United States
Environmental Protection
Agency
Air And Radiation
(6602J)
!PA 402-R-96-002
January 1996
Criteria For The Certification And
Re-Certification Of The Waste
Isolation Pilot Plant's Compliance
With The 40 CFR Part 191
Disposal Regulations
Background Information
Document For 40 CFR Part 194
Recycled/Recyclable
~ pnnted with Soy/Canola Ink oropaper that
contains at least 50% recyd«ltiber
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40 CFR Part 194
BACKGROUND INFORMATION DOCUMENT
FOR THE CRITERIA FOR THE CERTIFICATION
AND RE-CERTIFICATION OF THE
WASTE ISOLATION PILOT PLANT'S
COMPLIANCE WITH THE
40 CFR PART 191 DISPOSAL REGULATIONS
January 1996
U.S. Environmental Protection Agency
Office of Radiation and Indoor Air
Washington, D.C. 20460
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Contents
1. Introduction 1-1
1.1 EPA Authorities for the Rulemaking 1-1
1.1.1 Purpose of 40 CFR Part 194 1-1
1.1.2 Overview of 40 CFR Part 194 1-3
1.1.3 Statutory and Regulatory Basis 1-4
1.1.4 Compliance with Other Environmental Laws and Regulations . . 1-4
1.2 History of U.S. Policy on Geological Disposal 1-5
1.2.1 EPA's Development of the Generally Applicable 40 CFR part 191 1-7
1.3 Purpose and Scope of the Background Information Document 1-11
1.4 References 1-13
2. Quality Assurance Program (§194.22) 2-1
2.1 Introduction 2-1
2.2 DOE Quality Assurance Program 2-3
2.2.1 Management and Oversight 2-3
2.2.2 Key Regulatory Issue . 2-7
2.2.3 Key DOE Quality Assurance Documents . ... 2-7
2.2.3.1 DOE Order 5700.6C-Quality Assurance 2-7
2.2.3.2 Quality Assurance Requirements and Description
Document 2-9
2.2.3.3 Carlsbad Area Office Quality Assurance
Program Description 2-10
- 2.2.3.4 TRU Waste Characterization Quality Assurance
Program Plan (TRU QAPP) 2-13
2.2.4 Key WIPP Scientific Advisor Documents 2-13
2.2.5 Management and Operating Contractor Documents 2-14
2.2.6 TRU Waste Generator Sites 2-14
2.2.7 Qualification of Existing Data 2-14
2.2.7.1 Background 2-14
2.2.7.2 SNL Program For Qualifying Existing Data 2-15
2.2.7.3 SNL Program for Qualifying Existing Data 2-16
2.2.7.4 Current Status 2-17
2.2.8 Quality Assurance for Models and Codes 2-17
2.3 Summary of U.S. NRC Requirements 2-18
2.3.1 Data Gathering 2-20
2.3.2 Data Analysis 2-21
2.3.3 Data Modeling 2-21
2.3.4 Management and Oversight 2-23
2.3.5 Qualification of Existing Data 2-24
2.4 ASME NQA-1, NQA-2, AND NQA-3 Standards 2-24
2.4.1 ASME NQA-1 2-25
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Contents (Cont'd)
Page
2.4.2 ASME NQA-2, part 2.7 2-25
2.4.3 ASME NQA-3 2-25
2.5 References . . 2-26
3. Issues for the Selection and Development of Models and Computer Codes (§194.23) 3-1
3.1 Introduction .... 3-1
3.2 Review of Computer Code by Evolution Evaluation ... 3-2
3.2.1 Conceptual Models ... .3-3
3.2.2 Mathematical Models .... .... . . 3-6
3.2.3 Numerical Models ... 3-8
3.2.4 Computer Code 3-9
3.3 Code-Related Issues 3-10
3.3.1 Source Code Availability 3-11
3.3.2 History of Use 3-11
3.3.3 Quality Assurance 3-11
3.3.3.1 Code Documentation ... . . 3-13
3.3.3.2 Code Testing 3-16
3.3.4 Hardware Requirements . . . 3-17
3.3.5 Mathematical Solution Methodology 3-18
3.3.6 Code Dimensionality 3-18
3.4 Model Application 3-20
3.5 References 3-23
4. Waste Characterization (§194.24) .... 4-1
4.1 Introduction 4-1
4.1.1 Brief History of DOE's TRU Waste Characterization Program .... 4-1
4.1.1.1 DOE Wipp Waste Acceptance Criteria (WAC) 4-2
4.1.1.2 WIPP (TRU QAPP) 4-6
4.2 Regulatory Drivers for Waste Characterization 4-7
4.2.1 Agreement for Consultation and Cooperation July 1, 1981 . 4-8
4.2.2 WTPP LWA 4-9
4.2.3 NRC Regulations for the Packaging and Transportation of
Radioactive Waste (10 CFR part 71) 4-11
4.2.4 U.S. Department of Transportation Regulations:
49 CFR part 173 4-14
4.2.5 Resource Conservation and Recovery Act (RCRA) 4-14
4.2.5.1 RCRA part B Permit Application 4-16
4.2.6 Federal Facilities Compliance Act of 1992 4-18
4.2.7 (TSCA): 40 CFR part 761 PCB Manufacturing, Processing, Distribution
in Commerce and Use Prohibitions 4-21
11
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Contents (Cont'd)
Page
4.2.8 Environment Protection Standards for Management and
Disposal of Spent Nuclear Fuel, High-Level and Transuranic
Radioactive Wastes (40 CFR part 191) 4-22
4.2.9 Criteria for the Certification and Re-Certificationof the Waste
Isolation Pilot Plant's Compliance with the 40 CFR Part 191
Disposal Regulations (40 CFR Part 194) 4-23
4.3 Impacts of Waste Characteristics and Components on Performance
Assessment 4-23
4.3.1 Waste Characteristics 4-24
4.3.1.1 Mobility in Solution 4-24
4.3.1.2 Wave Strength 4-24
4.3.1.3 Gas Generation Within the Waste 4-25
4.3.1.4 Fluid Flow 4-27
4.3.2 Waste Components 4-27
4.3.3 Current and Projected Waste Inventory at the WIPP 4-28
4.3.4 Identification of Significant Radionuclides 4-33
4.3.5 Determination of Actinide Solubility Limits 4-34
4.3.6 Determination of Gas Generation Rates 4-36
4.3.6.1 Average Stoichiometry Model 4-38
4.3.6.2 The Reaction Path Model 4-41
4.3.7 Establishing the Waste Envelope 4-41
4.4 Methods for Characterizing WIPP Waste Inventory 4-42
4.4.1 Radioassay 4-47
4.4.2 Radiography 4-48
4.4.3 Headspace Sampling and Analysis 4-50
4.4.4 Solid Process Residues and Soils Sampling and Analysis 4-51
4.4.5 Visual Examination 4-51
4.4.6 Use of Acceptable Knowledge/Process Knowledge 4-52
4.5 Use of Process Knowledge (Acceptable Knowledge) to Characterize
TRU Wastes 4-52
4.5.1 Definition and Regulatory Precedent For the Use of Process
Knowledge (Acceptable Knowledge) 4-52
4.5.2 Using Process Knowledge for Waste Characterization 4-55
4.5.3 Use of Acceptable Knowledge/Process Knowledge for TRU
Inventory 4-56
4.5.4 Evaluating the Use of Process Knowledge 4-57
4.6 Technical Rationale For Waste Characterization Provision of
40 CFR part 194 4-59
4.6.1 General Information on Waste 4-59
4.6.2 Documentation of Waste Characteristics 4-60
4.6.3 Documentation of Waste Components 4-61
in
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Contents (Cont'd)
4.6.4 Limits on Waste Components 4-63
4.6.5 Quality Assurance 4-64
4.7 References 4-64
5. Future State Assumptions (§194.25) . . . 5-1
5.1 Introduction .... . . . . . . 5-1
5.2 Land Use Around the WIPP Site 5-2
5.3 Future States of Climate . . .5-8
5.4 Geologic Future States 5-11
5.5 Hydrologic Future States . . 5-14
5.6 References 5-15
6. Use of Expert Judgment (§194.26) 6-1
6.1 Introduction 6-1
6.1.1 Background . 6-1
6.1.2 NRC Publications on the Expert Judgment Process . . . . 6-3
6.2 Examples of the Use of Expert Judgment at Facilities other than the WIPP 6-5
6.2.1 Nuclear Regulatory Commission 6-5
6.2.1.1 NUREG/CR-1150 . . 6-5
6.2.1.2 Yucca Mountain Climate Study 6-8
6.2.2 Electric Power Research Institute 6-9
6.2.3 Other Countries 6-11
6.2.3.1 United Kingdom 6-11
6.2.3.2 European Space Agency . ... 6-13
6.3 Use in WIPP Parameter Assumptions .... . 6-14
7. Peer Review Procedures (§194.27) 7-1
7.1 Introduction „ 7-1
7.1.1 Background 7-1
7.2 Other Peer Review Programs 7-2
7.2.1 Definition and Use of Peer Review 7-2
7.2.2 Peer Review at the Department of Health and Human Services ... .7-4
7.2.2.1 Public Health Service . . 7-4
7.2.2.2 National Institutes of Health 7-4
7.2.3 Peer Review at the National Aeronautics and Space Administration . . 7-6
7.2.4 Peer Review at the U.S. Environmental Protection Agency 7-9
7.2.4.1 Peer Review of Proposed Sewage Sludge
Disposal Regulations ... 7-10
7.2.4.2 Ecological Risk Assessment Peer Review 7-12
7.2.4.3 The Science Advisory Board .... 7-13
7.2.5 Nuclear Regulatory Commission Peer Review Guidance . . 7-14
IV
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Contents (Cont'd)
Page
7.2.6 Peer Review at the Department of Energy 7-16
7.2.6.1 Methodology 7-17
7.2.6.2 Peer Review at the Yucca Mountain Site 7-17
7.3 Summary of Peer Review 7-20
7.4 References 7-21
8. Uncertainty and "Reasonable Expectation" (§194.34) 8-1
8.1. Introduction 8-1
8.1.1 Background 8-1
8.1.2 General Approach to Evaluating Compliance 8-2
8.1.3 Outline of Chapter 8 8-5
8.2 Probability of Compliance 8-6
8.2.1 Review of the Probabilistic Requirements of 40 CFR part 191 8-6
8.2.2 Statistical Interpretation of the Requirements of 40 CFR part 191 ... 8-8
8.2.3 Use of Expert and Peer Review for Determining
Level of Confidence . . 8-9
8.2.4 Use of Statistical Methods for Determining Compliance ... . 8-10
8.2.5 Use of Sampling Methods 8-14
8.2.6 Conditional Probabilities of Compliance 8-16
8.2.7 Unconditional Probability of Compliance with the Containment
Requirements 8-19
8.3 Comparison of Alternative Criteria for Compliance 8-20
8.3.1 Advantages and Disadvantages of Alternative Compliance Criteria
Using a Central Point Measure 8-20
8.3.2 Advantages and Disadvantages of Alternative Compliance Criteria
Using Percentiles 8-26
8.4 Other Regulatory Considerations 8-28
8.4.1 Environmental Protection Agency 8-28
8.4.1.1 40 CFR Part 268, Land Disposal Restrictions 8-28
8.4.1.2 40 CFR Part 148, Hazardous Waste Injection Restrictions
(Underground Injection Control) 8-29
8.4.2 Nuclear Regulatory Commission 8-30
8.4.3 Department of Energy 8-32
8.4.4 Non-U.S. Disposal Systems 8-32
8.4.4.1 Organization for Economic Co-operation and Development
(OECD) Nuclear Energy Agency 8-34
8.4.4.2 Canada 8-34
8.4.4.3 France : . 8-35
8.4.4.4 Sweden 8-37
8.4.4.5 Switzerland 8-37
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Contents (Cont'd)
Page
8.5 Conclusions 8-38
8.5.1 Essential Role of Peer Review 8-38
8.5.2 Selection of a Statistical Criterion for Compliance 8-39
8.6 References 8-41
9. Consideration of Human Intrusion (§194.32 and §194.33) .. .9-1
9.1 Introduction 9-1
9.2 Geologic Setting 9-2
9.3 Intrusion by Drilling 9-6
9.3.1 Oil and Gas Drilling 9-6
9.3.1.1 Permitting Practices 9-8
9.3.1.2 Well Drilling and Casing ..9-9
9.3.1.3 Detection of the Repository During Drilling . . ... 9-10
9.3.1.4 Borehole Plugging and Abandonment 9-12
9.3.1.5 Human Intruction Scenarios . . 9-17
9.3.2 Exploratory Drilling for Potash . . 9-17
9.3.3 Water Well Drilling ... . 9-20
9.3.4 Other Exploratory Drilling . . 9-22
9.4 Intrusion by Mining . . . . 9-23
9.4.1 Introduction 9-23
9.4.2 Mining Scenarios 9-30
9.4.3 Literature Review ..... 9-32
9.4.3.1 WIPP Related Studies . . 9-32
9.4.3.2 Other Relevant Studies .' . 9-36
9.4.4 Impact of Mining on Hydraulic Conductivity 9-38
9.4.4.1 Background Information 9-38
9.4.4.2 Strain Analysis 9-45
9.4.5 Consideration of Other Mining Impacts 9-60
9.4.5.1 Solution Mining 9-60
9.4.5.2 Change in Flow Direction of Water-Bearing Members
if a Vertical Hydraulic Connection if Created by
Subsidence 9-63
9.4.5.3 Formation of Subsidence-Related Surface Depressions
where Water could Accumulate and Alter Local Recharge
Characteristics . 9-63
9.4.5.4 Increased Hydraulic Gradient if Significant Flow from
Water-Bearing Strata into Mine Workings Occurs ..... 9-63
9.4.5.5 Damage to Borehole or Shaft Seals by Subsidence . . . 9-68
9.4.5.6 Increased Hydraulic Conductivity of the Salado
Formation due to Excavation-Induced Stress . . . . 9-68
VI
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Contents (Cont'd)
9.4.6 Summary 9-69
9.5 References 9-70
10. Active Institutional Controls (§194.41) 10-1
10.1 Introduction 10-1
10.1.1 Regulatory Requirements Relevant to Institutional
Controls at WIPP 10-1
10.1.2 WIPP Site Characteristics 10-3
10.1.2.1 Site Description 10-3
10.1.2.2 WIPP Facilities 10-3
10.1.2.3 Waste Characteristics 10-3
10.1.2.4 Operations 10-4
10.1.2.5 Closure/Post-Closure Activities 10-4
10.2 Active Institutional Controls Proposed for the WIPP Site 10-4
10.3 Institutional Controls at Other Facilities 10-7
10.3.1 Facilities Containing Special Nuclear Material 10-7
10.3.2 Retired Nuclear Reactor Facilities ... 10-7
10.3.3 Low-Level Radioactive Waste Disposal Facilities 10-8
10.3.4 Uranium Mill Tailings Disposal 10-10
10.3.5 Superfund Sites 10-11
10.3.6 Applicability to WIPP 10-12
10.4 Adequacy of Proposed Active Institutional Controls for WIPP 10-12
10.4.1 Inadvertent Intrusion Scenarios 10-13
10.4.1.1 Recreational 10-13
10.4.1.2 Agricultural 10-14
10.4.1.3 Home Construction 10-15
10.4.1.4 Groundwater Scenario 10-15
10.4.1.5 Drilling for Hydrocarbons 10-16
10.4.2 Intentional Intrusion Scenarios 10-17
10.4.2.1 Benign Intentional Intrusion Scenarios 10-17
10.4.2.2 Hostile Intentional Intrusion Scenarios 10-17
10.5 References 10-18
11. Monitoring (§194.42) 11-1
11.1 The Need for Monitoring 11-1
11.1.1 EPA Disposal Standards 11-1
11.1.2 RCRA Regulations 11-3
11.2 Pre-Closure Monitoring 11-4
11.2.1 Pre-Disposal Monitoring to Support Operations and Closure 11-4
11.2.2 Pre-Closure Monitoring to Support Assurance Requirements .. .. 11-5
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Contents (Cont'd)
Page
11.2.2.1 Brine Quantity, Flux Composition, and Spatial
Disposition .... 11-5
11.3 Post Disposal Monitoring 11-14
11.3.1 Jeopardizing Waste Isolation .... 11-14
11.3.1.1 Introduction . ... ... 11-14
11.3.1.2 Ramifications of Non-Threatening Monitoring . . 11-15
11.3.1.3 Bore Hole Techniques 11-16
11.3.2 Geophysical Methods 11-17
11.3.2.1 Surface Subsidence 11-17
11.3.2.2 Direct Repository Monitoring 11-18
11.3.2.3 Groundwater Monitoring 11-18
11.3.2.4 Electromagnetic Conductivity 11-20
11.3.2.5 Resistivity 11-21
11.3.2.6 Seismic Techniques 11-22
11.3.2.7 Gravitation ... . 11-23
11.3.2.8 Magnetization . 11-23
11.4 Conclusion . 11-24
11.5 References . . 11-24
12. Passive Institutional Controls (§194.43) 12-1
12.1 Introduction 12-1
12.1.1 Regulatory Background 12-1
12.1.2 General Background 12-3
12.2 Permanent Markers . 12-6
12.2.1 Archeological Analogues . . . . . 12-7
12.2.1.1 Introduction 12-7
12.2.1.2 Monuments of Mesopotamia 12-10
12.2.1.3 Intaglios of the California, Arizona, and Nevada
Deserts 12-13
12.2.1.4 Australian Rock Art 12-14
12.2.1.5 Rock Art in the Western United States (General) .... 12-16
12.2.1.6 Summary Observations on Archeological
Analogues 12-18
12.2.1.7 The WIPP Markers Panel Report 12-20
12.2.2 NRC Studies . 12-22
12.2.3 NASA Studies 12-22
12.2.4 NEA/OECD Studies . ........ 12-23
12.2.5 DOE Studies 12-24
12.2.5.1 ONWI Program 12-24
12.2.5.2 WIPP Program 12-27
12.2.5.3 Project PLOWSHARE and Related Tests 12-33
Vlll
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Contents (Cont'd)
12.2.6 No Marker Strategy 12-37
12.2.6.1 Futures Panel 12-39
12.2.6.2 Markers Panel 12-40
12.2.6.3 ANDRA Off-Site Marker Concept 12-42
12.3 Public Records and Archives 12-44
12.3.1 Regulations 12-45
12.3.1.1 WIPP Land Withdrawal Act (LWA) 12-45
12.3.1.2 40 CFR part 194 12-45
12.3.2 Historical Perspective on Use and Survivability of Records 12-46
12.3.3 Survivability of Land Ownership Records 12-48
12.3.3.1 Introduction 12-48
12.3.3.2 Historical Land Records 12-49
12.3.3.3 State Land Records in the United States 12-50
12.3.3.4 Public Land Records 12-51
12.3.3.5 Past Problems with Public Land Records 12-52
12.3.3.6 Automation of Public Land Records 12-57
12.3.4 Contemporary Examples of Lost Government Records 12-56
12.3.4.1 Oil and Gas Leases Near WIPP 12-57
12.3.4.2 Lost AEC Records 12-59
12.3.4.3 Spring Valley Munitions Dump 12-61
12.3.5 Format for Records 12-62
12.4 Government Ownership and Regulations 12-63
12.4.1 General Comments 12-63
12.4.2 WIPP Land Withdrawal Act 12-64
12.5 Other Methods of Preserving Disposal System Knowledge 12-66
12.5.1 Subsurface Markers 12-66
12.5.1.1 Passive Markers 12-66
12.5.1.2 Buried Sensors 12-66
12.5.2 Protective Barriers 12-67
12.6 References 12-68
Appendix 12A: Federal Register Notice Identifying WIPP
Land Withdrawal Area 12A-1
Appendix 12B: Letter to U.S. Archivist Transmitting WIPP Land Withdrawal
Information 12B-1
13. Engineered Barriers (§194.44) 13-1
13.1 Regulatory Background 13-1
13.1.1 Environmental Protection Agency Regulations 13-1
13.1.2 Nuclear Regulatory Commission Regulations 13-2
13.2 Consideration of Engineered Alternatives 13-3
IX
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Contents (Cont'd)
13.2.1 Engineered Alternatives Task Force 13-3
13.2.2 Engineered Barriers Study for §194.44 13-8
13.2.3 Waste Inventory 13-9
13.3 References . . 13-12
14. Individual and Ground Water Protection Requirements (§194.51-194.55) ... . 14-1
14.1 Introduction . 14-1
14.2 Ground-Water Protection 14-1
14.3 Individual Protection 14-3
14.3.1 Consideration of the Exposure Pathways 14-4
14.3.2 Location of Protected Individual 14-5
14.3.3 Calculation of Radiation Dose 14-5
14.3.4 EPA's Standardized Exposure Scenarios and Default Exposure
Parameter values for Human Health Risk Assignment 14-5
14.3.5 Exposure Scenarios Considered by DOE and the NRC 14-12
14.4 Scope of Compliance Assessments . . . 14-12
14.5 Results of Compliance Assessments (§194.55) 14-14
14.6 References 14-16
List Of Abbreviations xv
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Figures
2-1 Current WIPP Organizational Chart 2-4
2-2 Hierarchy of DOE WIPP QA Program Documents 2-8
2-3 DOE QAPD Source Documents by Category 2-11
4-1 Generalized Sequence for TRU Waste Characterization 4-45
4-2 Programmatic Approach to Visual Characterization of TRU Waste 4-49
5-1 Land ownership within 30 Miles of the WIPP Site 5-5
5-2 WIPP Land Withdrawal Area and Surroundings 5-6
5-3 Land Use Within 30 Miles of the WIPP Site 5-7
5-4 Estimated Mean Annual Precipitation Rate at the WIPP During the Late
Pleistocene and Holocene (SWI93) 5-10
5-5 Boundary Conditions for Regional Domain used to model
"Climate Change" 5-16
6-1 Principal Steps in NUREG/CR-1150 Expert Elicitation Process 6-7
8-1 Ten Hypothetical LHS CCDFs, with Maximum 8-17
8-2 Mean, 90th Percentile, and Median Curves from Set of 10 LHS CCDFs . . . 8-17
9-1 Stratigraphic Column and Potential Hydrocarbon Pay Zones
(Source: POW78) 9-4
9-2 Outline of the Delaware Basin (Source: POW78, Figure 6.3-8) 9-7
9-1 Generalized Stratigraphic Column of Permian and Younger Strata,
Eddy County, New Mexico (CHE78) 9-25
9-4 Location of BLM Lease Grade Mineralization Within the WIPP Site 9-26
9-5 Finite Elements Mesh Used for Strain Analysis
Mesh 4,500 ft by 6,750 ft 9-49
9-6 Half Width (1,500 ft) of Mined Panel 9-49
Subsidence-induced Culebra strains for subsidence factor of 52.5% 9-52
9-8 Aperture Change in Vertical Joints for Fracture Spacings of 3, 30,
and 300 inches and Subsidence Factor of 52.5% 9-54
9-9 Aperture Change in Horizontal Joints for Fracture Spacings of 3, 30,
and 300 inches and a Subsidence Factor of 52.5% 9-54
9-10 Relative Change hi Fracture Hydraulic Conductivity for Vertical Joints
with various Fracture Spacings, Subsidence Factor of 52.5%, and
Fracture Apertures of 10^, 10"3, and 10"2in.
a) 3 inch b) 30 inch c) 300 inch 9-55
XI
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Figures (Cont'd)
9-11 Relative Change in Fracture Hydraulic Conductivity for Horizontal Joints
with various Fracture Spacings, Subsidence Factor of 52.5%,
and Fracture Apertures of 10"*, 10"3, and 10"2in
a) 3 inch b) 30 inch c) 300 inch 9-57
9-12 Subsidence-induced Culebra Strains for Subsidence Factor of 67.5% . . . 9-58
9-13 Subsidence-induced Culebra Strains for Subsidence Factor of 67.5%
Horizontal Mesh Extended to 13,500 ft 9-58
9-14 Aperture Change in Vertical Joints for Fracture Spacings of 3, 30, and
300 inches and Subsidence Factor of 67.5% 9-59
9-15 Aperture Change in Horizontal Joints for Fracture Spacings of 3, 30, and
300 inches and a Subsidence Factor of 67.5% . . . 9-59
9-16 Relative Change in fracture Hydraulic Conductivity for Vertical Joints with
various Fracture Spacings, Subsidence Factor of 67.5%,
and Fracture Apertures of 10"4, 10~3, and 10"2in.
