the period up to 2000 years are dominated by the igneous eruption scenario
(TRWOOb). From the period between 2000 years to after 10,000 years, the igneous intrusion
scenario becomes the most important scenario (TRWOOb). Subsequent to that time, all three
scenarios (nominal, eruption, and igneous intrusion) play a role in establishing the dose curve.
TRWOOb also reported the results of uncertainty analyses that led to the mean dose rates shown
in Figure 7-54.
7-225
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E
(D
Q)
•4—*
CO
CD
CO
o
Q
c:
CO
CD
NRC 25 mrem/yr IPS Limit
EPA s Proposed IPS and HIS
EPA s Proposed GWS
.X Igneous Mean
Dose Rate
Combined Mean Dose Rate
/ Nominal Mean Dose Rate
i I I
10
1000
10,000
Time (years)
100,000
Figure 7-54 Comparison of Proposed Radiation Protection Standards with Expected Values of
TSPA-SR Calculations for a Repository at Yucca Mountain for Nominal and Igneous Scenarios
(Figure adapted from TRWOOb). \
\
Results from the TSPA-SR human intrusion analysis are shown in Figure 7-55! The base case
analysis is the result of an assumed intrusion event at 100 years after closure. Also shown on the
figure is the result of an inadvertent intrusion even at 10,000 years after closure.
-
f
Additional results were presented by TRWOOb for| comparison with proposed groundwater
criteria. The TSPA-SR indicates that no radionuchde 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 7-56 and 7-57. The groundwater 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 groundwater protection is 10,000 years. However, the analyses were carried out
to 100,000 years to ensure that no significant degradation of the performance occurs after 10,000
years. 1
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I I I I M I I
10,000 Year Package
Failure Time
I | I I I I I I
1000 10,000
Time (years)
100,000
Figure 7-55. Expected Values of TSPA-SR Calculations for a Repository at Yucca Mountain for
the Inadvertent Human Intrusion Scenario (Figure adapted from TRW.OOb).
10°
"E 1°2
E
CD
J,
0)
CO
iS -
I 1°-2
^ io-3
C
ro
OJ
-| 0-4
g 10-5
O 10-6-
1000
1-129
I I I I I I I l\
10,000
Time (years)
100,000
Figure 7-56. Summary of Groundwater Protection Performance Results of the TSPA-SR:
Combined Beta and Photon-Emitting Radionuclides (Figure adapted from TRWOOb).
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103
"ra
10°
' 10-2
10-4-
Gross Alpha lActivity (excluding Rn and U)
Total Radium Activity (Ra-226 and Ra-228)
J L
1000
10,000
Time (years)
100,000
Figure 7-57. Summary of Groundwater protectioiji Results for TSPA-SR for Gross Alpha
Activity (Figure adapted from TRWOOb). |
i
I
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 Ar, alysis 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 the system.
I
Uncertainty importance analyses were performed [beginning with stepwise linear rank regression
analysis. The results of this analysis were evaluated using classification and regression tree
analysis to determine decision rules that determine whether a particular realization would
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produce doses at the 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.
The uncertainty importance analyses showed that the waste package and saturate 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 analysis, as used in the TSPA-SR, refers 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.
Robustness analysis was conducted by setting a suite of parameters associated with a particular
barrier at their 5th or 95th percentile, 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. ;
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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
j
increase in dose rate over the base case, j
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
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
i
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
in 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
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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 reported by TRWOOa.
Concentration limits. This barrier represents the function of environmental conditions and
radionuclide solubility limits in limiting radionuclide transport out of the EBS. 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. In 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.
• EBS transport. This barrier represents the function of environmental conditions and
diffusion in the drift invert in limiting radionuclide transport out of the EBS. 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 5. . ;
• 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 shoed 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.
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.
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. !
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7.3.10.3 The Yucca Mountain Science and Engineering Report
A more in-depth discussion of the basis for the TSPA-SR was recently provided in the Science
and Engineering Report (DOE01). This basis was used to update information in the Draft EIS, to
reflect the most current information regarding design and safety, and was published recently as a
Supplemental EIS (DOEOla). The information presented in these reports is generally a synopsis
of information published in prior reports, but the format of the report permits a straightforward
comparison between scenarios, models, and parameter assumptions implemented in the TSPA-
SR and observations, elicitations, and measurements that support them. In addition, a series of
discussions are presented on alternative conceptual models omitted from the TSPA-SR, and the
rationale for their omission. These discussions of alternative conceptual models provide useful
contrasts with the TSPA-SR models, and the alternatives are briefly summarized here for each
component of the TSP A: ;
[
• For unsaturated zone flow, alternative conceptual models are discussed for lateral flow in
' the Ptn unit, fracture flow in the Ptn unit, episodic flow in the Tsw unit, low permeability
of faults in the Ptn and Chn/CFu units, ana for discrete fracture flow. Each of these
conceptual models was argued to be equivalent to, or less conservative than, the
conceptual model used in the TSPA-SR. The data needed to resolve between the TSPA-
SR conceptual model and these alternative's were argued to be either currently unavailable
or impractical to gather. i
• Alternative thermal-hydrological conceptual models were stated to be in one of several
categories: alternative representation of fractured rock in numerical models, alternative
selection of representative property values, and the potential for permanent changes in
those properties from the effects of heating. Specific attention is drawn to alternative
thermal-hydrologic-chemical and thermal-hydrologic-mechanical approaches. No
arguments are made as to the relative conservatism of the alternative approaches and the
TSPA-SR model.