a) 3 inch b) 30 inch c) 300 inch 9-61
9-17 Relative Change in Fracture Hydraulic Conductivity for Horizontal Joints
with various Fracture Spacings, Subsidence Factor of 67.5%,
and Fracture Apertures of 10^, 10"3, and 10"2in.
a) 3 inch b) 30 inch c) 300 inch 9-62
9-18 The Ratio of the Hydraulic Gradient Imposed by Mining to the
Ambient Gradient . ... 9-66
12-1 Human Interference Logic Flow ... 12-5
12-1 ONWI Conceptual Design of Site Markers for Geologic Repository .... 12-26
xn
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Tables
4-1 Summary of Waste Acceptance Criteria and Requirements 4-3
4-2 Transuranic Waste Disposal Inventory for WIPP 4-29
4-3 Estimated Composition of Waste Disposal Inventory at WIPP Repository
Capacity 4-30
4-4 Waste Matrix Code Group Names 4-31
4-5 WIPP CH-TRU Waste Material Parameter Disposal Inventory 4-32
4-6 Major Nuclides in Disposal Radionuclide Inventory 4-33
4-7 Waste Characterization Capabilities of Ten Main TRU Waste Generators . . 4-47
4-8 Classes and Examples of Acceptable Knowledge 4-56
4-9 Summary of Waste Characteristics and Waste Components Likely to be
Used in WIPP Performance Assessment 4-63
5-1 1979 Resident Population Within 50 Miles of the Site 5-3
5-2 Comparison of EPA, DOE, and NRC Intake Rates and
Exposure Assumptions 5-18
8-1 Measures of the Central Point 8-21
8-2 Measures of Spread or Dispersion 8-23
8-3 Numerical Measures of Compliance 8-24
8-4 Regulatory Status and Approach to Safety in Foreign Geologic
Repositories 8-33
9-1 Potash Reserves and Resources within WIPP Site Boundary (GRI95) .... 9-27
9-2 Active Potash Mines in New Mexico Showing Estimated Capacity,
Average Ore Grade, and Mine Life at the Average 1992
Price of $81.14/st product 9-29
9-3 Summary of IT Corp. Subsidence Prediction Results for WIPP
Repository (ITC94) 9-35
9-4 Rock Properties by Type 9-46
9-5 Strata Depth and Thickness 9-47
9-6 Summary of Results for Mining Scenario 9-65
9-7 Ground water Velocities at Hydraulic Conductives the Range from 2-1000 times
those Values Presented in Table 9-6 9-67
13-1 Potentially Useful Engineered Alternatives Considered By the Engineered
Alternatives Multidisciplinary Panel (EAMP)(From DOE91a) 13-5
13-2 Engineered Alternatives Evaluated by the EATF Relative to the Baseline
Case (From DOE91a) 13-6
13-3 Transuranic Waste Disposal Inventory for WIPP (Volumes in Cubic
Meters) (From BIR95) 13-10
14-1 Comparison of EPA, DOE, and NRC Intake Rates and Exposure Assumptions 14-7
Xlll
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List of Abbreviations
AEA Atomic Energy Act
AEC Atomic Energy Commission
AECB Atomic Energy Control Board
ALMRS Automated Land and Minerals Records System
ANDRA Agence National pour la Gestion des Dechets Radioatifs
ANL-E Argonne National Laboratory-East
ANL-W Argonne National Laboratory-West
ANPR advance notice of proposed rulemaking
AO Announcement of Opportunity
ASME American Society of Mechanical Engineers
ASTP Actinide Source Term Program
BID Background Information Document
BLM Bureau of Land Management
BSEP Brine Sampling and Evaluation Program
CAO Carlsbad Area Office
CCDF complementary cumulative distribution functions
CERCLA Comprehensive Environmental Response, Compensation, and Liability Act
CFR Code of Federal Regulations
CH-TRU contact-handled transuranic waste
CRW Committee on Radioactive Waste Management
CSRS Cooperative States Research Service
DHHS Department of Health and Human Services
DOE U.S. Department of Energy
DOT U.S. Department of Transportation
DQO data quality objective
DSIN Directorate for the Safety of Nuclear Installations
EATF Engineered Alternatives Task Force
EAMP Engineered Alternatives Multidisciplinary Panel
EEG Environmental Evaluation Group
EM electromagnetic
EPA U.S. Environmental Protection Agency
EPRI Electric Power Research Institute
ERDA Energy Research and Development Administration
ERP Environmental Restoration Project
ESA European Space Agency
FACA Federal Advisory Committee Act
FEIS Final Environmental Impact Statement
FERC Federal Energy Resources Council
FFCA Federal Facility Compliance Act
FLPMA Federal Land Policy and Management Act
GCDB geographic coordinate data base
GTP Generator Treatment Plan
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List of Abbreviations (Cont'd)
HANF Hanford Site
HITF Human Interference Task Force
HLW high-level radioactive waste
HSWA Hazardous and Solid Waste Amendments
ICPMS inductively coupled plasma / mass spectrometry
ICRP International Commission on Radiological Protection
INEL Idaho National Engineering Laboratory
IQR interquartile range
IRG Interagency Review Group
IWPF Idaho Waste Processing Facility
KPLA Known Potash Leashing Area
LANL Los Alamos National Laboratories
LHS Latin hypercube sampling
LLNL Lawrence Livermore National Laboratory
LLW low-level radioactive waste
LMR Life-of-Mine potash Reserves
LWA Land Withdrawal Act
MAD mean absolute deviation
MCL maximum contaminant level
MOU Memorandum of Understanding
MOUND Mound Plant
MRE measure of relative effectiveness
MWIR Mixed Waste Inventory Report
NAC National Advisory Council
NACEPT National Advisory Council on Environmental Policy
NAGRA National Cooperative for the Storage of Radioactive Waste
NAS National Academy of Sciences
NAS-NRC National Academy of Sciences National Research Council
NASA National Aeronautics and Space Administration
NCP National Contingency Plan
NEA Nuclear Energy Agency
NICS Nuclear Safety Research
NIH National Institutes of Health
NMED New Mexico Environment Department
NNWSI Nevada Nuclear Waste Storage Investigations
NPL National Priorities List
NPRM notice of proposed rule making
NRA NASA Research Announcement
NRC U.S. Nuclear Regulatory Commission
NSF National Science Foundation
NTPO National TRU Program Office
NTS Nevada Test Site
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List of Abbreviations (Cont'd)
NWPA Nuclear Waste Policy Act
NWTS National Waste Terminal
NYDEC New York Department of Ecology
OCD Oil Conservation Division
OECD Organization for Economic Co-operation and Development
OMB Office of Management and Budget
ONWI Office of Nuclear Waste Isolation
OPA Office of Program Analysis
OPvNL Oak Ridge National Laboratory
PA performance assessment
PAN passive active neutron
PCB polychlorinated biphenyl
PDP performance demonstration program
PHS Public Health Service
PI Petroleum Information
PLSS Public Land Survey System
PNL Pacific Northwest Laboratory
PRC peer review committee
QA quality assurance
QAO quality assurance objective
QAP quality assurance program
QAPD Quality Assurance Program Description
QAPP Quality Assurance Program Plan
QAPjP Quality Assurance Project Plan
QARD Quality Assurance Requirements & Description document
QMT Qualification Methods Team
RCRA Resource Conservation and Recovery Act
RF radio frequency
RFETS Rocky Flats Environmental Technology Site
REP Rocky Flats Plant
RH-TRU remote-handled transuranic waste
RME Reasonable Maximum Exposure
RMWTF Remote Mixed Waste Treatment Facility
RTR real time radiography
RYW Rhodes-Yates Waterflood Area
SAB Science Advisory Board
SAFSTOR safe storage
SAR Safety Analysis Report
SDWA Safe Drinking Water Act
SGS segmented gamma scan
SKB Swedish Nuclear Fuel and Waste Management Company
SKI Swedish Nuclear Power Inspectorate
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List of Abbreviations (Cont'd)
SNL Sandia National Laboratories
SNR signal to noise ratio
SOG Seismicity Owners Group
SQA Software Quality Assurance
SRS Savannah River Site
STP Source Team Panel
STTP Source Term Test Program
SWB Standard Waste Boxes
SWCF Sandia W1PP Central File
SWDA Solid Waste Disposal Act
TASC The Analytical science Corporation
TDS total dissolved solids
TRU transuranic
TRU QAPD Transuanic Waste Characterization Quality Assurance Program Plan
TRUCON TRUMPACT-II Content Codes
TRUMP AC TRUMPACT-II Authorized Methods for Payload Control
TSCA Toxic Substances Control Act
UIC underground injection control
ULF ultra low frequency
UMTRCA Uranium Mill Tailings Radiation Control Act
USD A U.S. Department of Agriculture
USDW underground source of drinking water
USGS U.S. Geological Survey
VLF very low frequency
VOC volatile organic compound
WAC Waste Acceptance Criteria
WACCC Waste Acceptance Criteria Certification Committee
WHPP Waste Handling and Packaging Plant
WID Waste Isolation Division
WIPP Waste Isolation Pilot Plant
WMC waste matrix code
WMCG waste matrix code groups
WTAC WIPP Technical Support Contractor
WTWBIR WIPP Transuranic Waste Baseline Inventory Report
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1. Introduction
The Environmental Protection Agency's (EPA) regulation, 40 CFR part 194, sets forth
criteria for determining if the Waste Isolation Pilot Plant (WIPP) will comply with EPA's
environmental radiation protection standards for the disposal of radioactive waste, found at
40 CFR part 191 subparts B and C. If the Administrator of EPA determines that the WIPP
will comply with the standards for disposal, then the Administrator will issue to the Secretary
of Energy a certification of compliance which will allow the emplacement of transuranic
waste in the WIPP to begin, provided that all other statutory requirements have been met. If
a certification is issued, EPA will also use 40 CFR part 194 to determine if the WIPP has
remained in compliance with EPA's environmental radiation protection standards, once every
five years after the initial receipt of waste for disposal at the WIPP. The final preamble and
regulation to 40 CFR part 194, as they appear in the Federal Register, take precedence over
any descriptions or interpretations of the final rule that appear in this document.
This document provides much of the necessary background information and technical
analyses which the Agency used during the development of 40 CFR part 194. The document
explicates fourteen issues considered by EPA in establishing the individual criteria contained
in 40 CFR part 194.
1.1 EPA'S REGULATORY OVERSIGHT OF THE WIPP
1.1.1 Purpose of 40 CFR Part 194
The criteria for compliance, 40 CFR part 194, implement the Environmental Protection
Agency's (EPA) environmental radiation protection standards, 40 CFR part 191, by applying
them to the proposed disposal of transuranic radioactive waste in the Waste Isolation Pilot
Plant (WIPP). The EPA previously promulgated 40 CFR part 191, "Environmental
Radiation Protection Standards for Management and Disposal of Spent Nuclear Fuel, High-
Level and Transuranic Radioactive Wastes," to provide standards that will apply to all sites
(except Yucca Mountain) for the deep geologic disposal of highly radioactive waste.
Complete descriptions of 40 CFR part 191 were published hi the Federal Register in 1985
(50 Fed. Reg. 38066-38089, Sep. 19, 1985) and 1993 (58 Fed. Reg. 66398 66416, Dec.
20, 1993). The WIPP is subject to 40 CFR part 191, and is being constructed by the
Department of Energy (DOE) near Carlsbad, New Mexico as a potential repository for the
safe disposal of transuranic radioactive waste. The EPA is required by the WIPP Land
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Withdrawal Act of 1992 (Pub. L. 102-579) to evaluate whether the WIPP will comply with
subparts B and C of 40 CFR part 191 — known as the "disposal regulations" — and to issue
or deny a certification of compliance. The Department of Energy is required to submit an
application to EPA that will be the basis of EPA's evaluation of whether a certification of the
WIPP's compliance with the disposal regulations should be issued. The Department of
Energy may not begin to emplace transuranic waste underground for disposal at the WIPP
until such time as a certification of compliance has been issued and all other requirements of
section 7(b) of the WIPP Land Withdrawal Act have been satisfied. With 40 CFR part 194,
the Agency establishes criteria by which to judge whether the WIPP is in compliance with
the "disposal regulations" and sets forth procedural requirements for this determination.
The criteria for compliance, 40 CFR part 194, also apply to the periodic re-certification of
the WIPP's compliance with the disposal regulations. The process of periodic re-
certification, established by section 8(f) of the WIPP Land Withdrawal Act, calls for EPA to
determine whether the WIPP continues to be in compliance with the disposal regulations,
assuming that an initial certification of compliance has been issued. The Secretary of Energy
must submit to the Administrator of EPA documentation of the WIPP's continued compliance
with the disposal regulations, every five years after the initial receipt of transuranic waste for
disposal at the WIPP, until the end of the decommissioning phase. The Agency will use the
criteria in determining whether or not the WIPP will have continued to be in compliance.
The WIPP was authorized in 1980 under section 213 of the Department of Energy National
Security and Military Applications of the Nuclear Energy Authorization Act of 1980 (Pub. L.
96-164, 93 Stat. 1259, 1265), "for the express purpose of providing a research and
development facility to demonstrate the safe disposal of radioactive wastes resulting from the
defense activities and programs of the United States." The waste proposed for disposal in
the WIPP, transuranic radioactive waste (TRU waste), is waste consisting of materials such
as rags, equipment, tools, protective gear and sludges which have become contaminated
during atomic energy defense activities. The WIPP Land Withdrawal Act defines transuranic
waste to be waste containing more than 100 nano-curies per gram of alpha-emitting radio-
isotopes, with half-lives greater than twenty years and atomic number greater than 92, per
gram of waste. The Act further stipulates that radioactive waste shall not be transuranic
waste if such waste also meets the definition of high-level radioactive waste, has been
specifically exempted from the disposal regulations with the concurrence of the
Administrator, or has been approved for an alternate method of disposal by the Nuclear
Regulatory Commission. The radioactive component of transuranic waste consists of man-
made elements created during the process of nuclear fission, chiefly isotopes of plutonium.
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1.1.2 Overview of 40 CFR part 194
The regulation, 40 CFR part 194, sets forth the criteria against which the WIPP's compliance
with the disposal regulations of 40 CFR part 191 will be evaluated and is divided into four
subparts, consisting of:
1) Subpart A, which specifies the protocols for submission of certification applications, the
terms of any certification, and the process for any subsequent suspension, modification, or
revocation of compliance status.
2) Subpart B, which outlines the information to be included with compliance applications to
ensure that EPA has adequate information to evaluate the basis for any demonstration of
compliance. Subsequent applications for continued compliance must note any changes in
such information that have occurred since the previous certification.
3) Subpart C, which implements the specific containment, assurance, individual, and
groundwater protection requirements of the disposal regulations at 40 CFR part 191.
General requirements, such as those for quality assurance and waste characterization, are
included to ensure that compliance applications are based on reliable information; they also
allow EPA inspection authority to confirm conditions reported in applications. Assessments
of disposal system performance are expressed to show the likelihood of release or exposure
occurring. Performance assessments for releases must account for the frequency and
consequences of potential human intrusion into the repository over the 10,000-year regulatory
time frame, as specified by 40 CFR part 194. Assurance requirements, designed to increase
confidence in the performance of the disposal system, include criteria for monitoring of
repository performance, and implementation of engineered barriers to protect against releases
from the disposal system.
4) Subpart D, which provides opportunities for public participation in the rulemaking
processes for initial certification of compliance and for modification or revocation of any
certification. It also provides for public input at critical junctures hi the re-certification
process. The subpart specifies criteria for notification of the public at each stage of
rulemakings, holding of public hearings, opportunity for public comment, and creation and
maintenance of public dockets hi Washington, DC, and New Mexico.
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1.1.3 Statutory and Regulatory Basis
40 CFR part 194 was mandated by Congress in section 8(c) of the WIPP Land Withdrawal
Act. The criteria promulgated in 40 CFR part 194 implement only those subparts of 40 CFR
part 191 that apply to the disposal of transuranic radioactive waste. 40 CFR part 194 does
not amend 40 CFR part 191. Subpart A of 40 CFR part 191 applies to the management of
spent nuclear fuel, high-level and transuranic radioactive wastes at sites designated for the
disposal of these wastes and is not the subject of 40 CFR part 194. However, section 9(a) of
the WIPP Land Withdrawal Act stipulates that the Secretary of Energy shall comply with
respect to the WIPP with Subpart A of 40 CFR part 191. With the Energy Policy Act of
1992, Congress mandated the development of regulations to replace 40 CFR part 191 for the
Yucca Mountain site only, but the entire standard, 40 CFR part 191, remains applicable to
the WIPP. See 106 Stat. 2921, section 801(a)(l). The entire 40 CFR part 191 standard was
developed to establish generally applicable standards for the protection of the general
environment from radioactive materials, specifically those disposed of in mined geologic
repositories. The standard was developed pursuant to the Agency's authorities under the
Atomic Energy Act (AEA) of 1954, as amended, and Reorganization Plan No. 3 of 1970
(NIX70). A more complete description of the development of 40 CFR part 191 may be
found later in this chapter.
1.1.4 Compliance with Other Environmental Laws and Regulations
The WIPP is regulated under the Resource Conservation and Recovery Act (RCRA) and is
subject to both the Part B licensing requirements and the land disposal restrictions of that
statute. The WIPP must comply with other environmental laws, including, among other
statutes, the Clean Air Act (40 U.S.C. 7401 et sea), the Toxic Substances Control Act (15
U.S.C. 2601 et seq) and the Comprehensive Environmental Response, Compensation, and
Liability Act of 1980 (42 U.S.C. 9601 et seq). The promulgation of 40 CFR part 194 does
not affect the need for DOE to comply with these and all other applicable environmental laws
with respect to the WIPP.
Much of the waste proposed for disposal in the WIPP is mixed waste, i.e., waste composed
of both radioactive and hazardous constituents, the latter's disposal being controlled by the
regulations set forth under RCRA. As mandated by section 7(a) of the WIPP Land
Withdrawal Act, the Secretary of Energy must obtain from the Administrator a determination
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of no-migration under the land disposal restrictions of RCRA, prior to commencing the
emplacement of waste in the WIPP. A conditional no-migration determination for the now
withdrawn test phase was granted by EPA on November 14, 1990. This conditional
determination must be amended and formal approval granted before disposal of radioactive
waste can begin.
1.2 HISTORY OF U.S. POLICY ON GEOLOGIC DISPOSAL
Since the 1940s, the Federal Government has assumed ultimate responsibility for the care and
disposal of high-level radioactive wastes, regardless of whether they are produced by
commercial or national defense activities. To respond to this need, in 1949 the Atomic
Energy Commission (AEC) initiated research and development work on the conversion of
high-level radioactive liquid wastes into a stable, solid form. Then, hi 1955, at the request
of the AEC, a National Academy of Sciences - National Research Council (NAS-NRC)
Advisory Committee was established to consider the disposal of high-level radioactive wastes
within the United States. Its report (NAS57), issued in 1957, recommended the following:
1. The AEC continue to develop processes for the solidification of high-
level radioactive liquid wastes, and
2. Naturally occurring salt formations are the most promising medium for
the long-term isolation of these solidified wastes.
Project Salt Vault, conducted from 1965 to 1967 by the AEC in an abandoned salt mine near
Lyons, Kansas, was initiated to demonstrate the safety and feasibility of handling and storing
solid wastes in salt formations (MCC70).
In 1968, the AEC again asked the NAS-NRC to establish a Committee on Radioactive Waste
Management (CRWM) to advise the AEC concerning its long-range radioactive waste
management plans and to evaluate the feasibility of disposing of solidified radioactive wastes
in bedded salt. The CRWM convened a panel to discuss the disposal of radioactive wastes in
salt mines. Based on the recommendations of the panel, the CRWM concluded that bedded
salt is satisfactory for the disposal of radioactive wastes (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
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years, however, in-depth site studies raised several questions concerning the safe plugging of
old exploratory wells and on proposals for expanded salt mining activities. These questions
and growing public opposition to the Lyons site prompted the AEC in late 1971 to pursue
alternatives (DOU72).
The Federal Government intensified its program to develop and demonstrate a permanent
disposal method for high-level radioactive wastes and the Office of Management and Budget
(OMB) established an interagency task force on commercial wastes in March, 1976. The
OMB interagency task force defined the responsibility of each Federal agency involved in
high-level waste management, including the preparation of environmental standards for high-
level wastes by EPA (LYN76, ENG77a, ENG77b).
A status report on the management of commercial radioactive nuclear wastes, published in
May 1976 by the President's Federal Energy Resources Council (FERC), emphasized the
need for coordination of administration policies and programs relating to energy. The FERC
established a nuclear subcommittee to coordinate Federal nuclear policy and programs to
assure an integrated government effort. This report called for an accelerated, comprehensive
government radioactive waste program plan and recommended the formation of an
interagency task force to coordinate activities among the responsible Federal agencies. EPA
was given the responsibility for establishing general environmental standards governing waste
disposal activities (FER76).
In 1976, President Ford issued a major policy statement on nuclear waste. As part of his
comprehensive statement, he announced new steps to assure that the United States would
have facilities for the long-term management of nuclear waste from commercial power
plants. The President's actions were based on the findings of the OMB interagency task
force formed in March 1976. He announced that the experts had concluded that the most
practical method for disposing of high-level radioactive wastes is in geologic repositories
located in stable formations deep underground. EPA's responsibilities were better defined to
include issuing general environmental standards governing nuclear waste facility releases 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.
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In December 1976, EPA announced its intent to develop environmental radiation protection
criteria for radioactive wastes to assure the protection of public health and the general
environment (EPA76). These efforts resulted hi a series of radioactive waste disposal
workshops, held in 1977 and 1978 (EPA77a, EPA77b, EPA78a, EPA78b).
In 1978, President Carter established the Interagency Review Group (IRG) to recommend an
administrative policy for addressing the long-term management of nuclear waste. The IRG
was to recommend programs that would support the policy when adopted. The IRG report
re-emphasized EPA's role in developing generally applicable standards for the disposal of
high-level wastes, spent nuclear fuel, and transuranic wastes (DOE79). In a message to
Congress on February 12, 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 high-level and transuranic wastes that
would rely on mined geologic repositories. The message repeated that EPA was responsible
for creating general criteria and numerical standards for nuclear waste management activities
(CAR80).
1.2.1 EPA's Development of the Generally Applicable 40 CFR part 191
In November 1978, EPA published proposed "Criteria for Radioactive Wastes," which were
intended as Federal Guidance for storage and disposal of all forms of radioactive wastes
(EPA78c). In March 1981, however, EPA withdrew the proposed criteria because the many
different types of radioactive wastes made the issuance of generic disposal guidance too
problematic (EPA81).
In 1982, under the authority of the Atomic Energy Act of 1954, 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).
Shortly after the publication of EPA's proposed rule, Congress passed the Nuclear Waste
Policy Act of 1982, wherein EPA was to "...promulgate generally applicable standards for
the protection of the general environment from off-site releases from radioactive material hi
repositories..." not later than January 1984 (NWP83).
After the first comment period on the proposed rule ended on May 2, 1983, EPA held two
public hearings on the proposed standards-one in Washington, D.C., on May 12-14, 1983,
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and one in Denver, CO, on May 19-21, 1983-and during a second public comment period
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 (EPA85a).
In parallel with its public review and comment effort, the Agency conducted an independent
scientific review of the technical basis 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 18, 1983, through September 21, 1983, and later
prepared and released a final report on February 17, 1984 (EPA83c, SAB84). The SAB
review found that the Agency's analyses in support of the proposed standards were
comprehensive and scientifically competent, but contained several recommendations for
improvement. The report was publicly released on May 8, 1984, and the public was
requested to comment on the findings and recommendations (EPA84). Public responses to
the SAB report were subsequently presented and released in August 1985 (EPA85b).
On February 8, 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 7, 1984, deadline mandated by the NWPA for promulgation of the
standards. A consent order was negotiated with the plaintiffs that required the standards to
be promulgated on or before August 15, 1985. EPA issued the final rule under 40 CFR part
191 on that date (EPA85c, EPA85d, EPA85e).
EPA standards were divided into two main sections, Subparts A and B. Subpart A addressed
the management and storage of waste. For any disposal facility operated by DOE and not
regulated by the Nuclear Regulatory Commission (NRC) or by Agreement States, under
Subpart A of the standard, the exposure limits to any member of the general public were 25
millirem (mrem) to the whole body and 75 mrem to any critical organ. For facilities
regulated by the Nuclear Regulatory Commission or Agreement States, the standards adopt
the annual dose limits given in 40 CFR part 190, the environmental standards for the
uranium fuel cycle: 25 mrem to the whole body, 75 mrem to the thyroid, and 25 mrem to
the critical organ.
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Subpart B .imposed limits on 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 over 10,000 years;
• Assurance requirements to compensate for uncertainties in achieving the
desired level of protection;
• Individual exposure limits based on the consumption of groundwater
and any other potential exposure pathways for 1,000 years after
disposal; and
• Groundwater protection requirements in terms of allowable radionuclide
concentrations and associated doses for 1,000 years after disposal.
§191.15 and §191.16 of Subpart B limited the annual dose to any member of the general
public to 25 mrem to the whole body and 75 mrem to any critical organ. The groundwater
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. For radium-226
and radium-228 alone, the concentration limit was 5 picocuries/liter. Appendix A of the
standards provided cumulative release limits for other radionuclides.
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 hi the U.S. Court of Appeals for the First Circuit in Boston. The
main challenges concerned:
1. Violation of the Safe Drinking Water Act (SDWA) underground
injection requirements;
2. Inadequate notice and comment opportunity on the groundwater
protection requirements; and
3. 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 hi the remand of the standards. The
Court began by looking at the definition of "underground injection," which is the "subsurface
emplacement of fluids by well injection." A "well" is defined by the SDWA and EPA as a
shaft "bored, drilled, or driven where the depth is greater than the largest surface
dimension." A "fluid" is a material or substance that flows or moves whether in a semi-
solid, sludge, gas, or any other form or state." In the view of the Court, the method
envisioned by DOE for disposal of radioactive wastes in underground repositories might fit
both of the latter definitions and 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. A
dichotomy appeared when endangerment was considered inside the "controlled area" versus
beyond the controlled area (i.e., in the accessible environment). Inside the controlled area,
the Court ruled that endangerment of groundwater was permitted. Therefore, EPA's
approach of using the geological formation as part of the containment was valid. However,
outside the controlled area where endangerment would not be permitted, the Court found that
§191.15 as promulgated would endanger drinking water supplies. In the context of the
SDWA, "endangerment" occurs when doses are higher than that allowed by the Primary
Drinking Water Regulations. §191.15 permits an annual dose of 25 mrem to the whole body
and 75 mrem to any critical organ from all pathways. On the other hand, the regulations
under the SDWA allow only 4 mrem doses from drinking water. The Court recognized that
less than 4 mrem may result from the groundwater pathway; however, it rejected this
possibility because the Agency stated that radioactivity may eventually be released into the
groundwater system near the repository which could result hi substantially higher doses.
Therefore, the Court decided that a large fraction of the 25 mrem could be received through
the groundwater exposure pathway. Accordingly, the Court found that the high-level waste
standards should have been consistent with the SDWA, or the Agency should have explained
that a different standard was adopted and justified its position.
The Court also noted that the Agency was not incorrect in promulgating the proposed
standards, but that the Agency neither acknowledged the interrelationship of the SDWA and
HLW rules, nor did it adequately explain the divergence between them. The Court also
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supported the petitioner's argument that the Agency arbitrarily selected the 1,000-year limit
for individual protection requirements (§191.15) under undisturbed performance. The Court
indicated that the 1,000-year criterion is not inherently flawed, but the administrative record
and the Agency's explanations did not adequately support this choice. The criterion was
remanded for reconsideration and a more thorough explanation for its basis. Finally, the
Court found that the Agency did not provide adequate opportunity for notice and comments
on §191.16 (Groundwater Protection Requirements), which was added to Subpart B after the
standards were proposed. This section was remanded for a second notice and comment
opportunity.
In August 1987, the Justice Department petitioned the First Circuit Court to reinstate all of
40 CFR part 191 except for §191.15 and §191.16, which were originally found defective.
The Natural Resources Defense Council filed an opposing opinion. In response, the Court
issued an Amended Decree that reinstated Subpart A, but continued the remand of Subpart B.
On October 30, 1992, the President signed the WIPP LWA. This Act reinstated Subpart B
of 40 CFR part 191, except §191.15 and §191.16, and required the Administrator to issue
final disposal standards. The reinstatement of these regulations does not apply to the
characterization, licensing, construction, operation, or closure of any site required to be
characterized under the NWPA Section 113(a) of Public Law 97-425. On December 20,
1993, EPA issued amendments to 40 CFR part 191 which: eliminated §191.16 of the
original rule; altered the individual protection requirements; and added Subpart C on
groundwater protection. The amended standards represent the Agency's response to the
above legislation and to the issues raised by the court pertaining to individual and
groundwater requirements. EPA did not revisit any of the regulations reinstated by the
WIPP LWA.
1.3 PURPOSE AND SCOPE OF THE BACKGROUND INFORMATION DOCUMENT
This Background Information Document (BID) provides much of the necessary background
information and technical analyses which support the Agency's development of 40 CFR part
194. The BID explicates fourteen issues considered by EPA in establishing the individual
criteria contained in 40 CFR part 194. For clarity of presentation, the issues generally have
been arranged to correspond to their relative placement in 40 CFR part 194. Following are
brief descriptions of the remaining chapters:
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GENERAL REQUIREMENTS
Chapter 2 An assessment of the DOE Quality Assurance (QA) program as it
relates to site characterization, data gathering, data analysis, and data modeling
at the WIPP. DOE, EPA, NRC, and other QA guidance are examined.
Chapter 3 A discussion of the use of appropriate models in the WIPP
performance assessment.
Chapter 4 A review of the DOE TRU waste characterization program.
Chapter 5 A review of background information and technical analyses
relevant to future state assumptions.
Chapter 6 A discussion of the formal use of expert judgment in scientific
investigation and how the technique has been applied at the WIPP.
Chapter 7 A review of peer review procedures and a discussion of their
application in the WIPP assessments.
CONTAINMENT REQUIREMENTS
Chapter 8 A discussion of background information on evaluation of
uncertainty, and a summary of regulatory approaches for dealing with
uncertainty, including "reasonable expectation."
Chapter 9 A discussion of resource drilling and mining.
ASSURANCE REQUIREMENTS
Chapter 10 A discussion of regulatory requirements relevant to active
institutional controls at the WIPP and DOE proposed action.
Chapter 11 A review of issues relevant to monitoring, including the necessity
for monitoring and potential techniques for pre- and post-disposal monitoring.
Chapter 12 A discussion on the use of passive institutional controls, including
permanent markers, public records and archives, and government ownership
and regulations.
Chapter 13 A review of the regulations concerning engineered barriers and
consideration of engineered barriers at the WIPP
Chapter 14 A discussion on the development of compliance criteria for
individual and groundwater protection requirements.
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1.4 REFERENCES
AEC70 Atomic Energy Commission Press Release No. N-102, dated June 17, 1970.
CAR80 The White House, President J. Carter, "The President's Program on
Radioactive Waste Management," Fact Sheet, February 12, 1980.
DOE79 Department of Energy, "Report to the President by the Interagency Review
Group on Nuclear Waste Management," Report No. TID-29442, March 1979.
DOU72 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.
ENG77a 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.
ENG77b 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.
EPA76 Environmental Protection Agency, "Environmental Protection Standards for
High-Level Wastes - Advance Notice of Proposed Rulemaking," Federal
Register, 41 FR 53363, December 6, 1976.
EPA77a 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.
EPA77b 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.
EPA78a Environmental Protection Agency, "Background Report - Consideration of
Environmental Protection Criteria for Radioactive Wastes," Office of Radiation
Programs, Washington, D.C., February 1978.
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EPA78b 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 Environmental Protection Agency, "Recommendations for Federal Guidance,
Criteria for Radioactive Wastes," Federal Register, 43 FR 53262-53268,
November 15, 1978.
EPA81 Environmental Protection Agency, "Withdrawal of Proposed Regulations,"
Federal Register, 46 FR 17567, March 19, 1981.
EPA82 Environmental Protection Agency, "Proposed Rule, Environmental Standards
for the Management and Disposal of Spent Nuclear Fuel, High-Level and
Transuranic Radioactive Wastes," Federal Register, 47 FR 58196-58206,
December 29, 1982.
EPA83a 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 Environmental Protection Agency, "Environmental Standards for the
Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic
Radioactive Wastes, Request for Post-Hearing Comments," Federal Register,
48 FR 23666, May 26, 1983.
EPA83c Environmental Protection Agency, Science Advisory Board Open Meeting:
High-Level Radioactive Waste Disposal Subcommittee, Federal Register,
48 FR 509, January 5, 1983.
EPA84 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, May 8, 1984.
EPA85a 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.
EPA85b 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.
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EPA85c 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.
EPA85d 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.
EPA85e Environmental Protection Agency, "40 CFR part 191, Environmental
Standards for the Management and Disposal of Spent Nuclear Fuel, High-
Level and Transuranic Radioactive Wastes; Final Rule," 50 FR 38066-38089.
September 19, 1985. " '
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, DC, 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.
NIX70 The White House, President R. Nixon, "Reorganization Plan No. 3 of 1970,"
Federal Register, 35 FR 15623-15626, October 6, 1970.
NWP83 Nuclear Waste Policy Act of 1982, Public Law 97-425, January 7, 1983.
SAB84 Science Advisory Board, U.S. EPA, "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, Washington, D.C.,
January 1984.
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USC87 United States Court of Appeals for the First Circuit, Natural Resources
Defense Council, Inc., et al., vs United States Environmental Protection
Agency, Docket Nos.: 85-1915, 86-1097, 86-1098, Amended Decree,
September 23, 1987.
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2. Quality Assurance Program
2.1 INTRODUCTION
Quality assurance is the set of planned and systematic actions necessary to provide adequate
confidence that a structure, system, or component will perform satisfactorily in service
(ASM89a). Standard "good practice" may lead to a quality outcome and it is possible that
quality outcomes will result without the imposition of a formal program. However,
formalizing this process helps to assure a quality outcome and the lack of formality can
impede the demonstration of the outcome's inherent quality. Because of the need to provide
confidence that the WIPP will comply with federal and state disposal standards, a carefully
structured quality assurance (QA) program is essential.
DOE's ability to demonstrate compliance with the regulatory requirements of 40 CFR part
191 Subpart B, Environmental Standards for Disposal, and Subpart C, Environmental
Standards for Ground Water Protection, depends in large part on the adequacy of its quality
assurance (QA) program. Demonstration of an appropriately implemented QA program can
provide confidence in the soundness of information and scientific data, thus enabling greater
defensibility for the technical basis of those measures intended to ensure waste isolation.
This is especially true in relation to establishing and maintaining the integrity of data and
models which form the technical basis of the WIPP's performance assessment (PA) process.
In §194.22, EPA has specified criteria aimed at ensuring the soundness of DOE's QA
program for modeling and data collection and analysis. Specific items which the QA
program must address include:
• waste characterization activities and assumptions;
• environmental monitoring, monitoring of the performance of the disposal
system, and sampling and analysis activities;
• field measurements of geologic factors, ground water, meteorologic, and
topographic characteristics;
• computations, computer codes, models, and methods used to demonstrate
compliance with the disposal regulations;
• procedures for implementation of expert judgment elicitations used to support
applications for certification or re-certification of compliance;
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• design of the disposal system and actions taken to ensure compliance with
design specifications;
• collection of data and information used to support compliance applications; and
• other systems, structures, components, and activities important to the
containment of waste in the disposal system.
40 CFR part 191 establishes the disposal system's performance requirements by specifying
criteria for containment, assurance of performance, individual protection, and groundwater
protection, but does not specify requirements for "Quality Assurance." However, §191.13(b)
requires a "reasonable expectation" that compliance with the requirements will be achieved
based upon the total record before the implementing agency. This statement implicitly
requires a mechanism to (1) produce such a record, and (2) to provide a basis for that record
to support the concept of "reasonable expectation." Quality assurance is an integral element
in the formalization of this mechanism. A fully implemented quality assurance program that
is in compliance with the appropriate requirements justifies a high level of confidence in the
scientific protocols and data which form the basis for waste isolation estimates.
To ensure that calculations of compliance with 40 CFR part 191 are based on sound data and
information, EPA requires in 40 CFR part 194 that DOE implement a QA program that
meets the requirements of the following documents:
• American Society of Mechanical Engineers' (ASME) "Quality Assurance Program
Requirements for Nuclear Facilities" (NQA-1-1989 edition) (ASM89a);
• ASME's "Quality Assurance Requirements of Computer Software for Nuclear Facility
Applications" (NQA-2a-1990 addenda (part 2.7) to ASME NQA-2-1989 edition)
(ASM89b); and
• ASME's "Quality Assurance Program Requirements for the Collection of Scientific
and Technical Information for Site Characterization of High-Level Nuclear Waste
Repositories" (NQA-3-1989 edition excluding Section 2.1(b) and (c)) (ASM89c).
The ASME national consensus standards are well established within the U.S. nuclear
industry. They have a long history of use and provide extensive supplemental guidance.
EPA believes the use of these standards offers the most comprehensive/credible and specific
set of QA requirements for all compliance-related elements of the disposal system. For
example:
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• NQA-1 sets forth requirements for the "establishment and execution of quality
assurance programs for the siting, design, construction, operation, and
decommissioning of nuclear facilities."
• NQA-2 (part 2.7) establishes requirements for "the development, procurement,
maintenance, and use of computer software, as applied to the design, construction,
operation, modification, repair, and maintenance of nuclear facilities;" and applies to
computer software "used to produce or manipulate data which is used directly in the
design, analysis, and operation of structures, systems, and components."
• NQA-3 sets forth quality assurance requirements for "the collection of scientific and
technical information for site characterization of high-level nuclear waste
repositories;" and applies to "activities which could affect the quality of scientific
and technical information collected as part of the site characterization phase of high-
level nuclear waste repositories...[which include] as a minimum: (a) readiness
reviews; (b) peer reviews; (c) data and sample management; (d) data collection and
analysis; (e) coring; (f) sampling; (g) in situ testing; and (h) scientific investigations."
This chapter describes the current DOE quality assurance program for the WIPP, describes
how NRC addresses similar regulatory requirements, and provides additional background on
the basis for selecting the ASME NQA requirements for 40 CFR part 194.
2.2 DOE QUALITY ASSURANCE PROGRAM
2.2.1 Management and Oversight
DOE Office of Environmental Management
Within the DOE headquarters organization, the Office of Environmental Management (EM-1)
is responsible for the overall management of DOE waste management programs. Under EM-
1, responsibilities of the Office of Waste Management (EM-30) include programmatic
management of site operations for storage, treatment, or disposal of radioactive, hazardous
and mixed waste materials including defense-generated TRU waste. Additionally, EM-30 is
responsible for assuring that waste is properly characterized, packaged, labeled, and
transported to the WIPP hi accordance with DOE priorities and objectives; and providing
management direction to the waste generators.
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to
Assistant Secretary
Office of Environment
Restoration and Waste
Management (EM-1)
1
Manager
DOE Carlsbad
Area Office
(DOE/CAO)
1
Deputy Assistant
Secretary,
DOE Office of
Waste Management
(EM-30)
CAO Chief,
Compliance
Branch
CAO Chief,
WIPP Site
Branch
Manager,
Westinghouse
Electric Corporation/
Waste Isolation
Division
1
CAO Chief,
Experimental
Programs
Branch
CAO Chief,
Administrative
Branch
Department
Manager,
Sandia National
Laboratories
(SNL)
Manager,
National TRU
Program Office
(NTPO)
Manager,
NTPO Waste
Characterization
and Technology
Program
|
Manager,
NTPO Program
Assessment and
Certification
DOE Field
Office
DOE Site
Project
Manager
Figure 2-1. DOE WIPP Organizational Chart
DOE Site
Project QA
Officer
-------
Carlsbad Area Office
The WIPP management structure contained many organizational levels among DOE
headquarters and field activities until early 1994 when DOE streamlined the organizational
structure. This streamlining resulted in the vesting of major responsibilities for WIPP in the
Carlsbad Area Office (CAO)1 (DOE93a, DOE93b). The current WIPP organization is
illustrated in Figure 2-1 (DOE94).
The mission of the CAO is to integrate the national transuranic (TRU) waste generator
activities and carry out the actions necessary to facilitate DOE's decision to operate the
WIPP as a disposal facility (REF). Overall responsibility for the development and
implementation of the CAO quality assurance program for all WIPP related activities resides
with the CAO Manager who reports directly to EM-1 as shown in Figure 2-1. The activities
under CAO can be assigned to three main areas, as listed below:
• WIPP site activities are performed by the WIPP Site Management and Operating
Contractor, Westinghouse Waste Isolation Division (WID) located at the WIPP Site
outside of Carlsbad, NM. WID fulfills the requirements of a contract managed by
CAO under the direction of the CAO Manager and is responsible for WIPP site
operation (including support of experiments) and maintenance and for monitoring the
site environment.
• WIPP experimental programs are conducted under the direction of the WIPP
Scientific Advisor, Sandia National Laboratories (SNL) located hi Albuquerque, NM.
SNL fulfills the requirements of a contract managed by DOE Albuquerque Operations
and overseen by the CAO Manager. In this capacity, SNL is responsible for
developing, confirming, and validating models used to simulate long-term disposal
system performance (i.e., performance assessment); and conducting research,
experiments, and tests to collect the data needed for input to the models (DOE92).
SNL sets forth its QA requirements through its Sandia National Laboratories Waste
Isolation Pilot Plant Quality Assurance Program Description and implementing
procedures (SNL95).
• TRU Waste Generator Site activities at the TRU waste generator sites are managed by
DOE Field and Site Offices, as discussed below. A detailed description of site
1 For example, in 1991, DOE EM-1 provided policy guidance and centralized management through EM-30 to
DOE Albuquerque Operations, which reported to the WIPP Project Integration Office. Policy guidance and
management now flow directly from DOE EM-1 to the Manager of the DOE Carlsbad Area Office. DOE EM-30
still ensures that program plans and operations are coordinated, integrated, and consistent with DOE Headquarters.
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activities is beyond the scope of this report. The TRU waste generator sites are
responsible for TRU and mixed TRU waste characterization and for the waste
certification programs. The National TRU Program Office (NTPO) Team Leader is
responsible for the day-to-day implementation of DOE Headquarters policy and
technical direction. The NTPO Team Leader is also responsible for overseeing waste
characterization activities and for providing an interface between DOE field offices
and CAO. NTPO is divided into two functional areas, described below:
• Waste Characterization and Technology is responsible for the development,
issuance and distribution of technical documents that control the TRU Waste
Characterization Program.
• Assessment and Certification is responsible for the verification of compliance
with the TRU Waste Characterization Program requirements at participating
TRU waste generator sites through audits (DOE94).
The CAO QA Manager has the overall responsibility to independently assess the effective
implementation of the QA program. Other responsibilities of the QA Manager include:
• Interfacing with CAO technical staff on quality related matters;
• Maintaining a liaison with the QA organizations of WIPP participants and other
affected organizations; and
• Review and approval of CAO procedures and contractor quality assurance program
descriptions.
77? U Waste Generator Sites
Each of the DOE sites that currently generate, process or store TRU wastes intended for
disposal at WIPP must comply with applicable federal and state regulations regarding waste
characterization, storage, transportation, etc. While these sites vary considerably in size,
complexity and function, each site prepares a site specific Quality Assurance Project Plan
(QAPjP). The QAPjP translates the applicable CAO and other federal and state regulations
into procedures for that site. As stated previously, the daily operations at TRU generator
sites are managed by DOE field and site offices with guidance provided by NTPO. CAO
assesses generator site activities through Quality Assurance audits and surveillances, focusing
primarily on waste characterization activities.
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2.2.2 Key Regulatory Issue
The Quality Assurance documents described in this section provide requirements for activities
associated with the generation, storage, transport and characterization of TRU waste intended
for disposal at WIPP. However, the ultimate compliance criterion is the determination that
once the appropriate Quality Assurance practices and criteria have been identified that they
are adequately implemented. This must be determined empirically by conducting compliance
audits and surveillances at all levels of operation. DOE is currently in the process of
evaluating the implementation of its QA program, identifying problem areas and preparing a
documentation record of these activities. EPA must evaluate DOE's program by a thorough
evaluation of records in conjunction with selected independent verification.
2.2.3 Key DOE Quality Assurance Documents
DOE has established a hierarchy of quality assurance documents consistent with the
organizational framework (See Figure 2-2). Some of the major documents are described in
this section.
2.2.3.1 DOE Order 5700. 6C Quality Assurance
This document establishes the basic quality assurance framework for the Department
(DOE91) and includes the following:
• Placing responsibility for mission accomplishment and Quality Assurance Program
(QAP) implementation with senior management;
• Training and qualification of all personnel performing assigned work; all important
work will be described in documents and records will be kept;
• Performance of all work to established standards using approved instructions; all
equipment used for data collection shall be calibrated and maintained;
• Verification and validation of the adequacy of all designed products by independent
personnel; and
• Periodic management assessment of the QAP to assure results, and independent
assessments to assess quality; all assessments conducted by personnel technically
qualified and knowledgeable in the areas under assessment.
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(Scientific Advisor)
SNL QAPD
Contractor QA
Plans and
Procedures
CRFs, DOE Orders
NQA-1, -2, -3
(DOE HQ Requirements)
I
EM-QARD
(DOE/CAO Requirements and Guidance)
CAO QAPD
WID QAPD
Contractor QA
Plans and
Procedures
(Management and Operating Contractor) (TRU Waste Generator Sites)
Waste
Characterization
QAPP
Generator QA
Project Plans and
Procedures
Figure 2-2. Hierarchy of DOE WIPP QA Program Documents
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2.2.3.2 Quality Assurance Requirements and Description Document
Based on requirements in Order 5700.6C, DOE EM developed the Quality Assurance
Requirements and Description document (QARD) (DOE91a). The QARD is intended to state
DOE/EM's commitment to specific requirements and to integrate their requirements, for
example, ASME NQA-1, NQA-2 (part 2.7) and NQA-3; and EPA QAMS-005/80. QARD
requirements include the following:
• Organizations shall develop, implement, and maintain a written Quality Assurance
Program (QAP) as identified in DOE Order 5700.6C. Appropriate standards, such as
ASME NQA-1 shall be used, wherever applicable, to develop and implement QAPs.
The QAP Description shall delineate the organizational structure, functional
responsibilities, levels of authority, and interfaces for those managing, performing,
and assessing adequacy of work.
• Personnel shall be trained and qualified to ensure they are capable of performing their
assigned work. Personnel shall be provided continuing training to ensure that job
proficiency is maintained.
• Organizations shall establish and implement processes to detect and prevent quality
problems and to ensure quality improvement. Items and processes that do not meet
established requirements shall be identified, controlled, and corrected. Correction
shall include identifying the causes of problems and preventing recurrence.
• Organizations shall ensure that procured items and services meet established
requirements and perform as specified. Prospective suppliers shall be evaluated and
selected on the basis of specified criteria.
• Inspection and acceptance testing of specified items and processes shall be conducted
using established acceptance and performance criteria. Equipment used for
inspections and tests shall be calibrated and maintained.
• Senior management shall periodically assess the integrated quality assurance program
and its performance. Problems that hinder organizations from achieving their
objectives shall be identified and corrected.
• Planned and periodic independent assessments shall be conducted to measure item
quality and process effectiveness, and to promote improvement.