• Alternative conceptual models for the physical and chemical environment of the
repository were described as alternative concepts for the thermal-hydrologic-chemical
seepage model, for the approach to representing precipitates and salts contacting the drip
shield and waste package, for evaluating rrucrobial communities in the repository, for
interactions of steel and titanium, and for modeling rockfall. It was argued by DOE01
that these alternative approaches would provide results comparable to those in the TSPA-
SR.
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Alternative conceptual models for waste-package and drip-shield degradation were
discussed for oxidation, localized corrosion thresholds, stress corrosion cracking, stress
mitigation, and hydrogen-induced cracking. The TSPA-SR model for stress corrosion
cracking is described as conservative compared to alternatives, as it is based on data for
stainless steel, which is more prone to such cracking than Alloy 22. Estimates of the
relative conservatism of the other conceptual models are not given by DOE01.
Alternative conceptual models associated with flow diversion are discussed by DOE01 in
the representation of seepage inflow as a function of time and location, in the
environment under the drip shield, and in water drainage from the drifts. It is stated that
the TSPA-SR model is expected to be similar in average behavior to the alternative
concepts for seepage inflow. The environment under the drip shield in the TSPA-SR is
conservatively assumed to permit microbially induced corrosion, which may not occur in
alternative conceptual models. The alternatives associated with drainage of water from
the drifts have not been fully evaluated, but they are argued to be unimportant.
DOEO1 states that alternative conceptual models for waste form degradation and
radionuclide release were considered in all aspects of the TSPA-SR model., Specific
alternatives are discussed for evaluation of inventory, in-package chemistry, cladding
degradation, dissolution of spent fuel (both commercial and DOE), glass degradation,
solubility, and colloid generation.
DOE01 states that alternative conceptual models for transport in the engineered barrier
system are improbable or not supported by data. It is argued that the use of a multi-
dimensional model would lead to small differences in performance.
Alternative conceptual models associated with unsaturated-zone transport below the
repository are argued to be similar to those for unsaturated zone flow above the
repository. In addition, the potential for a drift shadow of drier conditions is mentioned,
which would be less conservative than the TSPA-SR model. In addition, a discussion is
provided of matrix diffusion. It is argued that neglect of matrix diffusion is not realistic,
and that such a model should not be used.
For the saturated-zone system, DOE01 identifies four key areas of uncertainty for which
alternative models are possible: treatment as a porous medium or if discrete features need
to be included, behavior in the vicinity of large hydraulic gradient, recharge and the time
scale over which it functions, and scale dependence of data and model parameters. No
arguments are made by DOE01 as to the relative conservatism of the TSPA-SR model
and alternatives in this area.
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In addition to the discussion of alternative conceptual models for the nominal scenario, DOE01
provided in-depth discussions of alternative scenarios that might be considered in a TSPA, but
which have been discarded based on either probability of occurrence, reasonableness, or lack of
consequences. Scenarios associated with water table rise and nuclear criticality are discussed in
detail, with extensive presentations of the determination that neither is credible.
7.3.11 Uncertainties in Projecting Repository Performance Over Very Long Time Periods
Repository performance assessments, such as those for a potential repository at Yucca Mountain
discussed in Sections 7.3.1 through 7.3.9, are the means to assess potential doses to individuals
as a result of radionuclide releases from the repository. Potential doses are, in turn, the major
factor in regulatory compliance evaluations.
A 10,000-year time horizon for repository dose projections is well established in regulatory
approaches in the United States (40 CFR Part 191, Part 194, and Part 197; 10 CFR Part 60 and
draft Part 63) and in other nations (GAO94). Performance projections have'a fundamental role in
implementing these regulations and in regulatory decision making, and the reliability of
performance projections over such time frames fojr estimating potential doses to exposed
individuals and groups is inextricably tied to geologic stability issues for any particular repository
site. This section discusses uncertainties over the time period for assessing the performance of a
repository at Yucca Mountain, and the effect of the uncertainties on regulatory decision making.
As discussed in Sections 7.3.1 through 7.3.9, a series of Total System Performance Assessments
(TSPA) has been reported by DOE during the past decade, and additional TSPA evaluations will
be done for the License Application if the Yucca Mountain site is recommended to be suitable
for disposal (Section 7.3.10). Some TSPA evaluations have been performed for time periods up
to one million years because some of the radionuclides that could be released from the repository
to the environment have half lives that would enable potential doses to humans to occur for such
time periods (e.g., Np-237 and Pu-242). I
Results of the TSPA evaluations show that potential doses increase over long time frames, i.e.,
beyond 10,000 years. The TSPA-SR analyses (TRWOOb) showed, for example, that, for the
EDA II repository design and TSPA-SR models, pjeak doses would continue to increase over
100,000 years after disposal, but would not reach very high values even at that time.
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