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2.2.3.3 Carlsbad Area Office Quality Assurance Program Description
The Carlsbad Area Office Quality Assurance Program Description (CAO QAPD) is the
quality management document that identifies the federal and industry quality requirements
applicable to the CAO quality assurance program (DOE 94). The CAO QAPD Revision 1.0
states that compliance to its requirements, responsibilities, and authorities "is mandatory for
CAO personnel" while organizations supporting CAO are expected to use the CAO QAPD
for "guidance." The federal and industry quality program requirement source documents it
identifies are divided into three categories:
• Regulatory documents - these define the requirements necessary for WIPP to receive a
certificate of compliance and operational permits by the federal and state
governments, respectively;
• Commitment documents these have been imposed on WIPP operations by DOE
management; and
• Guidance documents - these provide additional information that may be useful in
developing quality assurance programs for WIPP activities.
A listing of QAPD source documents by category is provided in Figure 2-3. The CAO
QAPD provides a description of general, management, performance, and assessment
requirements, as well as supplementary quality assurance requirements for specific
application areas, such as Scientific Investigation Quality Assurance and Software Quality
Assurance, incorporating the applicable portions of ASME NQA-3 and NQA-2, part 2.7,
respectively. CAO QAPD requirements include the following:
• Identifying the responsibilities and authorities of those organizational line management
positions responsible for achieving and verifying quality.
• Allowing the CAO QA Manager direct access to responsible management at a level
where appropriate action can be effected.
• Performing and documenting planning to ensure work is accomplished under suitably
controlled conditions.
• Establishing and implementing processes to detect and prevent adverse quality
conditions and to ensure quality improvement.
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: REGULATORY REQUIREMENTS
DOCUMENTS
10 CFR part 830
40 CFR part 261
40 CFR part 268.6
10 CFR part 71
40 CFR part 284
COMMITMENT DOCUMENTS
DOE Order 5700.6C
EM-1 QARD
ASME NQA-1 (1989), with all supplements
ASME NQA-2 (1990) Part 2.7
ASME NQA-3 (1989) (with exceptions)
NUREG-1298 (1988)
GUIDANCE .DOCUMENTS
NUREG/BR-0167 (1993)
TITLE
Nuclear Safety Management
Identification and Listing of Hazardous Waste
Land Disposal Restrictions
Subpart H, Quality Assurance, Packaging and
Transportation of Radioactive Material
Standards for Owners and Operators of Hazardous
Waste Treatment Storage and Disposal Facilities
TITLE
Quality Assurance
Quality Assurance Requirements and Description
Quality Assurance Program Requirements for
Nuclear Facilities
Quality Assurance Requirements of Computer
Software for Nuclear Facility^Applications
Quality Assurance Program Requirements for the
Collection of Scientific and Technical Information
for Site Characterization of High-Level Waste
Repositories
Staff Position - Qualification of Existing Data for
High-Level Nuclear Waste Repositories
TITLE
Software Quality Assurance Program and
Guidelines
Figure 2-3. DOE QAPD Source Documents by Category
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• Analyzing performance data that affect quality and identifying lessons learned to
improve items, activities, and processes.
• Preparing, approving, issuing, and controlling documents which prescribe processes,
specify requirements, or establish design.
• Records shall be specified, prepared, reviewed, approved, controlled, and maintained
to accurately reflect completed work and facility conditions and to comply with
statutory or contractual requirements.
• Classifying Quality Assurance records as either "permanent" or "non-permanent."
• Performing work under controlled conditions using approved instructions, procedures,
drawings or other appropriate means.
• Items and processes shall be designed using sound engineering/scientific principles
and appropriate standards. The adequacy of design products shall be verified by
individuals or groups other than those who performed the work.
• Ensuring that procured items and services meet established technical and quality
assurance requirements and that they perform as specified. Prospective suppliers shall
be evaluated and selected on the basis of documented criteria.
• Inspecting and testing specified items and processes, and calibrating and maintaining
equipment used for such tests.
• Conducting planned and periodic assessments to measure management effectiveness,
item quality and process effectiveness, and to promote improvement. Persons
conducting assessments shall be technically qualified and knowledgeable in their
assigned roles.
• Controlling and identifying samples in a manner consistent with their intended use.
• Shall be defining, controlling, verifying and documenting scientific investigations.
• Verification of software shall include reviews that ensure that the requirements are
complete and correct, and shall include the appropriate testing.
EPA Office of Radiation and Indoor Air (ORIA) has conducted two formal reviews of the
CAO QAPD and transmitted comments on Revisions 0 and 1 to DOE CAO. DOE appears
to be revising the QAPD to address EPA concerns.
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2.2.3.4 TRU Waste Characterization Quality Assurance Program Plan (TRU QAPP)
The TRU QAPP presents detailed technical information focusing on analytical techniques for
the collection and analysis of samples at a stated, statistically derived confidence interval for
physical, chemical and radiological parameters. In addition to technical information, the
TRU QAPP provides Quality Assurance information, much of which overlaps with areas
covered in the CAO QAPD (DOE95). The TRU QAPP identifies the quality of data
necessary and the techniques designed to attain and ensure the required quality to meet the
objectives of the WIPP Waste Characterization Program (DOE94), and also contains specific
Quality Assurance Objectives for TRU waste. The waste characterization requirements
presented hi the TRU QAPP focus on the Resource Conservation and Recovery Act (RCRA)
and Department of Transportation (DOT) regulations (DOE94). §194.24 requires DOE to
perform an analysis which identifies waste characteristics influencing waste containment
within the disposal system. Once DOE has completed the analysis, the QAPP should be
revised to address the requirements of 40 CFR part 194. According to CAO, the TRU
QAPP addresses all of the basic requirements of ASME NQA-1 (DOE94); any exceptions to
ASME NQA-1 requirements must be noted in each site's QAPjP. Each TRU generator site's
QAPjP must integrate the TRU QAPP's requirements for all TRU waste intended for
shipment to WIPP, which is accomplished by NTPO/CAO reviewing and approving the
generator site QAPjPs (DOE94).
2.2.4 Key WIPP Scientific Advisor Documents
In their capacity as the WIPP Scientific Advisor, Sandia National Laboratories (SNL) has
developed the SNL WIPP Quality Assurance Program Description (SNL QAPD) (SNL95b)
and an extensive list of implementing procedures to address specific WIPP related Quality
Assurance and technical activities. These activities are generally classified as experimental
programs and they cover a wide range of technical and QA activities from certifying pressure
relief valves to Root Cause Analysis. As of 9-14-95, SNL lists forty one Quality Assurance
Procedures as "Active WIPP Controlled Documents" (REF). EPA needs to determine the
conformance of these SNL documents to the ASME NQA standards and the degree of their
implementation.
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2.2.5 Management and Operating Contractor Documents
Westinghouse Waste Isolation Division (WID) functions in the capacity of the WIPP Site
Management and Operations Contractor. As such, WID is required to comply with all
applicable federal and state regulations. These requirements are integrated in the WID
Quality Assurance Program Description (WID92) and associated implementing procedures
and instructions.
2.2.6 TRU Waste Generator Sites
There are approximately ten major TRU generator sites. Each site is required to develop a
site specific Quality Assurance Project Plan (QAPjP) and supporting procedures. These
documents are the means whereby the site translates CAO and other federal requirements
into operating procedures. The TRU generator site QAPjPs must be approved by CAO
(DOE94).
2.2.7 Qualification of Existing Data
2.2.7.1 Background
An important factor in performance assessment is the use of data that were not generated
under a Quality Assurance Program that complies with the requirements of ASME NQA-1.2
These data are referred to as existing data or old data1 The majority of the early
experimental work performed under and above ground at the WIPP was conducted by the
SNL, beginning in the early 1980's. SNL has only recently begun to configure its Quality
Assurance Program to be consistent with ASME NQA-1, and therefore many of the technical
data generated since the early 1980s in support of various scientific investigations concerning
waste isolation are considered existing data. Under 40 CFR part 194, data that were not
2 In the context of Section 2.2.7, ASME NQA-1 means ASME-NQA-1-1988, ASME NQA-2A-1990, addenda
part 2.7 to ASME NQA-1-1989, and ASME NQA-3-1989 [excluding Section 2.1(b) and (c) in Section 17.1].
"Existing data is data developed prior to the implementation of an NQA-1, -2, -3 QA program by SNL and
its contractors, or data developed outside the SNL-WIPP program, such as by oil companies, national laboratories,
universities, or data published in technical or scientific publications. Existing data does not include information that
is accepted by the scientific and engineering community as established fact (e.g., engineering handbooks, density
tables, gravitational laws, etc.)." (SNL95)
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generated under an ASME NQA-1 quality assurance program must be qualified in a manner
consistent with one of the following four approaches:
• peer review
• confirmatory testing
• use of corroborating data
• a quality assurance program that is equivalent in effect to ASME NQA-1
Additionally, the specific methodology used must be approved by the administrator.
In commenting on EPA's Advance Notice of Proposed Rule-making on compliance criteria,
DOE proposed the following steps to verify data used in compliance assessment (DOE93c):
• The data will be examined against currently approved QA procedures. This
examination will be directed to show that if the data had been obtained under current
QA practices, the results would be equivalent to the ongoing data collection.
• If QA equivalency of the data cannot be shown, and the data are crucial to
compliance demonstration, an independent peer review group will be established to
assess the validity of the data, and DOE will submit the findings to EPA.
• If an acceptable QA level cannot be demonstrated to EPA, and the data are crucial to
compliance, DOE will do statistical resampling to establish the quality of the data or
initiate an activity to reacquire the needed data. However, the original data will not
be discarded. Instead, they will be evaluated for use as confirmation of the newly
acquired data.
2.2.7.2 SNL Program For Qualifying Existing Data
In 1994, SNL began to address the issue of qualifying existing data (QED). The process
initially followed the approach outlined by the US NRC (NRC88a) and consisted of three
main areas:
• Identifying those data necessary for compliance calculations, for settlement of
compliance issues or for submission of DOE's certificate of compliance and
organizing them into groups called Data Records Packages (DRPs);
• Determining whether the selected DRPs are acceptable with respect to technical and
Quality Assurance criteria through the use of a team of qualified, independent
personnel with expertise in the areas of interest, called Independent Review Teams
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(IRT). Each data package is evaluated by an IRT that determines whether the data
were collected under a QA program equivalent to ASME NQA-1, and whether the
Technical/Scientific Protocols employed during data collection are acceptable when
evaluated against a pre-established check list.
• Remediating those data that were judged to be unacceptable by virtue of technical
and/or Quality Assurance flaws by IRT assessment. If a DRP is determined to be
inadequate for technical and/or Quality Assurance reasons by an IRT, alternative
methods for qualifying the data are to be identified by a Qualification Methods Team
(QMT). Such methods include the use of corroborating data, confirmatory testing and
peer review. Data that cannot be suitably qualified must be abandoned.
2.2.7.3 SNL Program for Qualifying Existing Data
In 1994, SNL began to address the issue of qualifying existing data (QED). The process
initially followed the approach outlined by the US NRC (NRC88a) and consisted of three
main areas:
• Identifying those data necessary for compliance calculations, for settlement of
compliance issues or for submission of DOE's certificate of compliance and
organizing them into groups called Data Records Packages (DRPs);
• Determining whether the selected DRPs are acceptable with respect to technical and
quality assurance criteria through the use of a team of qualified, independent
personnel with expertise in the areas of interest, called Independent Review Teams
(IRT). Each data package is evaluated by an IRT that determines whether the data
were collected under a QA program equivalent to ASME NQA-1. and whether the
Technical/Scientific Protocols employed during data collection are acceptable when
evaluated against a pre-established check list.
• Remediating those data that were judged to be unacceptable by virtue of technical
and/or quality assurance flaws by IRT assessment. If a DRP is determined to be
inadequate for technical and/or quality assurance reasons by an IRT, alternative
methods for qualifying the data are to be identified by a Qualification Methods Team
(QMT). Such methods include the use of corroborating data, confirmatory testing and
peer review. Data that cannot be suitably qualified must be abandoned.
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2.2.7.4 Current Status
The assessment of existing data focused on data in four technical areas: natural barriers,
disposal system design and engineered barriers, waste interactions, and human initiated
processes and events. As of August 30, 1995, 46 DRPs had been identified as high priority ,
meaning that it was likely that DOE would use them in whole or part for the submission of
their compliance application. Of these, 26 were assessed as not adequate to support
compliance; 20 were assessed as adequate to support compliance. Of the 26 DRPs assessed
as not adequate, 23 were pending QMT review.
SNL began the QED process by evaluating DRPs in November 1994, and has made
considerable progress to date, as discussed previously. Personnel from EPA Office of
Radiation and Indoor Air-Las Vegas Facility and their technical support contractor have
observed IRT assessments of approximately ten DRPs, some as late as June, 1995.
Additionally, CAO performed a Quality Assurance Audit of the SNL QED process in
September, 1995 (DOE CAO Audit A-95-05). These observations have provided EPA with
insight into SNL's approach to the QED process. The governing document for the QED
process (SNL QAP 20-3) (SNL95a) has been evolving as SNL has progressed through the
QED process. Once SNL's methodology is completed, it would be presented to EPA for
approval.
2.2.8 Quality Assurance for Models and Codes
Sandia National Laboratories is conducting iterative performance assessments to provide
interim guidance prior to preparing a final compliance evaluation (SNL92). These
performance assessments describe the conceptual basis for consequence modeling and
performance assessment methodology, including the selection of scenarios for analysis, the
determination of scenario probabilities, and the estimation of scenario consequences
(SNL92).
The modeling process described in references SNL92 and SNL92a includes significant
participation of peer review groups external to Sandia. The iterative nature of the work leads
to a constant updating of models. If, during review of a compliance application, it is
determined that one or more parts of the model(s) is based upon data obtained in the early
stages of the WIPP program when a less stringent QA program was used, the model's
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validity could become an issue. If found to exist, such issues must be settled using the QA
standards and criteria applicable to the performance assessment process itself.
WIPP procedure No. PAP02, Computer Software Supporting Performance Assessments of
the Waste Isolation Pilot Plant, describes four classes of software:
A- Adjudicated (full QA status)
C- Candidate (partial QA status, possibly undergoing continued refinement)
D- Dormant (obsolete software formerly in Class A or C)
X- Experimental (entry level, software in early stages of development or
experimentation, no QA requirements)
A Software Review Committee decides whether to classify software as Class A. It is SNL
policy to use only Class A software for the performance assessment to support the
application for certification of compliance. This procedure is not as rigorous as that
specified in NQA-2 (part 2.7).
2.3 SUMMARY OF U.S. NRC REQUIREMENTS
The purpose of this section is to identify the NRC QA requirements for data gathering,
analyses, and modeling applicable to high-level radioactive waste disposal systems.4 Useful
parallels may be drawn between NRC and EPA requirements since both involve modeling of
geologic nuclear waste repositories.
Quality assurance requirements for disposal of high-level radioactive wastes in geologic
repositories are specified hi Subpart G of 10 CFR part 60. Subpart G requires DOE to
implement a QA program based on the criteria of Appendix B of 10 CFR part 50 as
applicable5, and which is appropriately supplemented by additional criteria. Specific QA
criteria which the NRC staff use to review the DOE QA program are provided in "Review
Plan for High-Level Waste Repository Quality Assurance Program Descriptions" (NRC89).
This document provides NRC's position on the meaning of the term "as applicable" as used
in Appendix 2B of the disposal system program.
4 References to DOE in this section refer to DOE's Office of Civilian Radioactive Waste Management.
5 10 CFR Part 50 does not address requirements for software or data collection for siting such as are addressed
by ASME NQA part 2.7 and ASME NQA-3.
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The NRC Review Plan endorses ASME NQA-1-1986 and incorporates the lessons learned
from the Ford Study (NRC84), such as the use of technical audits and readiness reviews.
This document also accounts for differences between power reactor projects and the high-
level nuclear waste disposal system program and references the NRC staffs Technical
Positions.
Each section of the Review Plan corresponds to one of the 18 criteria of 10 CFR part 50,
Appendix B, and provides the acceptance criteria the NRC staff uses to evaluate QA program
descriptions or plans. The areas addressed by each of these 18 criteria are listed below:
Criterion 1 Organization
Criterion 2 Quality Assurance Program
Criterion 3 Design Control
Criterion 4 Procurement Document Control
Criterion 5 Instructions, Procedures, and Drawings
Criterion 6 Document Control
Criterion 7 Control of Purchased Material, Equipment, Items and Services
and Software
Criterion 8 Identification and Control of Items, Services, and Software
Criterion 9 Control of Special Processes
Criterion 10 Inspection
Criterion 11 Test Control
Criterion 12 Control of Measuring and Test Equipment
Criterion 13 Handling, Storage, and Shipping
Criterion 14 Inspection, Test, and Operating Status
Criterion 15 Nonconformances
Criterion 16 Corrective Action
Criterion 17 Quality Assurance Records
Criterion 18 Audits
As stated in §60.151, the QA program applies to all systems, structures, and components
"important to safety," to design and characterization of barriers "important to waste
isolation," and to activities related thereto. These activities include site characterization,
facility and component construction, facility operation, performance confirmation, permanent
closure, and decontamination and dismantling of surface facilities.
Section 60.2 defines the term "important to safety" as those "engineered structures, systems,
and components essential to the prevention or mitigation of an accident that could result in a
radiation dose to the whole body, or any organ, of 0.5 rem or greater at .or beyond the
nearest boundary of the unrestricted area at any time until the completion of permanent
closure."
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NUREG-1318, "Technical Position on Items and Activities in the High-Level Waste Geologic
Repository Program Subject to QA Requirements" (NRC88b) provides guidance for the
identification of items important to safety and waste isolation. The NRC QA requirements
relating to data that are important to safety or to waste isolation are summarized in Sections
2.3.1, 2.3.2 and 2.3.3, below.
2.3.1 Data Gathering
Site characterization involves data gathering. As required by Subpart B of 10 CFR part 60,
DOE must conduct a program of site characterization hi accordance with the following:
• Investigations to obtain the required information shall be conducted in such a manner
as to limit adverse effects on the long-term performance of the geologic disposal
system to the extent practical.
• The number of exploratory boreholes and shafts shall be limited to the extent practical
consistent with obtaining the information needed for site characterization.
• To the extent practicable, exploratory boreholes and shafts in the geologic disposal
system operations area shall be located where shafts are planned for underground
facility construction and operation or where large unexcavated pillars are planned.
• Subsurface exploratory drilling, excavation, and in situ testing before and during
construction shall be planned and coordinated with geologic disposal system
operations area design and construction.
DOE must submit to the NRC a description of the QA program to be applied during the site
characterization phase. As a result of meeting the requirements of the QA Plan, Q-lists will
be generated. The criteria developed hi preparing Q-lists are essential to identifying quality-
affecting, activities and, of necessity, require a disciplined, systematic analysis of the entire
project (NRC88b).
The Q-list identifies structures, systems, and components important to safety and engineered
barriers important to waste isolation. A quality activities list identifies the site
characterization activities that may provide data for use in assessments of the waste isolation
and containment capabilities of natural and engineered barriers, those activities related to the
actual assessments, and those activities that may adversely impact the waste isolation
capabilities of these barriers.
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Data and information needs are identified by compliance assessments, performance
allocation6 among the various components of the natural and engineered barrier systems,
design, and modeling of the geologic disposal system. The need to collect additional data
depends on the availability and quality of existing data. The QA criteria for data gathering
are provided in Criteria 3, 5, 8, 9, 10, 11, 12, 13, 17, and 18 listed above.
2.3.2 Data Analysis
Data analysis is a design activity subject to the requirements of design control, design
verification, and design changes control. The QA criteria for data analysis are provided in
Criteria 3, 17 and 18.
2.3.3 Data Modeling
The data needed for construction of an adequate model of the disposal system and
compliance assessments, and the associated computer modeling, are subject to 10 CFR 60
Subpart G QA requirements.
Computer programs should be developed, controlled, and used in accordance with the QA
program. Guidance for documentation of computer codes is provided by NUREG-0856,
"Final Technical Position on Documentation of Computer Codes for High-Level Waste
Management" (NRC83)7. Documentation includes five categories: software summary,
description of mathematical models and numerical methods, user's manual, code assessment
and support, and continuing documentation and code listing.
General recommendations for software quality assurance programs are provided in
NUREG/CR-4640 "Handbook of Software Quality Assurance Techniques Applicable to the
Nuclear Industry" (NRC87). The handbook is intended to be used by the nuclear power
industry as an aid for structuring QA programs and assessing the adequacy of existing
software practices including development and use.
6 Performance allocation: This term applies to the process of deriving subsystem and component performance
goals from performance objectives. A systematic process of assigning confidence levels with their desired,
associated performance goals for the mined geologic disposal systems, subsystems, and components (NRC88b).
7 NUREG-0856 does not comply completely with ASME NQA Part 2.7.
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Guidance for NRC organizations and NRC contractors in the development and maintenance
of software for use by NRC staff is provided in NUREG/BR-0167 "Software Quality
Assurance Program and Guidelines" (NRC93). Those guidelines apply to technical
application software used hi safety decisions by the NRC. The applicability of those
guidelines depend on the purpose and use of the software and management's judgment of the
cost-effectiveness of each software quality activity. Most projects incorporate verification
and validation, configuration management, and documentation control activities.
Verification is the process of ensuring that the products and processes of each of the major
life cycles' activities meet the standards for the products and the objectives of that activity.
Validation is the process of demonstrating that the as-built software meets its requirements
and is accomplished by review and demonstration in a live or simulated environment.
Verification and validation activities include planning, formal life cycle reviews and audits,
peer inspections, and testing. Testing is the process of detecting errors and verifying
performance. Testing typically includes unit integration, qualification, and acceptance
testing.
Fundamental to configuration management are the concepts of a baseline and change control.
A baseline is a document or software that has been formally reviewed and agreed upon by
the developer and sponsor, and thereafter serves as the basis for further development. It can
be changed only through formal change control procedures. Change control is the process
by which a change to a baseline is proposed, evaluated, approved or rejected, scheduled, and
tracked.
Peer reviews may be employed for data modeling and computer models. Guidance on the
use of the peer review process is provided in NUREG-1297 "Peer Review for High-Level
Nuclear Waste Repositories" (NRC88). A peer review is a documented, critical review
performed by experts who are independent of the work being reviewed. NUREG-1297
provides guidance on areas where a peer review is appropriate, the acceptability of peers,
and the conduct and documentation of a peer review. Peer review is discussed in greater
detail in Chapter 7.
The QA criteria relating to data and computer modeling are provided in Criteria 3, 6, 8, 9,
11, 14, 15, 17 and 18 from the above list.
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2.3.4 Management and Oversight
NRC QA procedures describe how DOE and prime contractors exercise responsibility for the
overall QA program. DOE and its prime contractors are required to identify a management
position within each respective organization that retains overall authority and responsibility
for the QA program. This position must:
• Be at the same or higher organization level as the highest line manager directly
responsible for performing activities affecting quality and be sufficiently independent
from cost and schedule;
• Have effective communication channels with other senior management positions; and
• Have no duties or responsibilities unrelated to QA that would prevent full attention to
QA matters.
Persons and organizations performing QA functions must have sufficient authority and
organizational freedom to:
• Identify quality problems;
• Initiate, recommend, or provide solutions through designated channels;
• Verify implementation of solutions; and
• Assure that further processing, delivery, installation, or operation is controlled until a
nonconformance, deficiency, or unsatisfactory condition has been corrected.
The QA program should provide control over all activities affecting the quality of the
identified activities, structures, systems, and components to an extent consistent with their
required performance (NRC88b).
The QA criteria related to management and oversight are provided in Criteria 1, 2, 15, 16,
and 18.
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2.3.5 Qualification of Older Data
Data pertinent to waste isolation systems and/or components may exist that were developed
before the implementation of a 10 CFR part 60 Subpart G QA program by DOE and its
contractors. Additionally, data that were developed outside of the DOE disposal system
program may be identified as pertinent and used by DOE for purposes of waste isolation,
such as data generated by oil companies, national laboratories, universities, or data that have
been published in technical or scientific publications. These are considered "existing data".
This category does not include information accepted by the scientific and engineering
community as established facts such as are found in engineering handbooks, density tables,
gravitational laws.
NRC specifies that procedures should be established describing methods of reviewing and
qualifying existing data. NUREG-1298 "Qualification of Existing Data for High-Level
Nuclear Waste Repositories, Generic Technical Position" (NRC88a) describes four general
approaches, but does not provide implementation guidance for this process.
2.4 ASME NQA-1, NQA-2, AND NQA-3 STANDARDS
EPA is requiring that DOE implement a QA program that meets the requirements of the
American Society of Mechanical Engineers' "Quality Assurance Program Requirements for
Nuclear Facilities" (NQA-1-1989 edition); "Quality Assurance Requirements of Computer
Software for Nuclear Facility Applications" (NQA-2a-1990 addenda (part 2.7) to ASME
NQA-2-1989 edition); and "Quality Assurance Program Requirements for the Collection of
Scientific and Technical Information for Site Characterization of High-Level Nuclear Waste
Repositories" [NQA-3-1989 edition, excluding Section 2.1(b) and (c)]8. EPA has mandated
8 EPA proposes the use of the ASME NQA 1989 editions instead of the ASME NQA 1994 edition for the
following reasons:
• The ASME NQA 1994 edition allows a reduction in the level of structure for personnel
qualification and certification for designers/verifiers, inspection personnel, testing and audit
personnel to only a subjective level by supervisory analysis.
• ASME NQA-3-1989 contains added amplification requirements related to Scientific and
Technical Data (S&TD) applications to support the WIPP that are not included in ASME NQA
1994 edition. These areas include: (1) planning,quality standards and criteria for the collection
of S&TD; (2) surveillance, including in-process, deficiencies, and follow-up; (4)
communication; (5) design control; (6) peer reviews; (7) data processing; and (8) qualification
of existing data.
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the use of the ASME standards because these national consensus standards offer the most
comprehensive and specific set of QA requirements for all compliance-related elements of the
WIPP disposal system. NRC has taken a similar approach by specifying equivalent criteria
hi 10 CFR part 50, Appendix B.
2.4.1 ASME NOA-1
This Standard sets forth requirements for the establishment of a quality assurance program
for the siting, design, construction, operation, and decommissioning of nuclear facilities.
This Standard's requirements apply to activities which could affect the quality of structures,
systems, and components of nuclear facilities. These activities include:
• attaining quality objectives;
• assuring that an appropriate quality assurance program is established; and
• verifying that activities affecting quality have been correctly performed.
Activities affecting quality include siting, designing, purchasing, fabricating, handling,
shipping, receiving, storing, cleaning, erecting, installing, inspecting, testing, operating,
maintaining, repairing, modifying and decommissioning (ASM89a).
2.4.2 ASME NOA-2. Part 2.7
Part 2.7 provides requirements "for the development, procurement, maintenance, and use of
computer software, as applied to the design, construction, operation, modification, repair,
and maintenance of nuclear facilities" (ASM89b). It supplements the requirements of
NQA-1.
2.4.3 ASME NOA-3
NQA-3 was developed specifically for site characterization of high-level nuclear waste
repositories. (The same QA considerations apply to repositories for transuranic wastes.)
NQA-3 is to be used in conjunction with NQA-1 to set forth QA program requirements and
guidance for the collection of scientific and technical information for site characterization of
high-level nuclear waste repositories. "The requirements of NQA-1 and NQA-3 are intended
to meet and clarify the criteria of 10 CFR 50, Appendix B, and 10 CFR 60, Subpart G, for
high-level nuclear waste repositories" (ASM89c).
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2.5 REFERENCES
ASM89a American Society of Mechanical Engineers, "Quality Assurance Program
Requirements for Nuclear Facilities," ASME NQA-1-1989 Edition.
ASM89b American Society of Mechanical Engineers, "Quality Assurance Requirements
of Computer Software for Nuclear Facility Applications," Part 2.7, ASME
NQA-2-1990 Edition (First Addenda to ASME NQA-2-1989 Edition).
ASM89c American Society of Mechanical Engineers, "Quality Assurance Program
Requirements for the Collection of Scientific and Technical Information for
Site Characterization of High-Level Nuclear Waste Repositories," ASME
NQA-3-1989 Edition.
DOE91 U.S. Department of Energy, "Quality Assurance," DOE Order 5700.6C,
August 21, 1991.
DOE91a U.S. Department of Energy, Office of Environmental Restoration and Waste
Management, "Quality Assurance Requirements and Description (QARD),"
Revision 0, October 31, 1991.
DOE92 U.S. Department of Energy, "Quality Assurance Program Plan for the Waste
Isolation Pilot Plant Experimental-Waste Characterization Program," Waste
Isolation Pilot Project, DOE/EM/48063-1, Final Draft, Revision 2.0, October
1992.
DOE93a U.S. Department of Energy (M. Frei) letter dated July 9, 1993, to EPA (J.
William Gunter), identification of issues.
DOE93b U.S. Department of Energy (P. D. Grimm) letter dated March 31, 1993, to
U.S. Environmental Protection Agency (M. H. Shapiro), Attn: Docket No.
A-92-56, comments on ANPR 58 FR 8029.
DOE94 U.S. Department of Energy, Carlsbad Area Office, "TRU Waste
Characterization Quality Assurance Program Plan," CAO-94-1010, DRAFT,
Revision B, July, 1994.
DOE95 U.S. Department of Energy, Carlsbad Area Office, "Quality Assurance
Program Description," CAO-94-1012, DRAFT, Revision 1.0, August, 1995.
EPA80 U.S. Environmental Protection Agency Interim Guidelines and Specifications
for Preparing Quality Assurance Program Plans, QAMS-005-080.
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NRC83 U.S. Nuclear Regulatory Commission, S! A. Silling, Office of Nuclear
Materials Safety and Safeguards, "Final Technical Position on Documentation
of Computer Codes for High-Level Waste Management," NUREG-0856, June
1983.
NRC84 U.S. Nuclear Regulatory Commission, Altaian, W., et al.,"Improving Quality
and the Assurance of Quality in the Design and Construction of Commercial
Nuclear Power Plants, NUREG-1055, May 1984.
NRC87 U.S. Nuclear Regulatory Commission/Pacific Northwest Laboratory, J. L.
Bryant, N. P. Wilburn, "Handbook of Software Quality Assurance Techniques
Applicable to the Nuclear Industry," NUREG/CR-4640 (PNL-5784), August
1987.
NRC88 U.S. Nuclear Regulatory Commission, Altaian, W.D., et al., "Peer Review
for High-Level Nuclear Waste Repositories," NUREG-1297, February 1988.
NRC88a U.S. Nuclear Regulatory Commission, W. D. Altaian, J. P. Donnelly, J. E.
Kennedy, Office of Nuclear Materials Safety and Safeguards, "Qualification of
Existing Data for High-Level Nuclear Waste Repositories," Generic Technical
Position, NUREG-1298, February 1988.
NRC88b U.S. Nuclear Regulatory Commission, A. B. Duncan, S. J. Bilhorn, J. E.
Kennedy, Office of Nuclear Materials Safety and Safeguards, "Technical
Position on Items and Activities in the High-Level Waste Geologic Repository
Program Subject to Quality Assurance Requirements," NUREG-1318, April
1988.
NRC89 U.S. Nuclear Regulatory Commission, "Review Plan for High-Level Waste
Repository Quality Assurance Program Descriptions," Revision 2, March
1989.
NRC93 U.S. Nuclear Regulatory Commission, "Software Quality Assurance Program
and Guidelines," NUREG/BR-0167, February 1993.
NRC93a U.S. Nuclear Regulatory Commission (B. Youngblood) letter dated March 22,
1993, to EPA (J. W. Gunter), comments on ANPR 58 FR 8029.
SNL92 Sandia National Laboratories, WIPP Performance Assessment Department,
"Preliminary Performance Assessment for the Waste Isolation Pilot Plant,
December 1992, Volume 2: Technical Basis," SAND92-0700/2, December
1992.
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SNL92 Sandia National Laboratories, WIPP Performance Assessment Department,
"Preliminary Performance Assessment for the Waste Isolation Pilot Plant,
December 1992, Volume 1: Third Comparison with 40CFR191, Subpart B,"
SAND92-0700/2, March 1993.
SNL95a Sandia National Laboratories, Waste Isolation Pilot Plant, Quality Assurance
Procedure QAP 20-3, "Qualification of Existing Data", Revision 2, June 28,
1995.
SNL95b Sandia National Laboratories, "Waste Isolation Pilot Plant Quality Assurance
Program Description," Rev. R, July, 1995.
WTA94 WTAC Independent Review Team (IRT) Desk Instruction IRT DI 1.0,
Revision 1, Independent Review Team Qualifications, Selection and Roles,
November, 1994.
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3. Issues for the Selection and Development of Models and Computer Codes
3.1 INTRODUCTION
In assessing the containment characteristics of the WIPP disposal system, DOE will need to
consider many complex processes upon which their decisions regarding performance
assessment will be based. Although some decisions can be made using engineering
judgment, many analyses must be performed where human reasoning alone is inadequate to
synthesize the many factors involved in complex problems. The best tool available to help
scientists meet the challenge of such analyses and predictions is a model.
A model is a system designed to represent a simplified version of a real system. Models,
can be valuable predictive tools for performance assessment if properly constructed. The
validity of the predictions will depend on how well or conservatively the model approximates
the physical system being modeled.
DOE is currently developing models and computer codes to meet performance assessment
objectives for the WIPP. EPA will ultimately accept or reject DOE's selection and
application of models and computer codes. This section identifies the issues considered by
EPA in the development of criteria for WIPP model and code selection, description,
implementation, and justification (§194.23). These criteria are based on full implementation
of quality assurance procedures and the complete documentation of the procedures.
DOE has already selected a number of computer codes at the WIPP to gain an insight into
the kinds of problems that may be encountered in the modeling analyses that will be
conducted for the performance assessment. In the process of DOE's continued formulation
and testing of the various components of the disposal system conceptual models, attention is
given to many aspects of the system and possible avenues of analysis. This includes
developing a conceptual model of the site that defines the physical framework, relevant
processes, boundary conditions, and what approaches are justifiable and relevant to meeting
performance objectives. After the conceptual model has been formulated, appropriate codes
are selected by matching a detailed description of the modeling needs with well-defined,
quality-assured characteristics of existing codes, while taking into account the compliance
assessment objectives of the study. If a good match between model requirements and code
characteristics cannot be found, modification of an existing code or the development of a new
code may be considered.
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In the 1992 PA, DOE presented components of its site conceptual model. These components
focused on the Culebra Dolomite Member of the Rustler Formation overlying the Salado
Formation. The Culebra Dolomite is thought to present the most likely avenue for
radioactive waste to reach the accessible environment in addition to direct releases to the
accessible environment. However, there is considerable uncertainty regarding the physical
system boundary conditions and flow and transport mechanisms.
Another consideration is that computer codes are generally not designed to be universally
applicable. Code development is normally aimed at solving a specific environmental problem
or range of problems. Therefore, a single code will not simulate all of the components of
the conceptual model. For example, most WIPP performance assessment scenarios consider
three time-dependent gas-generation processes which are expected to be involved in the
degradation of transuranic (TRU) waste in the disposal system: (1) oxic and anoxic
corrosion of metals, (2) aerobic and anaerobic microbial degradation of cellulosic materials,
and (3) radiolysis. The potential for large quantities of gas has strong links to other
processes associated with closure of the disposal rooms and panels. After the repository is
closed, the surrounding halite will close (creep) inward upon and compact the waste and
backfill. Gas generating processes in the waste and the impact of creep closure will
potentially increase pressure in the room. The pressure within these materials may force the
brine and compressed gas through natural and induced fractures toward the regulatory
boundaries.
Section 3.2 of this chapter discusses how a code review would first determine if the selected
code was compatible with the modeling objectives set forth in the performance assessment.
Section 3.3 focuses on the code development process, code capabilities, and the quality of the
accompanying documentation. Issues related to the application of a computer code are
considerably different than those associated with code development and selection (i.e., a
properly selected code can be improperly applied). Although the primary objective of this
chapter is to present issues related to model and code selection rather than code application,
Section 3.4 discusses code application criteria in a global sense.
3.2 REVIEW OF COMPUTER CODE BY EVOLUTION EVALUATION
Computer code selection is the process of choosing the appropriate software capable of
simulating the characteristics of the physical system to be modeled. The evolution of the
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computer code can be traced from the inception of the conceptual model to the formulation
of the mathematical model, and finally to the development of the computer code where
computer instructions for performing the operations specified in the mathematical model are
programmed. The formulation of a conceptual model is an integral component of the
modeling process. Components of the conceptual model may be simplified to meet either
limited objectives or limitations in the data. It is often useful to simulate only certain
components of the conceptual model. For instance, even if there are data that indicate
different geologic zones in the hydrogeologic unit, it is common practice to evaluate the
system as a function of average values. While different aspects of the conceptual model may
be simulated in a variety of ways, the selected approach must remain consistent with the
objectives. The review and acceptance of a model is an evolutionary process that depends
upon the modeling goals and availability of data. The following illustration is focused on
hydrogeologic models although the process would be nearly identical for other models.
3.2.1 Conceptual Models
The performance assessment will use numerous conceptual models to describe the physical,
biological and chemical processes expected to occur at WIPP. At the most basic level,
conceptual models describe very fundamental processes, for example, the type of
microorganisms present in the repository, their population size and metabolic rates. These
basic conceptual models are integrated further to form the basis for more complete
conceptual models that predict processes such as gas generation, room closure rates, and
contaminant transport mechanisms.
The conceptual model of a groundwater system is an interpretation of the characteristics and
dynamics of the physical hydrogeologic system. The purpose of the conceptual model is to
consolidate site and regional hydrogeologic and hydrologic data into a set of assumptions and
concepts that can be evaluated quantitatively. The system conceptualization should include:
the geologic and hydrologic framework, characteristics of geologic formations (e.g.,
fractured or porous), the nature of relevant physical and chemical processes, time dependent
processes, geometry of the system, initial and boundary conditions, hydraulic properties; and
sources and sinks (water budget). A brief discussion of each of the typical components of
the hydrogeological conceptual model that should be considered and reviewed is presented
below.
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Geologic framework. The geologic framework is the distribution and configuration of
various rock units (e.g., fractured dolomite or intact halite). Of primary interest are the
thickness, continuity, lithology, and geologic structure of those units that are relevant to the
purpose of the study.
Hydrologic framework. The hydrologic framework in the conceptual model includes the
physical extent of the flow system, hydrologic features that impact or control the
groundwater flow system, analysis of groundwater flow directions, and media type. The
conceptual model must address the degree to which the system behaves as a porous media.
If the system is significantly fractured or solution channeled, the conceptual model must
address these issues.
Hydraulic properties. The hydraulic properties include the transmissive and storage
characteristics of the rocks and properties of the fluids. Specific examples of rock and fluid
properties include transmissivity, hydraulic conductivity, storativity, fluid viscosity and
densities. Hydraulic conductivity may also have directionality (anisotropy).
Sources and sinks. Sources or sinks of water or gas impact the pattern and rate of flow and
may affect the transport of radionuclides from the repository. The most common examples
of sources and sinks include pumping or injection wells, infiltration, evapotransporation,
drains and flow from surface water bodies. At the WIPP reactions between brine and waste
in the repository may provide a source of gas.
Boundary and initial conditions. Boundary conditions are the conditions the modeler
specifies, typically on the perimeter of the model domain, in order to solve for the unknowns
in the problem domain. These values may be associated with either the groundwater flow or
the contaminant transport aspects of the problem. Groundwater boundaries may be described
in terms of where water and/or gas are flowing into the groundwater system and where water
and/or gas are flowing out. Many different types of boundaries exist, Including: surface
water bodies, groundwater divides, rainfall, wells, and geologic features such as faults and
sharp contrasts in lithology. For example, at the WIPP, pressure boundary conditions for the
Salado have been set to far field fluid pressures.
The most common contaminant-source boundaries specify the source concentration or
prescribe the mass flux of contamination entering the system. Both type of source boundaries
are currently used in the modeling at the WIPP
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Initial conditions are defined as values of ground water elevation, pressure, flow volumes, or
contaminant concentrations which are initially assigned to interior areas of the modeled
regions. At the WIPP, pressures in the repository are initially set to atmospheric conditions.
Transport processes. The transport of radionuclides by flow through either a porous matrix
or a fracture system in a porous matrix will be affected by various mechanical and
geochemical processes. The dominant mechanical processes are advection, dispersive effects
(hydrodynamic dispersion, channeling) and diffusion. The chemical processes potentially
affecting radionuclide transport include: adsorption on mineral surfaces (both internal and
external to the crystal structure), speciation, precipitation, colloidal transport, radiolysis,
biofixation, natural organic matter interactions, anion exclusion, and complexation.
Spatial dimensionality. Groundwater flow and contaminant transport are seldom constrained
to one or two dimensions. However, in some instances, it may be appropriate to restrict the
analysis to one or two dimensions. The decision to model a site in a particular number of
dimensions should be based upon the modeling objectives and the availability of field and/or
laboratory data.
Temporal dimensionality. Either steady-state or transient flow simulations can be performed.
At steady-state, it is assumed that the flow field remains constant with time, whereas a
transient system simply means one that changes with time. Steady-state simulations produce
average or long-term results and require that a true equilibrium case is physically possible.
Transient analyses are typically performed when boundary conditions are varied through time
or when study objectives require answers at more than one tune.
The conceptual model is based on the modeler's experience, technical judgment and
represents the modeler's understanding of the physical system being modeled. The
conceptual model will become more complex as more processes are identified and
interrelationships of important components within the systems are considered. The
transformation of the conceptual model into a mathematical model is only an extrapolation of
a basic understanding of the system will result in simplifications of the system. For
example, the mathematical models assume that there is a direct scaling between the model
simulations and the scale at which the data are collected. The lack of knowledge about the
system resulting from limited information also contributes to simplifications of the
mathematical models.
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In addition to the unavoidable simplifications of the conceptual model, there are
simplifications in which the modeler decides what physical characteristics and processes are
important to the model application. Examples of these kind of simplifying assumptions
include:
• Flow through the unsaturated zone is vertical and in one
dimension.
• Chemical reactions are reversible and instantaneous.
• Soil or rock medium is isotropic and/or homogeneous.
• Flow field is uniform and under steady-state conditions.
As more data become available, simplifying assumptions are removed and the conceptual
model complexity increases. This process creates mathematical model development which
allows for the systematic integration of previously neglected conceptual model components.
3.2.2 Mathematical Models
A conceptual model describes the present condition of the system. To make predictions of
future behavior it is necessary to develop mathematical models. Laboratory sand tanks are
physical scale models that simulate groundwater flow directly. The flow of groundwater can
also be implied using electrical analog models. Mathematical models, including analytical
and numerical methods, discussed below, are more widely used because they are easier to
develop and manipulate.
A mathematical model is essentially a mathematical representation of a process or system
conceptual model. For example, the mathematical model for groundwater flow is derived by
applying principles of mass conservation and conservation of momentum. The generally
applicable equation of motion in groundwater flow is Darcy's law for laminar flow, which
originated in the mid-nineteenth century as an empirical relationship. Later, a mechanistic
approach related this equation to the basic laws of fluid dynamics. In order to solve the flow
equation, both initial and boundary conditions are necessary.
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Solution Methodology
Every ground water model is based upon a set of mathematical equations. Solution
methodology refers to the strategy and techniques used to solve these equations. The
equations are normally solved for water elevations in the subsurface (head) and/or
contaminant concentrations.
Mathematical methods developed to solve groundwater flow and transport equations can be
broadly classified as either deterministic or stochastic. Deterministic methods assume that a
system or process operates such that the occurrence of a given set of events leads to a
uniquely definable outcome. Stochastic methods presuppose the outcome to be uncertain and
are structured to account for this uncertainty.
Most of the stochastic methods are not purely stochastic because they often utilize a
deterministic representation of soil processes and derive their stochastic nature from their
representation of inputs and/or spatial variation of soil characteristics and resulting chemical
movement. While the deterministic approach results in a specific value of a soil variable
(e.g., solute concentration) the stochastic approach provides the probability of a specific
value occurring at any point.
The development of stochastic methods for solving groundwater flow is a relatively recent
endeavor that has occurred as a result of the growing awareness of the importance of
intrinsic variability of the hydrogeologic environment. Stochastic methods are still primarily
research tools; however, as computer speeds continue to increase, stochastic methods will be
able to move further away from the research-oriented community and more into mainstream
management applications. This discussion focuses primarily on deterministic methods.
Deterministic methods may be classified as either analytical or numerical. Analytical
methods usually involve approximate or exact solutions to simplified forms of the differential
equations for water movement and solute transport. Simple analytical methods are based on
the solution of differential equations which give qualitative estimates of the extent of
contaminant transport. Such methods are simpler to use than numerical methods and can
generally be solved with the aid of a calculator or computers. Analytical methods are
restricted to simplified representations of the physical situations and generally require only
limited site-specific input data. They are useful for screening sites and scoping the problem
to determine data needs or the applicability of more detailed numerical methods.
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Analytical methods are used in groundwater investigations to solve many different kinds of
problems. For example, aquifer parameters (e.g., transmissivity, storativity) are obtained
from aquifer tests through the use of analytical methods. To avoid confusion, only analytical
methods designed to estimate groundwater flow and radionuclide transport rates are discussed
in this section.
Analytical methods that solve groundwater flow and contaminant transport in porous media
are comparatively easy to use. Analytical solutions are generally restricted to either radial
flow problems or to cases where velocity is uniform over the area of interest. Except for
some radial flow problems, almost all available analytical solutions are developed for systems
having a uniform and steady flow. This means that the magnitude and direction of the
velocity throughout the system are invariable with respect to time and space.
Equations of flow and continuity in the form of partial differential equations do not lend
themselves easily to rigorous analytical solutions when boundaries are complex. Generally, a
realistic analytical expression for hydraulic head or concentration as a function of space
cannot be written from the governing equations, boundary and initial conditions, and
therefore analytical methods are generally abandoned and replaced by more sophisticated
numerical methods.
Numerical methods provide solutions to the differential equations describing water movement
and solute transport using methods such as finite differences and finite elements. Numerical
methods can account for complex geometry and heterogenous media, as well as for
dispersion, diffusion, and chemical retardation processes (e.g., sorption, precipitation,
radioactive decay, ion exchange, degradation). These methods always require a digital
computer, greater quantities of data than analytical modeling, and an experienced modeler-
hydrogeologist.
3.2.3 Numerical Models
A numerical model for groundwater flow consists of the mathematical framework for the
solution of the material balance equations that govern laminar flow through porous or
fractured media. These mass balance equations are dependent upon physical constraints and
constitutive relationships. The constraints simply state conditions that components of the
mass balance equations must satisfy, whereas the constitutive relationships describe the
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dependence of parameters, in the mass balance equations, on other physical processes.
Furthermore, the mass balance equations are composed of both spatial and temporal terms
both of which require discretization within the model domain. These terms describe the head
or pressure in space and time. Either finite element or finite difference methods can be used
to discretize the spatial term in the mass balance equations, whereas finite differences are
almost always used to discretize the temporal term.
The mass balance equations, physical constraints, and constitutive relationships lead to a
series of equations that must be solved in space and time. The means by which the equations
are discretized, linearized (e.g. Newton-Raphson), organized (i.e, matrix construction) and
solved with either direct or iterative methods is all part of the numerical model.
3.2.4 Computer Code
It is important that the progression from the conceptual model to the computer model is
documented in detail. The discussion for each component of the conceptual model should
begin with the laboratory or field studies that provide the fundamental characterization data.
Next, the data analysis results that support a particular conceptualization should be presented.
As part of the data analysis discussion the basis for screening out reasonable alternative
conceptual models should be provided. This type of discussion should be presented not only
for major components of the conceptual model (e.g., Darcian versus non-Darcian flow) but
also for more obscure assumptions (e.g., sorption may be described by a linear isotherm).
Following the development of each major and minor component of the conceptual model, the
formulation of the mathematical models and numerical models should be presented.
The linkage between the conceptual model(s), mathematical model(s) and numerical model(s)
should be clearly described. For example, conceptual models generally assume a continuum
in space and time, whereas mathematical models frequently divide space and time into user
specified segments. Furthermore, the implications that this type of simplifying assumption
may have on the modeling predictions should be presented.
Following the formulation of the numerical model, the computer program is developed. The
computer program consists of the assembly of numerical techniques, bookkeeping, and
control language that represents the model from acceptance of input data and instructions to
delivery of output.
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In summary, the conceptual model is a working description of the characteristics and
dynamics of a physical system. Model construction is the process of transforming the
conceptual model into a non-unique, simplified, mathematical description of the physical
system, coded in computer programming language together with a quantification of the
simulated system. An intermediate step in the model transformation process is the
mathematical model which consists of two aspects: a process equation and a solution
technique to solve the process equation. An analytical solution solves a very simple process
equation analytically by hand calculations. An analytical model solves a more complex, but
still relatively simple, process equation analytically with a computer program. A numerical
model solves a simple or complex process equation numerically with a computer program.
In the context of this document, mathematical model refers to all three solution techniques of
a process equation. The complexity of the process equation dictates the solution technique
required. The model formulation process concludes with the coding of the mathematical
model into computer programming language for performing a specified set of operations.
3.3 CODE-RELATED ISSUES
The determination of a computer code's acceptability for a particular application at the WIPP
depends on whether the code can meet the modeling objectives. The code evaluation process
must also consider attributes that are integral components of the computer code(s) including:
• Source Code Availability
• History of Use
• Quality Assurance
• Code Documentation
• Code Testing
• Hardware Requirements
• Solution Methodology
• Code Dimensionality
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3.3.1 Source Code Availability
Detailed documentation of the software, source code and developmental history is required to
facilitate independent review. Independent evaluation of the reproducibility of the
verification and validation results require compiled version of the code (i.e., computer code
in machine language) should be available to the reviewer, together with files containing the
original test data used in the code's verification and validation.
3.3.2 History of Use
The evaluation process should rely on documented user experience, in addition to hands-on
experience and testing. User experience is especially valuable in determining whether the
code functions as documented or has significant errors and shortcomings. In some instances
users independent of the developer should perform extensive testing and bench-marking.
3.3.3 Quality Assurance
Code acceptance issues are closely tied to the quality assurance procedures followed during
the developmental process of the computer code. These criteria are associated with the
adequacy of the code testing and documentation.
Quality assurance in modeling is the procedural and operational framework put in place by
the organization managing the modeling program, to assure technically and scientifically
adequate execution of all project tasks included in the program, and to assure that all
modeling-based analysis is verifiable and defensible (TAY85).
The American Society of Mechanical Engineers (ASME) Committee on Nuclear Quality
Assurance has developed standards for the development and use of computer software used in
the design and operation of nuclear facilities (ASM90). The standard in NQA-2a-1990
Addenda (Part 2.7) was developed under procedures accredited as meeting the criteria for
American National Standards. It addresses the following:
• general requirements
• software life cycle
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• software verification and validation
• software configuration control
• documentation
• verification reviews
• problem reporting and corrective action
• access control
• software procurement
• records
Quality assurance requirements in 40 CFR part 194 (§194.23(b)) mandate the use of this
standard as the basis for documenting any computer codes used to support a compliance
application.
The two major elements of quality assurance are quality control and quality assessment.
Quality control refers to the procedures that ensure the quality of the final product. These
procedures include the use of appropriate methodology in developing and applying computer
simulation codes, adequate verification and validation procedures, and proper usage of the
selected methods and codes (HEI92). To monitor the quality control procedures and to
evaluate the quality of the studies, quality assessment is applied (HEI89).
Software quality assurance (SQA) consists of the application of procedures, techniques, and
tools through the software life cycle, to ensure that the products conform to pre-specified
requirements (BRY87). This requires that in the initial stage of the software development
project, appropriate SQA procedures (e.g., auditing, design inspection, code inspection,
error-prone analysis, functional testing, logical testing, path testing, reviewing, walk-through)
and tools (e.g., text-editors, software debuggers, source code comparitors, language
processors) need to be identified and the software design criteria be determined (HE92).
Quality assurance for code development and maintenance implies a systematic approach,
starting with the careful formulation of code design objectives (section 3.2), criteria, and
standards, followed by an implementation strategy. The implementation strategy includes the
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design of the code structure and a description of the way in which software engineering
principles will be applied to the code. In this planning stage, measures are to be taken to
ensure complete documentation of code design and implementation, record-keeping of the
coding process, description of the purpose and structure of each code segment (functions,
subroutines), and record-keeping of the code verification process.
Records for the coding and verification process may include a description of the fundamental
algorithms describing the physical process(es) to be modeled; the means by which the
mathematical algorithms have been translated into computer code (e.g., FORTRAN); results
of discrete checks on the subroutines for accuracy; and comparisons among the codes'
numerical solutions with either analytical or other independently verified numerical solutions.
Software used for compliance assessment should have both internal and external
documentation. Internal documentation, which is part of the source code, describes the
operation of the program and includes the name of the author, other sources of the software,
and its revision history. External documentation includes a software abstract, an on-line help
file stored on the applicable computer system, records of verification and changes, and
formal reports including a theory manual and a user's manual.
Code verification or testing ensures that the underlying mathematical algorithms have been
correctly translated into computer code. The verification process varies for different codes
and ranges from simply checking the results of a plotting routine to comparing the results of
the computer code to known analytical solutions or to results from other verified codes.
Traceability describes the ability of the performance assessment analyst to identify the
software that was used to perform a particular calculation, including it's name, date, and
version number, while retrievability refers to the availability of the same version of the
software for further use.
3.3.3.1 Code Documentation
In general, the code documentation should describe the theoretical framework represented by
the model on which the code is based, code structure and language standards applied, and
code use instructions regarding model setup and code execution parameters. The
documentation should also include a complete treatment of the equations on which the model
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is based, the underlying mathematical and conceptual assumptions, the boundary conditions
that are incorporated in the model, the method and algorithms used to solve the equations,
and the limiting conditions resulting from the chosen approach. The documentation should
also include user's instructions for implementing and operating the code and preparing data
files. It should present examples of model formulation (e.g., grid design, assignment of
boundary conditions), complete with input and output file descriptions, and include an
extensive code verification and validation or field testing report.
Code Documentation Issues. An integral part of the code development process is the
preparation of the code documentation. This documentation of QA in model development
consists of reports and files pertaining to the development of the model and could include:
• A report on the development of the code including the
(standardized and approved) programmer's bound notebook
containing detailed descriptions of the code verification process;
• Verification report including verification scenarios, parameter
values, boundary and initial conditions, source-term conditions,
dominant flow and transport processes;
• Orientation and spacing of the grid and justification;
• Time-stepping scheme and justification;
• Changes and documentation of changes made in code after
baselining;
• Executable and source code version of baselined code;
• Input and output (numerical and graphical) for each verification
run;
• Notebook containing reference material (e.g., published papers,
laboratory results, programmer's rationale ) used to formulate
the verification problem.
Furthermore, the purpose of the software documentation is to (GAS79):
• record technical information that enables system and program
changes to be made quickly and effectively;
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• enable programmers and system analysts, other than software
originators, to use and to work on the programs;
• assist the user in understanding what the program is about and
what it can do;
• increase program sharing potential;
• facilitate auditing and verification of program operations;
• provide managers with information to review at significant
developmental milestones so that they may independently
determine that project requirements have been met and that
resources should continue to be expended;
• reduce disruptive effects of personnel turnover;
• facilitate understanding among managers, developers,
programmers, operators, and users by providing information
about maintenance, training, and changes in and operation of the
software;
• inform other potential users of the functions and capabilities of
the software, so that they can determine whether it serves their
needs.
The user's information could consist of items such as:
• an extended model description;
• model input data description and format;
• type of output data provided;
• code execution preparation instructions;
• sample model runs;
• trouble shooting guide; and
• contact person/affiliated office.
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The programmer's information could consist of items such as:
• model specifications;
• model description;
• flow charts;
• descriptions of routines;
• database description;
• source listing;
• error messages; and
• contact person/affiliated office.
The analyst's information could consist of items such as:
• a functional description of the model;
• model input and output data;
• code verification and validation information; and
• contact person/affiliated office.
The code itself should be well structured and internally well documented; where possible,
self-explanatory parameter, variable, subroutine, and function names should be used.
3.3.3.2 Code Testing
Before a code can be used as a planning and decision-making tool, its credentials must be
established through systematic testing of the model's correctness and evaluation of the
model's performance characteristics (HEI89). Of the two major approaches available, the
evaluation or review process is qualitative in nature, while code-testing results can be
expressed using quantitative performance measures. Code testing (or code verification) is
aimed at detecting programming errors, testing embedded algorithms, and evaluating the
operational characteristics of the code through its execution on carefully selected example test
problems and test data sets. ASME standard in NQA-2a-1990 Addenda (Part 2.7) defines
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software verification as the process of determining whether or not the product of a given
phase of the software development cycle fulfills the requirements imposed by the previous
phase.
It is necessary to distinguish between generic simulation codes based on an analytical solution
of the governing equation(s) and codes that include a numerical solution. Verification of a
coded analytical solution is restricted to comparison with independently calculated results
using the same mathematical expression, i.e., manual calculations, using the results from
computer programs coded independently by third-party programmers. Verification of a code
formulated with numerical methods might take two forms: (1) comparison with analytical
solutions, and (2) code intercomparison between numerically based codes, representing the
same generic simulation model, using synthetic data sets.
It is also important to distinguish between code testing and model testing. Code testing is
limited to establishing the correctness of the computer code with respect to the criteria and
requirements for which it is designed (e.g., to represent the mathematical model). Model
testing (or model validation) is more inclusive than code testing, as it represents the final step
in determining the validity of the quantitative relationships derived for the real-world system
the model is designed to simulate.
Attempts to validate models must address the issue of spatial and temporal variability when
comparing model predictions with limited field observations. If sufficient field data are
obtained to derive the probability distribution of contaminant concentrations, the results of a
stochastic model can be compared directly. For a deterministic model, however, the
traditional approach has been to vary the input data within its expected range of variability
(or uncertainty) and determine whether the model results satisfactorily match historical field
measured values. This code-testing exercise is sometimes referred to as history matching.
See Chapter 2 for additional Quality Assurance discussions.
3.3.4 Hardware Requirements
•In general, hardware requirements are rarely a discriminatory factor in the selection of a
computer code. However, a number of the codes that DOE intends to use in modeling the
WIPP will require very sophisticated hardware, not so much because of the intrinsic
requirements of the code but because the processes to be modeled are very complex.
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3.3.5 Mathematical Solution Methodology
Every groundwater or contaminant transport model is based upon a set of mathematical
equations. Solution methodology refers to the strategy and techniques used to solve these
equations. In groundwater modeling, the equations are normally solved for head and/or
contaminant concentrations. Other disposal system processes will be modeled with codes that
solve for gas-filled porosities and the quantity of radioactive material (in curies) brought to
the surface as cuttings generated by a drilling operation that penetrates the disposal system.
Analytical solutions are used in modeling investigations to solve many different kinds of
problems. For example, aquifer parameters are obtained from aquifer pumping and tracer
tests through the use of analytical models, and groundwater flow and contaminant transport
rates can also be estimated with the use of analytical models.
Numerical models provide solutions to the differential equations describing room collapse,
water movement, and solute transport using numerical methods such as finite differences and
finite elements. Numerical methods account for complex geometry and heterogeneous
media, as well as dispersion, diffusion, matrix deformation, salt creep and chemical
retardation processes (e.g., sorption, precipitation, radioactive decay, ion exchange,
degradation). These methods almost always require a digital computer, greater quantities of
data than analytical modeling, and experienced modelers.
The validity of the results from mathematical models depends strongly on the quality and
quantity of the input data. Stochastic, numerical, and analytical codes have strengths and
weaknesses inherent within their formulations.
3.3.6 Code Dimensionality
The determination as to the number of dimensions that a code should be able to simulate is
based primarily upon the modeling objectives and the dimensionality of the processes the
code is designed to simulate.
In determining how many dimensions are necessary to meet the objectives, a basic
understanding is needed of how the physical processes (e.g., salt creep, groundwater flow,
transport, dose rate) are affected by the exclusion or inclusion of an additional dimension.
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The movement of groundwater and contaminants is usually controlled by advective and
dispersive processes which are inherently three-dimensional. Advection is more responsible
for the tune (i.e., travel tune) it takes for a contaminant to travel from the source term to a
downgradient receptor, while dispersion directly influences the concentration of the
contaminant along its travel path. This fact is very important in that it provides an intuitive
sense for the effect dimensionality has on contaminant migration rates and concentrations.
As a general rule, the fewer the dimensions, the more the model results will over-estimate
concentrations and under-estimate travel times. In a model with fewer dimensions, predicted
concentrations will generally be greater because dispersion, which is a three-dimensional
process, will be dimension limited and will not occur to the same degree as it actually would
in the field. Similarly, predicted travel times will be shorter than the actual travel time, not
because of a change in the contaminant velocities but because a more direct travel path is
assumed. Therefore, the lower dimensionality models tend to be more conservative in their
predictions and are frequently used for screening analyses.
One-dimensional simulations of contaminant transport usually ignore dispersion altogether,
and contamination is assumed to migrate solely by advection, which results in a highly
conservative approximation. Vertical analyses in one dimension are generally reserved for
evaluating flow and transport in the unsaturated zone. In the 1992 PA modeling of the
Culebra Dolomite, advective and dispersive flow and transport were modeled in two
dimensions with SECO, whereas matrix diffusion was confined to one dimension. This type
of mixed-dimensionality approach is not uncommon early in a modeling analysis.
Two-dimensional analyses of an aquifer flow system can be performed as either a planar
representation, where flow and transport are assumed to be horizontal (i.e., longitudinal and
transverse components), or as a cross-section where flow and transport components are
confined to vertical and horizontal components. In most instances, two-dimensional analyses
are performed in an areal orientation, with the exception of the unsaturated zone, and are
based on the assumption that most contaminants enter the saturated system from above and
that little vertical dispersion occurs. However, a number of limitations accompany two-
dimensional planar simulations. These include the inability to simulate multiple layers (e.g.,
aquifers and aquitards) as well as any partial penetration effects. Furthermore, because
vertical components of flow are ignored, a potentially artificial lower boundary on
contaminant migration has been automatically assumed which may or may not be the case.
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A two-dimensional formulation of the flow system is frequently sufficient for the purposes of
risk assessment provided that flow and transport in the contaminated aquifer are essentially
horizontal. The added complexities of a site-wide, three-dimensional flow and transport
simulation are often believed to outweigh the expected improvement in the evaluation of risk.
Complexities include limited site-wide hydraulic head and lithologic data with depth and
significantly increased computational demands.
Quasi three-dimensional analyses remove some of the limitations inherent in two-dimensional
analyses. Most notably, quasi three-dimensional simulations allow for the incorporation of
multiple layers; however, flow and transport in the aquifers are still restrained to longitudinal
and transverse horizontal components, whereas flow and transport in the aquitards are even
further restricted to vertical flow components only. Although partial penetration effects still
cannot be accommodated in quasi three-dimensional analyses, this method can sometimes
provide a good compromise between the relatively simplistic two-dimensional analysis and
the complex, fully three-dimensional analysis. This is the case particularly if vertical
movement of contaminants or recharge from the shallow aquifer through a confining unit and
into a deeper aquifer is suspected.
Fully three-dimensional modeling generally allows both the geology and all of the dominant
flow and transport processes to be described in three dimensions. This approach is usually
the most reliable means of predicting groundwater flow and contaminant transport
characteristics, provided that sufficient representative data are available for the site.
3.4 MODEL APPLICATION
The application of a generic simulation model to a site-specific problem is often called
"model application" or "(computer/simulation) code application." The application of a
generic model to site-specific conditions should follow a well-structured model application
protocol. Such protocols are described by Mercer and Faust (MER81), van der Heijde et. al
(HEI88), and Anderson and Woessner (AND92), among others. Quality assurance in these
types of studies follows the same pattern as discussed for' generic model development projects
and consists of using appropriate data, data analysis procedures, modeling methodology and
technology, administrative procedures and auditing. To a large extent, the quality of a
modeling study is determined by the expertise of the modeling and quality assessment teams.
The following discussion is consistent with procedures found in an EPA-sponsored
publication (HEI92).
3-20
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Quality assurance in code application addresses all facets of the modeling process, including
such issues as:
• Historical review of code verification and benchmarking process;
• Correct and clear formulation of problems to be solved;
• Project description and objectives;
• Type of modeling approach to the project;
• Decision whether modeling is the best available approach and if
so, that the selected code is appropriate and cost-effective;
• Conceptualization of .system and processes, including
hydrogeologic framework, boundary conditions, stresses, and
controls;
• Detailed description of assumptions and simplifications, both
explicit and implicit (to be subject"to critical peer review);
• Data acquisition .and interpretation;
»
• Code selection considerations, or justification for modifying an
existing code or developing a new one;
• Model preparation (parameter selection, data entry or
reformatting, gridding);
• Validity of the parameter values used in the model application;
• Protocols for parameter estimation and model calibration to
provide guidance, especially for sensitive parameters;
• Level of information hi computer output (variables and
parameters displayed; formats; layout);
• Identification of calibration goals and evaluation of how well
they have been met (e.g., root-mean square errors, etc.);
• Role of sensitivity analysis in evaluating parameter uncertainties
and creating probability distributions;
3-21
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• Post-simulation analysis (including verification of reasonableness
of results, uncertainty analysis, and the use of manual or
automatic data processing techniques, as for contouring);
• Establishment of appropriate performance targets which should
characterize the limits of the data;
• Presentation and documentation of results;
• Evaluation of how closely the modeling results answer the
questions raised by management.
QA for model application should include complete record-keeping of each step of the
modeling process. The paper trail for QA should consist of reports and files addressing the
following items:
• Assumptions and limitations;
• Parameter or input values and sources including rationale for
their selection, range, and distribution;
• Boundary and initial conditions;
• Nature of grid and grid design justification;
• Changes and verification of changes made in code;
• Actual input used;
• Output of model runs and interpretation;
• Validation (or at least calibration) of model.
As is the case with code development QA, all data files, source codes, and executable
versions of computer software used in the modeling study should be retained for auditing or
post-project re-use (in hard-copy and, at higher levels, in digital form) including:
• Version of the source and executable image of the code used;
• Calibration input and output;
• Verification input and output;
3-22
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• Application input and output (e.g., for each of the scenarios
studied).
If the code used in the modeling study is modified, then the code should be tested again
according to a standard testing protocol; the code should be subject to the full QA procedure
for code development, including accurate record-keeping and reporting. All new input and
output files should be saved for inspection and possible re-use together with existing files,
records, codes, and data sets.
3.5 REFERENCES
AND92 Anderson, M.P., and W.W. Woessner, 1992. Applied Groundwater
Modeling: Simulation of Flow and Advective Transport. Academic Press,
Inc., San Diego, California.
ASM90 American Society of Mechanical Engineers (ASME), 1990. Quality Assurance
Program Requirements for Nuclear Facilities. ASME NQA-2a-1990, PART
2.7, Quality Assurance Requirements of Computer Software for Nuclear
Facility Applications, Am. Soc. of Mechanical Engineers, New York.
AST84 American Society for Testing and Materials (ASTM), 1984. Standard
Practices for Evaluating Environmental Fate Models of Chemicals. Annual
Book of ASTM Standards, E 978-84, Am. Soc. for Testing and Materials,
Philadelphia, Pennsylvania.
BRY87 Bryant, J.L., and N.P. Wilburn, 1987. Handbook of Software Quality
Assurance Techniques Applicable to the Nuclear Industry. NUREG/CR-4640,
Off. of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,
Washington, DC.
GAS79 Gass, S.I., 1979. Computer Model Documentation: A Review and an
Approach. NBS Special Publ. 500-39, Inst. for Computer Science and
Technology, Nat. Bur. of Standards, U.S. Dept. of Commerce, Washington,
DC.
HEI88 van der Heijde, P.K.M., and M.S. Beljin, 1988. Model Assessment for
Delineating Wellhead Protection Areas. EPA 440/6-88-002, Office of
Groundwater Protection, U.S. Environmental Protection Agency, Washington,
DC.
HEI89 van der Heijde, P.K.M., 1989. Quality Assurance and Quality Control in
Groundwater Modeling. GWMI 89-04. Internal. Ground Water Modeling
Center, Holcomb Research Inst., Indianapolis, Indiana.
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HEI92 van der Heijde, P.K.M., and O.A. Elnawawy, 1992. Compilation of
Groundwater Models. GWMI 91-06. International Ground Water Modeling
Center, Colorado School of Mines, Golden, Colorado.
MER81 Mercer, J.W. and C.R. Faust, 1981. Ground Water Modeling, Nat. Water
Well Assoc., Dublin, Ohio.
TAY85 Taylor, J.K., 1985. What is Quality Assurance? In: J.K. Taylor and T.W.
Stanley (eds.), Quality Assurance for Environmental Measurements, pp. 5-11.
ASTM Special Technical Publication 867, Am. Soc. for Testing and Materials,
Philadelphia, Pennsylvania.
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4. Waste Characterization
4.1 INTRODUCTION
Proper characterization of the waste slated for disposal at the WIPP site is an essential
element in assessing whether the repository meets the disposal standards set forth in 40 CFR
part 191. To set the stage for the subsequent discussion, it is first necessary to define certain
terms. Key definitions are as follows:
• A waste characteristic is a property of the waste that has an impact on the
containment of the waste in the disposal system. Examples of waste
characteristics include solubility of radionuclides, ability of the radionuclides
to become part of stable colloids, gas generation potential from corrosion,
microbial degradation or radiolysis, and various strength properties.
• A waste component is an ingredient of the total inventory of the waste that
influences a waste characteristic. Examples of waste components include the
quantity of metals, cellulosics and organic ligands, and the quantity of
radioactivity (curies) associated with each radionuclide.
• Waste characterization is the process of determining the chemical, radiological,
and physical properties of the waste. Waste characterization techniques
include the use of process knowledge, laboratory and field experimentation,
literature search, technical judgement, non-destructive examination/assay, and
destructive analysis.
This chapter discusses the various regulations including 40 CFR part 191 which drive waste
characterization, the ways in which waste characteristics can impact performance assessment
(PA), the methods used for characterizing the waste, and the rationale for the waste
characterization requirements of the 40 CFR part 194 rule.
4.1.1 Brief History of DOE's TRU Waste Characterization Program
DOE's TRU waste characterization program is based on the requirements developed for the
proper handling and disposal of TRU wastes intended for the WIPP. Historically, this
characterization has focused on two types of techniques; empirical — laboratory analyses to
quantify hazardous and radioactive waste constituents such as metals, volatile organic
compounds, Pu-239, etc.; and, informational based — the use of process/acceptable
knowledge derived from site operations to classify wastes according to established categories.
4-1
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This report describes the various facilities at TRU waste generator sites that are dedicated to
characterizing TRU wastes via routine or modified chemical and radiochemical analyses.1
Consistent with the 40 CFR part 194 rule, these aspects relate to waste components, as
defined above. However, this definition of waste characterization excludes waste
characteristics, which have historically been addressed under experimental programs, such as
the Actinide Source Term Program (ASTP). The definition of waste characterization must
be expanded accordingly to include these other aspects. As discussed in subsequent sections,
the current DOE waste acceptance criteria and waste characterization guidance documents do
not address the requirements of 40 CFR part 194 concerning waste characteristics.
4.1.1.1 DOE WIPP Waste Acceptance Criteria (WAC)
DOE developed and published tentative criteria for the acceptance of TRU waste produced
under the defense related programs (DOE91) in 1980. These criteria were developed to
ensure the safety of all operations at the WIPP. The waste acceptance criteria document was
intended to provide: 1) criteria for use in project design; 2) technical justification for the
WAC; and 3) quantitative guidelines to be used by waste generators for certifying TRU
wastes. The criteria do not specifically stipulate whether further waste treatment or
processing will be required, but DOE recognized that this decision would have to be made in
the future. Revision 4 of the WAC, published in 1991, included additional criteria relevant
to waste transportation and regulatory requirements for hazardous waste in order to provide a
single, comprehensive document for all parties involved with the shipment and handling of
WIPP waste. These criteria are summarized in Table 4-1. Revision 4 of the WAC also
describes the relationship among the various DOE guidance documents that address
characterization of TRU wastes, including the WIPP TRU Waste Characterization Quality
Assurance Program Plan (TRU QAPP) (DOE94b) and the generator and/or storage site
Quality Assurance Project Plans (QAPjPs). However, as discussed below, the WAC
document is outdated and is not integrated with the TRU QAPP.
The Waste Acceptance Criteria Certification Committee (WACCC) is responsible for
developing the WIPP WAC and verifying compliance of TRU waste with the WIPP WAC at
the generator/storage facilities. According to DOE, compliance will be demonstrated
through audits and surveillances at waste generators (DOE91). It should be noted that WAC
1 Many standard analytical protocols have been modified to accommodate the analytical and radiological
aspects of analyzing materials heavily contaminated with Pu-239.
4-2
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Table 4-1. Summary of Waste Acceptance Criteria and Requirements1
WAC Criterion/
Requirement &
Section
CH
or
RH
WIPP
Operations and
Safety Criteria
Transportation: Waste
Package Requirements:
TRAMPAC/RH-Cask2
RCRA
Requirements
Performance
Assessment
Criteria
Waste Container Requirements/Criteria
Waste Containers
3.2.1
Waste Package
Size
3.2.2
Waste Package
Handling
3.2.3
CH
RH
CH
RH
CH
RH
Type A,
Noncombustible
Type A,
Noncombustible
Maximum
dimension
specified
RH Canister
Drum and Box
Handling
Attachments
Axial Pintle
55-gal drums, SWBs,
or SWB Overpack of
55-gal Drums or Test
Bin
RH Canister
55-gal Drums in Two
Seven Packs, of Two
SWBs
RH Canister
Seven Packs, or SWBs
Axial Pintle
No Additional
Requirements
No Additional
Requirements
None
None
None
None
Same as
Transportation
None
Same as
Transportation
None
No Additional
Requirements
None
Waste Form Requirements/Criteria
Immobilization
3.3.1
Liquids
3.3.2
Pyrophoric
Materials
3.3.3
Explosives and
Compressed
Gases
3.3.4
CH
&
RH
CH
&
RH
CH
&
RH
CH
&
RH
< 1% Below 10
Microns,
< 15% Below
200 Microns
Only Residual
Liquids (see
definitions in
Section 3.3.2.1)
< 1%
Radionuclides,
No Non-
Radionuclide
Pyrophorics
Not Permitted,
49 CFR 173
Subpart C and G
None
< 1 Volume Percent
< 1% Radionuclides,
No Non-Radionuclide
Pyrophorics
Explosives and
compressed gases are
not permitted
No Additional
Requirements
No Additional
Requirements
See Section
3.3.5.3
See Section
3.3.5.3
Same as WIPP
Operation
< 1 Volume
Percent
Same as
Transportation
No Additional
Requirements
4-3
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Table 4-1. Summary of Waste Acceptance Criteria and Requirements (Continued)
WAC Criterion/
Requirement &
Section
CH
or
RH
WIPP
Operations and
Safety Criteria
Transportation: Waste
Package Requirements:
TRAMPAC/RH-Cask2
RCRA
Requirements
Performance
Assessment
Criteria
Waste Form Requirements/Criteria (Cent.)
TRU Mixed
Wastes
3.3.5
Specific Activity
of Waste
3.3.6
CH
&
RH
CH
RH
Hazardous
Waste must be
Reported
> 100 nd/g
TRU
> lOOnd/g
TRU
< 23 Ci/liter
total
Corrosives are not
permitted
Same as WIPP
Operations
Same as WIPP
Operations
WIPP RCRA
parts A & B
Permit
Applications,
WAP, NMD
None
None
No Additional
Requirements
Same as WIPP
Operations
Same as WIPP
Operations
Waste Package Requirements/Criteria
Waste Package
Weight
3.4.1
Nuclear
Criticality
(Pu-239 FGE)
3.4.2
Pu-239
Equivalent
Activity
3.4.3
Surface Dose
Rate
3.4.4
CH
RH
CH
RH
CH
&
RH
CH
RH
< 21,000 Ibs
< 8, 000 Ibs
See List in
3.4.2.1
< 600 g
< 1000 PE-Cif
package
< 200 mrem/hr
95% < 100
rem/hr.
5% < 1000
rem/hr.
1000 Ibs per drum,
4000 Ibs per SWB,
7265 Ibs per
TRUPACT-H payload,
19,250 Ibs per
TRUPACT-II,
80,000 Ibs GVW (DOT)
RH-Cask TBD
< 200 g/drum
< 325 g/SWB, or
< 325 g/TRUPACT-II
< 325 gi 'cask
None
< 200 mrem/hr, DOT
Package Limits, and
Shielded Packages per
SARP
RH-Cask TBD and
DOT Package Limits
None
None
None
None
None
None
None
None
None
Same as
Transportation
Same as
Transportation
None
Same as WIPP
Operations
None
4-4
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Table 4-1. Summary of Waste Acceptance Criteria and Requirements (Continued)
WAC Criterion/
Requirement &
Section
CH
or
RH
WIPP
Operations and
Safety Criteria
Transportation: Waste
Package Requirements:
TRAMPAC/RH-Cask2
RCRA ;:
Requirements
Performance;
^ : Assessment: :
"A:;- ^Criteria':;,:";
.-•- "-•-.;; • -\:-. . •' '
Waste Package Requirements/Criteria (Cont. ) : ; '••;•: : ; ;
Removable
Surface
Contamination
3.4.5
Thermal Power
3.4.6
Gas Generation
3.4.7
Labeling
3.4.8
CH
&
RH
CH
RH
CH
RH
CH
RH
< SOpCi/100
en? alpha,
< 450pCi/100
en? beta-gamma
No Limit Report
if > 0.1
watts/ft3
< 300
watts/canister
Vented
Vented
Id Number,
DOT
Id Number,
DOT
None
Refer to Limits in
TRUPACT-II SAR
Section 1.2.3.3
RH-Cask TBD
TRAMPAC Limits: See
requirements in Section
3.4.7.2, < 500 ppm
Flammable VOCs;
Chemical compatibility
study; all trace
chemicals < 5 weight
percent
RH-Cask TBD
Id Number and Waste
Shipping Category
RH-Cask TBD
None
None
None
NMD
Requirements
Apply
None
Same as DOT
TBD
Same as WIPP
Same as
Transportation
None
SNL Test Plan
Same as CH
None
None
Data Package! Requirements/Criteria I : ;
Data Package/
Certification
3.5.1
CH
RH
Certification,
WWIS
Information,
Data Format
Certification,
WWIS
Information,
Data Format
Tables 13.1 to 13.3 in
Appendix 1.3. 7
(TRAMPAC)
RH-Cask TBD
Hazardous
Waste
Manifest per
40 CFR part
262 NMD
andQAPP
Requirements
TBD
PA Data
Package,
QAPP
Requirements
None
4-5
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Table 4-1. Summary of Waste Acceptance Criteria and Requirements (Continued)
WAC Criterion/
Requirement &
Section
CH
or
RH
WIPP
Operations and
Safety Criteria ;
Transportation: Waste
Package Requirements :
TRAMPAC/RH-Cask2
RGRA
Requirements
Performance
Assessment
Criteria •
Other Requirements/Criteria
Additional
Requirements
3.6.1
CH
RH
None
None
One Shipping Category
per TRUPACT-II,
Authorized TRUCON
Content Codes, Waste
Aspirated per SARP,
Payload Control
Procedures
RH-Cask TBD
Regulations
or Permit
Conditions as
Determined
by NMED
TBD
None
None
1 Limiting parameters are shown in bold italics.
2 - RH Cask limits have been finalized.
Source: Waste Acceptance Criteria for the Waste Isolation Pilot Plant (DOE91)
audits have not been performed at waste generator sites for the last two years. Historically,
the WIPP WAC has been disconnected from waste characterization activities at the generator
sites, precluding a prospective incorporation of WAC related requirements in generator site's
ongoing waste generation practices. DOE plans to integrate the WAC in site waste
generation activities in support of its greater reliance on process knowledge as the main
waste characterization tool for newly generated wastes (DOE94d). The TRU waste generator
sites differ in their individual approaches to the generation and characterization of TRU
waste.
4.1.1.2 WIPP TRU QAPP
DOE released Revision B of the TRU QAPP in July, 1994. This document replaced the
Waste Characterization Program Plan for WIPP Experimental Waste and the Quality
Assurance Program Plan for the WIPP Experimental Waste Characterization Program. In
the TRU QAPP, DOE
identifies the quality of data necessary, and techniques designed to attain and
ensure the required quality, to meet the specific objectives associated with the
Department of Energy (DOE) Waste Isolation Pilot Plant (WIPP) Transuranic
Waste Characterization Program (DOE94b).
4-6
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This document provides guidance for the TRU waste generator sites in developing their site-
specific QAPjPs. The QAPjPs contain detailed information regarding how the site will
achieve the data quality objectives (DQOs) for the various waste characterization techniques.
It is worth noting that neither this document nor the DOE Carlsbad Area Office's (CAO)
guidance on the use of acceptable knowledge provides DQOs for waste characterization
performed using acceptable knowledge (DOE94b, DOE95c).
4.2 REGULATORY DRIVERS FOR WASTE CHARACTERIZATION
This section briefly summarizes the various laws, regulations, and agreements which underlie
the WIPP WAC and the specific waste characterization requirements which are distilled from
these sources. The sources discussed include:
• Agreement for Consultation and Cooperation between DOE and the State of
New Mexico
• The WIPP Land Withdrawal Act (LWA)
• NRC regulations for the packaging and transportation of radioactive waste (10
CFR part 71)
• Department of Transportation regulations
• Resource Conservation and Recovery Act and amendments
• The Federal Facilities Compliance Act (FFCA) of 1992
• EPA Toxic Substance Control Act regulations
EPA 40 CFR part 191
• EPA WIPP compliance criteria (40 CFR part 194)
From the summary it will be clear that waste characterization, in various forms, is required
not only to satisfy 40 CFR parts 191 and 194, but a variety of other regulations and
agreements as well.
In certain instances, the regulatory framework separates the radioactive waste into two
categories: 1) contact-handled transuranic waste (CH-TRU) and 2) remote-handled
transuranic waste (RH-TRU). Definitions and restrictions applicable to each type of
radioactive waste are presented in the ensuing discussion, where appropriate.
4-7
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4.2.1 Agreement for Consultation and Cooperation—July 1, 1981
An Agreement for Consultation and Cooperation (the Agreement) between the State of New
Mexico and the U.S. Department of Energy was signed by the parties on July 1, 1981.
Appendix B to this Agreement is entitled Working Agreement for Consultation and
Cooperation (the Working Agreement). Article IV of the Working Agreement provides a
basis for the State to comment on waste acceptance criteria as described in IV.E.l.(c):
DOE has provided this documentation to the State. Any State comments as to
public health and safety concerns shall be provided to the DOE WIPP Project
Manager within 2 calendar days after receipt of documentation from DOE.
DOE shall respond to the State comments within 2 calendar days after
receipt of such comments. Nothing herein shall preclude further discussions of
the matter or any updates prepared by DOE. Reasonable time frames for State
comments and DOE response to any DOE updates shall be negotiated by the
principal representatives of the parties.
The Agreement and the Working Agreement were modified in November 1984 under the
First Modification to the July 1. 1981 "Agreement for Consultation and Cooperation" on the
WIPP bv the State of New Mexico and the U.S. Department of Energy. Article VI.B of the
Agreement was revised to set certain limitations on remote-handled transuranic waste
including the following maximum values for specified parameters:
• volume — 250,000 cubic feet3
• surface dose rate — 1,000 rem/h
• volume with surface dose greater than 100 rem — 12,500 cubic feet
• activity level (averaged over canister volume) — 23 Curies (Ci)/l
• amount of radioactivity — 5.1 million Ci
The First Modification further specified that the concentrations of radionuclides in the
RH-TRU canisters would be determined by one or more of the following methods: "(1)
materials accountability; (2) classification by source; (3) gross radioactivity measurements;
(4) direct measurements of major contributing radionuclides; or (5) such other methods as the
parties may agree to."
2 To be negotiated in original agreement.
3 This is 4% of the total waste volume.
4-8
-------
A second modification to the Agreement was implemented on August 4, 1987 which
included, among other things, amendment of Article VI.E to contain the following
paragraph:
"4. The transportation of radioactive waste to WIPP shall comply with the applicable
regulations of the U.S. Department of Transportation and any applicable
corresponding regulations of the U.S. Nuclear Regulatory Commission. All waste
shipped to the WIPP will be shipped in packages which the Nuclear Regulatory
Commission has certified for use."
4.2.2 WIPP LWA
The WIPP LWA was signed into law on October 30, 1992. Several items in the LWA relate
to waste characterization including relevant definitions and limitations (particularly those
involving RH-TRU waste). The following definitions from Section 2 of the LWA are
important to waste characterization:
"(20) TRANSURANIC WASTE — The term "transuranic waste" means waste
containing more than 100 nanocuries of alpha-emitting transuranic isotopes per gram
of waste, with half lives greater than 20 years, except for—
(A) high-level radioactive waste
(B) waste that the Secretary has determined with the concurrence of the
Administrator, does not need the degree of isolation required by the disposal
regulations; or
(C) waste that the Nuclear Regulatory Commission has approved for disposal
on a case-by-case basis in accordance with part 61 of title 10, Code of Federal
Regulations."4
"(3) CONTACT-HANDLED TRANSURANIC WASTE — The term "contact-handled
transuranic waste" means transuranic waste with a surface dose rate not greater than
200 millirem per hour."
4 The apparent intent of exceptions (B) and (C) is to preclude shipment to the WIPP of wastes which meet
the transuranic waste definition, but can be properly disposed in other than a geologic repository (e.g., greater
than Class C wastes (as defined in §61.55)).
4-9
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"(4) REMOTE-HANDLED TRANSURANIC WASTE — The term "remote-handled
transuranic waste" means transuranic waste with a surface dose rate of 200 millirem
per hour or greater. "5
Section 7 of the LWA imposes the following waste-related limitations:
• Restrictions of remote-handled transuranic waste (RH-TRU)
1,000 rem/h maximum surface dose rate
surface dose rate less than 100 rem/h for 95% by volume of all
RH-TRU
Canister activity limited to 23 Ci/liter (averaged over the canister
volume)
Total RH-TRU radioactivity is limited to 5.1 x 106 Ci
• Repository capacity — 6.2 million cubic feet of transuranic waste
Most of the waste requirements in Section 7 are also included in the First Modification to the
Agreement for Consultation and Cooperation between DOE and the State of New Mexico
(Section 4.2.1 above).
In Section 12, Congress made clear its intent that disposal at the WIPP be limited to TRU
wastes by prohibiting the shipment and disposal of high-level radioactive waste or spent
nuclear fuel.6 In Section 16, it further specified that the TRU waste must be shipped to the
WIPP in containers whose design is certified by NRC and whose QA requirements meet
NRC standards.
5 According to the LWA definitions of CH-TRU and RH-TRU, waste with a surface dose of exactly 200
millirem per hour meets both definitions.
6 These terms are defined in Section 2 of the Nuclear Waste Policy Act of 1982 as follows:
12. "High-level radioactive waste": (A) The highly radioactive material resulting from the reprocessing of
spent nuclear fuel, including liquid waste produced directly in reprocessing and any solid material derived from
such waste that contains fission products in sufficient concentrations; and (B) other highly radioactive material
that the Commission, consistent with existing law, determines by rule requires permanent isolation.
23. "Spent nuclear fuel" means fuel that has been withdrawn from a nuclear reactor following irradiation, the
constituent elements of which have not been separated for reprocessing.
4-10
o
-------
4.2.3 NRC Regulations for the Packaging and Transportation of Radioactive Waste (10
CFR part
In 10 CFR part 71, the NRC sets "(1) requirements for packaging, preparation for shipment,
and transportation of licensed material; and (2) procedures and standards for NRC approval
of packaging and shipping procedures for fissile material and for a quantity of other licensed
material in excess of a Type A quantity"7 (§71.0). Under this rule, packages must be
approved for each specific use. Subpart D of the rule defines the contents of the application
for approval of a transportation package. (In the case of WIPP, the application approval
package is the Safety Analysis Report (SAR) for the TRUPACT-II Shipping Package.
Revision 0 was issued in February 1989. The latest revision is No. 14 which was issued in
October, 1994.) The package description in the approval application must include the
following information with regard to the contents of the shipping package (§71.33):
• Identification and maximum radioactivity of the radioactive constituents
• Identification and maximum quantities of fissile constituents
• Chemical and physical form
• Extent of reflection, the amount and identity of nonfissile materials used as
neutron absorbers or moderators, and the atomic ratio of moderator to fissile
constituents
• Maximum normal operating pressure
• Maximum weight
• Maximum amount of decay heat
• Identification and volumes of any coolants
The DOE shipping package application for the TRUPACT-II Shipping Package has been
assigned Docket No. 71-9218 by the NRC who issued a Certificate of Compliance No. 9218
7 A Type A quantity is an amount of radioactive material which does not exceed certain isotope-specific
limits stipulated in Appendix A of 10 CFR Part 71.
4-11
-------
(DOE93a) for use of this container to ship CH-TRU.8 Revision 5 (June 9, 1994) of this
certificate specifies the following limitations on the contents of the TRUPACT-II based on
the items from §71.33 listed above:
"Dewatered, solid or solidified transuranic wastes. Waste must be packaged in
55-gallon drums, standard waste boxes (SWB) or bins. Waste must be restricted to
prohibit explosives, corrosives, nonradioactive pyrophorics, and pressurized
containers. Within a drum, bin, or SWB radioactive pyrophorics must not exceed 1
percent by weight and free liquids must not exceed 1 percent by volume. Flammable
organics are limited to 500 ppm in the headspace of any drum, bin, or SWB."
"Contents not to exceed 7,265 pounds including shoring and secondary containers,
with no more than 1,000 pounds per 55-gallon drum and 4,000 pounds per SWB."
"Fissile material not to exceed 325 grams Pu-239 equivalent with no more than
200 grams Pu-239 equivalent per 55-gallon drum and 325 grams of Pu-239 equivalent
per SWB."
"Decay heat must not exceed values specified in Tables 6.1 through 6.3 of
"TRUPACT-II Content Codes," (TRUCON), DOE/WIPP 89-004, Rev. 6."
"Physical form, chemical properties, chemical compatibility, configuration of waste
containers and contents, isotopic inventory, fissile content, decay heat, weight and
center of gravity, radiation dose rate must be limited in accordance with Appendix
1.3.7 of the application, "TRUPACT-II Authorized Methods for Payload Control,"
(TRAMPAC)."
"Each drum, bin, or SWB must be assigned to a shipping category in accordance with
Table 5, "TRUPACT-H Content Codes," (TRUCON), DOE/WIPP 89-004, Rev. 6, or
must be tested for gas generation and meet the acceptance criteria in accordance with
Attachment 2.0 of Appendix 1.3.7 of the application."
As noted above, the Certificate of Compliance specifies that waste properties are determined
and limited according to the specifications in TRAMPAC. TRAMPAC (Appendix 1.3.7 to
the SAR) is the document which provides acceptable methods for the preparation and
As of the date of publication of this document, the most recent revision to the TRUPACT-II Shipping
Package Application is Revision 14 submitted to NRC by Westinghouse (on behalf of DOE) on October 14,
1994. The current revision and revision date for other related documents are as follows:
• TRUPACT-II Content Codes (TRUCON) - Revision 8, October 1994
• TRUPACT-II Safety Analysis Report (SAR) Revision 14, October 1994
• Certificate of Compliance No. 9218 - Revision 6, March 30, 1995
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characterization of payloads for transport in TRUPACT-II. Parameters for which
TRAMP AC specifically identifies restrictions are as follows:
• Physical and chemical form of the CH-TRU waste
• Chemicals to ensure chemical compatibility between all constituents in a given
shipment
• Maximum pressure in a package during a sixty-day transport period
• Amount of potentially flammable gases that might be present or generated in
the payload during a sixty-day transport period
• Layers of confinement (e.g., plastic bagging) in payload containers
• Fissile material content for individual payload containers and the total package
• Decay heat for individual payload containers and the total package
• Weight of the individual payload containers and the loaded TRUPACT-II
• Center of gravity for the payload assembly to be transported in TRUPACT-II
• Dose rate of individual payload containers, the total package, and three loaded
packages on a truck trailer
The forego ing discussion is specific to CH-TRU waste. Currently there is no approved
shipping container for RH-TRU waste. DOE plans call for RH-TRU to be shipped in the
RH-72B, which is a scaled down version (5/8 scale) of the NuPac 125B container certified
by NRC and used to ship waste from Three-Mile Island Unit 2 (DOE93a).
The TRAMP AC provides some detail on how various parameters are to be tested. For
example, Section 9.4, Methods of Determination and Control of Radionuclides. specifies five
allowable methods for the identification and quantification of radionuclides in TRU waste
including:
• passive gamma
• radiochemical assay using alpha and gamma spectroscopy
• passive neutron coincidence counting
• passive-active neutron assay
• calorimetry
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Attachment 3.0 to the TRAMP AC discusses each of the allowable methods including typical
errors, sensitivities, calibration standards, assay procedures, and operator training. These
topics are addressed further in the WIPP Transuranic Waste Quality Assurance Program Plan
(DOE94b) and the site specific Quality Assurance Project Plans.
4.2.4 U.S. Department of Transportation Regulations: 49 CFR part 173 — Shippers —
General Requirements For Shipments and Packaging
The Department of Transportation (DOT) has jurisdiction over hazardous materials shipments
affecting intrastate and interstate commerce (DOE93a). This authority is derived from the
Hazardous Materials Transportation Act of 1975 as amended by the Hazardous Materials
Transportation Uniform Safety Act of 1990. Subpart I of 49 CFR part 173 sets out DOT
regulations for the shipment of radioactive materials. Basically, the DOT regulations provide
that any package which meets the applicable requirements of NRC regulation 10 CFR part 71
is authorized for shipment (49 CFR 173.416(b)). The DOT regulations add no additional
waste characterization requirements beyond those already imposed by the NRC.
4.2.5 Resource Conservation and Recovery Act (RCRA)
The Resource Conservation and Recovery Act of 1976 (RCRA) and the Hazardous and Solid
Waste Amendments of 1984 (HSWA) provide the statutory framework for the regulation of
hazardous wastes at the WIPP. Under HSWA, certain "listed" and "characteristic
hazardous" wastes are prohibited from land disposal unless the wastes meet specified
treatment standards or it can be demonstrated to a reasonable degree of certainty that there
will be no migration of hazardous constituents from the disposal unit for as long as the
wastes remain hazardous. Migration of hazardous constituents outside the unit boundary
must not exceed health-based limits (EPA92). The approach being taken by DOE at the
WIPP is to seek a no migration variance rather than meet the technology-based treatment
standards.
Requirements of a petition to seek a no migration variance are set forth in 40 CFR part
268—Land Disposal Restrictions. Specific requirements (§268.6) which relate to waste
characterization are:
§268.6(a)(2) A waste analysis to fully describe the chemical and physical
characteristics of the subject waste [must be provided]
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§268.6(b)(l) All waste .... sampling, test, and analysis data must be accurate and
reproducible to the extent that state-of-the art techniques allow
§268.6(b)(2) All sampling, testing, and estimation techniques for physical and
chemical properties of the waste .... must have been approved by the Administrator
§268.6(b)(3) Simulation models must be calibrated for the specific waste ....
conditions and verified for accuracy by comparison with actual measurements.
The No Migration Guidance Manual for Petitioners (EPA92) elaborates on the waste analysis
dictated under §268.6(a)(2) noting that "proper management of wastes for as long as they
remain hazardous requires that potential incompatibilities and waste transformation
mechanisms be assessed." Some additional guidance provided in the Manual regarding
details of waste descriptions is summarized below:
• Waste types and sources
applicable waste codes (EPA and industrial)
waste-generating processes
hazardous constituents and their properties
quantities of waste to be disposed
rate of disposal
handling and storage practices
• Waste characteristics
potential for leachate formation
waste solubilities
hazardous-constituent vapor pressures
other factors affecting waste mobility
analytical testing results for 40 CFR part 261 Appendix VIII hazardous
constituents reasonably expected to be present in the waste
• Waste incompatibilities
potential chemical interactions
identification and characteristics of reaction products
• Waste transformation mechanisms
biodegradation
photodegradation
hydrolysis
oxidation/reduction
volatilization
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In 1990, EPA granted DOE a conditional no migration variance to permit DOE to implement
an underground test program with a limited quantity of actual TRU waste at the WIPP
(55 FR 47700). DOE subsequently canceled the test program so the no migration variance
was never exercised. However, some of the conditions imposed by EPA in this conditional
variance are instructive as presaging future EPA requirements when DOE seeks a final no
migration variance to dispose of TRU wastes in the repository.9 It is recognized that the
conditional variance was based on short-term no migration considerations over a ten-year test
phase with particular focus on air emissions. Thus, some of the conditions specified in
granting the variance may not be indicative of requirements for permanent disposal.
In granting the conditional variance, EPA imposed the following requirements relating to
waste analysis:
To ensure that each waste container had no layer of confinement which Contains
flammable mixtures of gases or mixtures of gases which could become flammable
when mixed with air, samples of gas from the head space in each container must be
analyzed for hydrogen, methane, and volatile organic compounds. It must also be
demonstrated that the headspace gas is representative of the gas within all layers of
confinement hi a container.
To ensure that the wastes to be emplaced are compositionally similar to the wastes on
which the no migration petition was based, representative samples of headspace gas
must be analyzed and compared to compositions supplied with the petition. If the
results are not comparable, the waste may not be shipped to WIPP (without treatment
or modification)
A key finding in the conditional no migration determination was that "if adequate data are
not collected, EPA will not be in a position to approve any no-migration petition for the
operational or post-closure phase." EPA clearly stated that further characterization of the
waste would be required before a final no migration petition could be considered by the
Agency. EPA noted that, at a minimum, wastes should be analyzed for 32 organic
compounds and six metals (Cd, Cr, Pb, Se, Hg, Ag). Testing should include headspace
analysis of all waste types for the organics and analysis of sludges for both organics and
metals.
4.2.5.1 RCRA part B Permit Application
Since New Mexico is authorized by EPA to permit facilities which treat, store and dispose of
radioactive mixed waste, the RCRA part B Permit Application must be submitted to the New
DOE submitted a draft petition to EPA for a disposal phase no-migration variance in May, 1995.
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Mexico Environment Department (NMED). In February 1991, DOE submitted a RCRA part
B Permit Application for the Test Phase and in May 1995 (Revision 5) for the disposal
phase.10
The draft part B Permit Application contains a Waste Analysis Plan which was prepared in
accordance with EPA guidance (EPA94). According to the Permit Application (Revision 5),
the following waste is unacceptable for management at the WIPP facility:
• Ignitable, reactive and corrosive waste (Free liquids, explosives, compressed
gases, oxidizers, and non-radioactive pyrophorics are prohibited.)
• Headspace gas volatile organic compounds (VOCs) in concentrations resulting
in emissions not protective of human health and the environment
• Incompatible wastes (Waste must be compatible with container, cask, and
TRUPACT II materials as well as other waste.)
• Compressed gases
• Free liquids (Residual liquids in well-drained containers must be less than 1 %
by volume.)
• Waste with 50 parts per million or more of polychlorinated biphenyls (PCBs)
• particulate waste not solidified, stabilized, or consolidated
• Wastes with EPA codes not listed in the RCRA part A permit application
The Waste Analysis Plan further specifies all waste containers (for both newly-generated and
retrievably-stored wastes) undergo headspace gas analysis for total VOC concentrations.
Based on results and trends DOE may propose in the future to reduce the sampling
frequency. Homogeneous solids and soil/gravel wastes will be periodically sampled for
VOCs, semi-volatile organic compounds, and metals. Debris wastes will be characterized on
the basis of acceptable knowledge rather than examination and/or assay. The physical form
of all retrievably-stored wastes will be determined by radiography or visual examination.
10 New Mexico's RCRA regulations (HWMR-7) mirror the Federal RCRA regulations.
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4.2.6 Federal Facilities Compliance Act of 1992 (Public Law 102-386)
The FFCA is an amendment to the Solid Waste Disposal Act (SWDA) (42 U.S.C 6981)
which, among other things, imposes certain restrictions on DOE regarding the storage of
mixed wastes.11 After October 6, 1995, DOE can continue to store mixed waste without
violation of Section 3004(j) of the SWDA only if a plan has been submitted to EPA, or to a
state agency authorized by EPA to regulate the hazardous components of the mixed waste,
and has been approved by the appropriate agency. An order requiring compliance with the
plan must also have been issued. According to Sec. 102 ® of the FFCA, the requirement
does not apply to facilities subject to existing agreement, permit, administrative, or judicial
order. For example, a tri-partite compliance agreement among DOE, EPA Region X, and
the State of Washington exists for the Hanford Site which takes precedence (DOE94a).
While the FFCA does not, per se, require waste characterization, the compliance plans may.
The FFCA does, however, require that DOE generate an inventory of mixed wastes. Some
of the specified elements of this inventory include:
• a description of each type of mixed waste including the name of the waste
stream.
• the EPA hazardous waste code for each type of mixed waste that has been
characterized at each DOE facility12
• an inventory of each type of waste that has not been characterized by sampling
and analysis at each DOE facility
• the basis of DOE's determination of the applicable hazardous waste code for
each type of mixed waste and a description of whether the determination is
based on sampling and analysis or on process knowledge
The FFCA also requires that DOE develop and submit Site Treatment Plans for the
development of treatment capacity and technologies for handling mixed waste. Required
inventory reports and plans are described in Section 3021 of the FFCA. Mixed waste
11 Mixed wastes are wastes which contain a hazardous component regulated under the Resource
Conservation and Recovery Act and a radioactive component regulated under the Atomic Energy Act.
i: EPA Hazardous Waste Codes are found in 40 CFR parts 260 270.
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inventory reports have been completed (DOE94c) and Draft Site Treatment Plans have been
summarized in a recent DOE report (DOE94a). The National Summary Report (DOE94a)
noted that about one-third of the existing mixed TRU waste can probably be shipped to the
WIPP without further treatment, but the balance will require additional treatment to meet the
expected waste acceptance criteria. Thus, at least implicitly, the FFCA requirements will
result in increased understanding of the characteristics of the waste destined for the WIPP.
Existing and proposed DOE facilities to treat mixed TRU waste are as follows:
• Existing Facilities
Idaho National Engineering Laboratory
Rocky Flats Environmental Technology Site
Argonne National Laboratory East
West Valley Demonstration Project13
• New Facilities
Hanford Site
Argonne National Laboratory West
Idaho National Engineering Laboratory
Nevada Test Site
Rocky Flats Environmental Technology Site
Oak Ridge Reservation
Savannah River Site14
Comments derived from information contained in the Site Treatment Plans for several of
these mixed TRU treatment facilities are noted below.
Rocky Flats Environmental Technology Site (RFETS)
RFETS estimates the following distribution of mixed TRU wastes (RFP94):
• Meets WIPP WAC and TRAMP AC — 52.4%
• Test and possibly repackage for TRAMP AC — 30%
• Immobilize for WIPP WAC — 5.4%
13 TRU wastes at West Valley are not defense related and therefore are not slated for disposal at the WIPP.
14 According to DOE94a, Vol II, the Savannah River Site has deferred treatment until more definitive
information is available regarding the WIPP WAC.
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• Neutralize for WIPP WAC — 4.4%
• Oxidize for WIPP WAC — 1.2%
• Incomplete data — 6.6%
According to the draft site treatment plan, RFETS proposes to construct a facility which will
include capabilities for repackaging, immobilization, neutralization, and oxidation. This
facility is planned for operation in 2008.
As noted above, approximately 30% of the RFETS waste will require testing to determine
whether the gas-generation requirements of TRAMP AC (see Section 4.2.3) will be met.
Nevada Test Site (NTS)
Mixed TRU waste at the Nevada Test Site (NTS) was shipped there from Lawrence
Livermore Laboratory between 1974 and 1990 (NTS94). Since this waste is poorly
characterized and may be in oversized packaging, NTS is proposing to construct a TRU
Waste Certification Building, which, if funded, would be operational in FY 1999.
Operations will include breaching, sampling, and repackaging waste, and certifying that the
containers meet the WIPP waste acceptance criteria.
Idaho National Engineering Laboratory (INEL)
INEL has identified 52 waste streams, some portion of which will require treatment to meet
the WIPP WAC (IDA94). Facilities proposed to handle these projected needs include the
Remote Mixed Waste Treatment Facility (RMWTF), the Idaho Waste Processing Facility
(IWPF), and several Generator Treatment Plan (GTP) sites to handle small waste volumes.
The IWPF is designed to include the following treatment technologies: stabilization,
amalgamation, sizing, and incineration. The RMWTF is designed to handle RH- and
CH-TRU wastes containing reactive metals.
Argonne National Laboratory East (ANL-E)
ANL-E has a few cubic meters of acidic waste water which must be treated in a proposed
precipitation/filtration unit prior to shipment to WIPP (ANL94). The wastewater will be
neutralized, heavy metals will be precipitated, and residual sludge will be stabilized. Other
ANL-E waste streams can be shipped without treatment.
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Oak Ridge National Laboratory (ORNL)
ORNL is proposing the construction of a Waste Handling and Packaging Plant (WHPP) to
handle five of its six waste streams (OAK94). The sixth waste stream is subject to CERCLA
action under an existing agreement involving the State of Tennessee. The WHPP would
contain a sludge mobilization facility which would fluidize waste from storage tanks and
transfer it to the processing facility. In the processing facility, which consists mainly of a
bank of hot cells, wastes will be remotely dried, assayed, packaged, and checked for
contamination. Hot cell operation is dictated by the fact that a significant fraction of the
ORNL wastes are RH-TRU. Start up tests for the WHPP are projected for 2005.
Other DOE Locations
Draft Site Treatment Plans at other TRU waste generator sites such as Lawrence Livermore
and Los Alamos National Laboratories, the Savannah River Site, and the Mound Plant are
much less specific as to planned actions.
4.2.7 TSCA: 40 CFR part 761— PCB Manufacturing. Processing. Distribution in
Commerce and Use Prohibitions
Unlike the RCRA regulations, the TSCA regulations do not provide for the issuance of no
migration variances. Thus, waste containing PCBs must be treated to meet TSCA
requirements before disposal (IDA94). Generally speaking, §761.60—Disposal
requirements—specifies that PCBs at concentrations of 50 ppm or greater must be treated in
a licensed incinerator. Alternate methods of disposal which achieve the same level of
performance in destroying PCBs as incinerators may be approved by EPA (§761.60(e)).
The draft site treatment plans prepared by INEL and Rocky Flats have noted that PCB-
contaminated TRU waste at those facilities must be treated (IDA94 and RFP94). INEL
states that wastes "will be treated to meet TSCA requirements" while Rocky Flats says that
"PCBs must be destroyed or oxidized to meet WTPP WAC."
As discussed above in Section 4.2.5.1, Revision 5 of the WIPP RCRA part B Permit
Application prohibits "waste with equal to or more than 50 parts per million (50 milligrams
per liter) polychlorinated biphenyls (PCBs)." The Waste Analysis Plan indicates that
transformer oils containing PCBs have been identified in a few waste streams included in the
organic sludges summary category and consequently these streams must be examined for
PCBs.
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Revision 1 of the WIPP Baseline Inventory Report (WTWBIR) states that 13 TSCA waste
streams cannot be accepted at the WIPP under the terms of the draft RCRA part B Permit
Application and are consequently excluded from the WTWBIR (DOE95a).
4.2.8 Environmental Protection Standards for Management and Disposal of Spent Nuclear
Fuel. High-Level and Transuranic Radioactive Wastes (40 CFR part 191)
Subpart A, Environmental Standards for Management and Storage, of this rule sets annual
dose equivalent exposure standards for the maximum off-site individual during facility
operation as follows (50 FR 38085):
whole body — 25 mrem
thyroid — 75 mrem
other critical organ — 25 mrem
These standards, coupled with DOE Order 6430.1-General Design Criteria, were used as the
basis for setting the upper limit on TRU waste packages received at the WIPP at 1,000
Curies of Pu-239 equivalent activity15 (DOE87). Inhalation dose calculations are based on
particles having a 1 fim Activity Median Aerodynamic Diameter. Assumed accident
scenarios set the particle size distribution for drum handling mishaps which, in turn, lead to a
particle size specification in the WIPP WAC. Wastes not meeting the particle size
specification mat require treatment prior to shipment to the WIPP. Distribution of drums of
waste with high curie contents may be important in analyzing releases from drilling
intrusions.
Subpart B, Environmental Standards for Disposal, and Subpart C, Environmental Standards
for Ground-Water Protection, of the amended rule (58 FR 66414) prescribe the long-term
containment requirements which the WIPP must meet and defines performance assessment as
the basis for assessing compliance with the cumulative release limits in Subpart B.
Performance assessment will establish, through iterative calculations, an envelope of waste
acceptance criteria which, if met, should provide a reasonable expectation that the disposal
standards can be achieved for the regulatory life of the repository.
15 Pu-239 equivalent curies are used to normalize the inhalation hazard of various transuranic nuclides to
that posed by Pu-239.
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4.2.9 Criteria for the Certification and Re-certification of the Waste Isolation Pilot Plant's
Compliance with the 40 CFR part 191 Disposal Regulations (40 CFR part 194)
The WIPP LWA orders EPA to promulgate, through a formal rulemaking process, the
criteria which the Agency would use to assess DOE's compliance with the 40 CFR part 191
disposal standards at the WIPP. §194.24 of the rule deals with waste characterization and
requires DOE to identify the chemical, radiological, and physical characteristics of all
existing waste, and to the extent practicable, to-be-generated waste, proposed for disposal at
the WIPP. DOE can use process knowledge, non-destructive examination/assay, and other
methods to provide this waste description.
DOE is further required to substantiate that all waste characteristics which could impact
containment of wastes by the disposal system have been identified and their impact assessed.
Waste characteristics include, inter alia, radionuclide solubility, ability of radionuclides to
exist in stable colloidal suspensions, gas generation potential, and shear strength. DOE must
also substantiate that all waste components which influence the critical waste characteristics
are identified and their impact assessed. Waste components include, but are not limited to,
such items as the activity of each radionuclide, metals, cellulosics, cheating agents, and
water and other liquids.
Using this information, DOE is required to set limits on those waste components judged to
be important and to show that, when all of these components are set at the designated
limits,16 the disposal system will meet the numeric requirements of §191.34 and §194.55. It
is then incumbent on DOE to ensure that the waste actually emplaced in the WIPP falls
within these limits.
4.3 IMPACTS OF WASTE CHARACTERISTICS AND COMPONENTS ON
PERFORMANCE ASSESSMENT
Generally speaking, waste characteristics are determined through laboratory and fields studies
or through literature assessment combined with technical judgement. Waste components tend
to be measured on an on-going basis. As discussed in the following paragraphs, both waste
characteristics and waste components can affect performance assessment.
16 In some cases, the upper limit on a component will produce the more conservative result while in other
cases the lower limit will be controlling. For example, solubility of actinide elements generally increases as the
pH of the solution is lowered. Thus one would want to specify the minimum quantity of components which
would tend to increase pH.
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4.3.1 Waste Characteristics
Waste characteristics can be broadly divided into four categories according to what they
affect: mobility of actinide elements in solution, strength, fluid flow, and gas generation in
the sealed repository (SAN92). These categories are discussed in general terms in this
section and in more detail in sections 4.3.5 and 4.3.6.
4.3.1.1 Mobility in Solution
It is expected that under certain conditions the WIPP wastes will be exposed to brines.
These brines can be the result of seepage of Salado Formation brines through the repository
walls, seepage of brines from the overlying Rustler Formation through poorly sealed shafts
or boreholes, or from flow of Castile Formation brines released by an inadvertent borehole
(or boreholes) into the waste-filled rooms of the repository. The quantity of brine to which
the waste is exposed is dependent on several factors including the stage of the creep closure
of the room, the source of the brine, capillary effects, and the gas pressure in the room. The
brine can mobilize the actinide elements in the waste by two mechanisms—solubility and
formation of stable colloids. Solubility of the actinides is a complex function of brine
strength, pH, oxidation state of the actinide species, carbon dioxide levels, and presence of
organic ligands which can form soluble complexes with the actinides. Conditions for the
formation of various types of stable colloids are still being examined in the laboratory. Once
the actinide elements are mobilized, there are several conceptual mechanisms available by
which they can be transported to the accessible environment. If the actinide elements are not
mobilized in the brine, the only mechanism available for release from the disposal system is
via waste-laden material brought to the surface as the result of inadvertent drilling.
4.3.1.2 Waste Strength
Waste strength enters into performance assessment calculations in several ways. The
crushing resistance of the waste provides a back stress which opposes the creep closure of
the bedded salt surrounding the waste and consequently slows the closure process. Room
collapse, which is in part linked to crushing resistance, determines the porosity in the waste
as a function of time. This porosity, in turn, is used in equations to calculate the flow of
brine through the waste. In addition to the crushing resistance (i.e., volumetric plasticity as
a function of pressure), other waste parameters needed for the constitutive equations used to
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calculate the waste response to stress are shear modulus, bulk modulus, and yield function
constants. The constitutive parameters have been assumed based on educated guesses.
While SNL believes these parameters are of secondary importance, they have recommended
that bounding studies be conducted using extreme values of these parameters to provide an
indication of disposal room response (LAB95). The crushing resistance has been obtained
from laboratory experiments using simulated waste mixtures. From these experiments, a
composite repository-wide curve of mean stress versus volumetric strain was developed based
on an assumed waste weight mix of 28% metals (including the container), 28% combustibles,
and 44% sludges (LAB95).17 This curve was used in the 1992 WIPP Performance
Assessment (PA) (SAN92, vol. 3, p. 2-71). Conceptually, the waste mix fits the definition
of a waste component which influences a waste characteristic—the crushing resistance.
Assessment of the response of the room to collapse also requires knowledge of the initial
waste porosity. This waste characteristic can also be derived from the densities of the
components which make up the waste.
The shear strength of the waste is needed to analyze the amount of waste which might be
eroded from the wall of an intruding borehole by the action of the drilling fluid. Depending
on the type of analysis performed, the tensile strength of the waste may also be needed to
assess the amount of spallation which occurs in a borehole due to gas pressure release within
the waste.
4.3.1.3 Gas Generation Within the Waste
Several mechanisms have been identified which can cause significant quantities of gas to be
generated by the wastes after emplacement (BRU94). The principal gas generation
mechanisms are related to the anoxic corrosion of certain metals and the anaerobic microbial
degradation of cellulosics and other organic compounds. (Oxygen initially present when the
disposal rooms are sealed is consumed in a reasonably short period producing an oxygen-free
environment.) Quantities of gas produced by radiolysis and release of volatile organic
compounds are small by comparison.
The anoxic corrosion of ferrous metals requires the presence of water, which is consumed in
the reaction while hydrogen is produced. This water can be initially present in the waste,
17 Based on information in DOE95a, the current weight fractions are 0.59 solid organics, 0.13 solid
inorganics, and 0.28 sludges.
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brine which seeps into the disposal room from the surrounding formations, and/or brine
which is released by an intruding borehole from a reservoir in the underlying Castile
Formation. For gas generation to proceed at a significant rate, the ferrous metals must be
inundated by water. The rate is reduced by three to four orders of magnitude when exposure
is limited to water vapor. Aluminum and its alloys can be similarly involved in anoxic
corrosion also producing hydrogen.
Microbial degradation of cellulosics and, perhaps, plastics and rubber, can produce a variety
of gases including hydrogen, methane, hydrogen sulfide, carbon dioxide, and nitrous oxide.
For this to happen, the following conditions must all be met:
• the microorganisms are present when the repository is sealed
• the microorganisms persist for a significant fraction of the 10,000-year
repository life
• adequate water is present
• sufficient oxidants are present
• sufficient nutrients such as P and N are available
If the gas pressure generated by these mechanisms exceeds the lithostatic pressure of the
surrounding rock formation (i.e., about 14.8 MPa or 150 atmospheres), several disposal
system responses are possible. The relatively brittle anhydrite interbeds above and below the
repository horizon could fracture providing enhanced pathways for transport of radionuclides
to the accessible environment, the creep closure process could be reversed, and/or brine
could be driven from the disposal rooms causing the gas-producing reactions to cease.
Recent work has shown that gas spallation processes can cause significant quantities of waste
to be transported to the surface from an intruding borehole. These processes become
significant when the pressure in the waste exceeds the fluid pressure of the drilling mud at
the base of the borehole (about 8MPa).
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4.3.1.4 Fluid Flow
.SNL uses the computer code BRAGFLO to model two-phase flow in various regions of the
repository. The mass balance equations in BRAGFLO employ effective permeability k,
which is the product of the intrinsic permeability and the relative permeability of the zth
phase (i.e., gas or water). In the 1991 and 1992 WIPP performance assessments, the
intrinsic permeability of the waste was set at IxlO"13 m2 based on some experimental work
with simulated waste (SAN92). The relative permeabilities of the gas and the liquid were
derived from empirical composite curves based on measurements in many porous materials
such as sand, sandstone, and clay as a function of liquid saturation (i.e., the amount of pore
space in the waste occupied by liquid at any point in time). These empirical curves require
assumptions as to the residual brine saturation, the residual gas saturation and a pore shape
distribution parameter. In addition, the BRAGFLO equations also require specification of the
capillary pressure which is assumed, based on an empirical relationship, to be a function of
intrinsic permeability and a factor which reflects parametric uncertainty. Since no WIPP
waste-specific data exist for capillary pressure or relative permeability, a high degree of
parametric uncertainty exists for waste-related flow properties.
4.3.2 Waste Components
Waste components can be generally divided into those which influence certain waste
characteristics thus indirectly influencing PA, and those which directly influence performance
assessment. The former category would include such items as quantities of gas generating
materials, physical waste composition (i.e., waste volume mix), and quantities of constituents
affecting waste mobility (e.g, organic ligands). The total quantity of various actinide
elements present in the waste will govern whether the amount of the actinide species
mobilized in the waste is limited by the solubility of the element in intruding brines or by the
total inventory of the element in the waste. Waste components which directly influence
performance assessment generally relate to the quantities of radioactivity hi the waste (i.e.,
its curie content).
The curie content of waste enters explicitly into performance assessment calculations in two
ways. First, it is used to set the release limits hi accordance with Table 1 of 40 CFR part
191. For TRU radionuclides, the release limits hi Table 1 are based on one million curies of
alpha-emitting transuranic radionuclides with half-lives greater than 20 years. Thus, if the
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WIPP repository hypothetically contained 5 million curies of TRU radionuclides, the release
limits used in determining compliance with §191.13 would be five times the values listed in
Table 1. A feature of 40 CFR part 191 is that the allowable release is linearly proportional
to the amount of TRU waste emplaced in the repository. Second, the variability in the curie
content from drum to drum may be used to calculate the variability hi the quantity of
radioactivity released to the land surface from a borehole which inadvertently intercepts the
waste. The quantity of radioactivity in the waste also enters into performance assessment
indirectly. For example, when coupled with the amount of brine in-flow into a disposal
room, the quantity of radioactivity determines whether the concentration of a nuclide in
solution is limited by solubility (including colloidal formation) or by the total radionuclide
inventory.
At a more fundamental level, the quantity of radioactivity determines whether the waste
meets that TRU waste definitional specification of 100 nanocuries per gram of waste.
Wastes containing less than 100 nanocuries per gram are classified as low-level wastes and
are excluded from disposal at the WIPP
4.3.3 Current and Projected Waste Inventory at the WIPP
Waste destined for disposal at the WIPP is to be packaged in 55 gallon steel drums, Standard
Waste Boxes (which are 1.9 m3 steel containers designed to fit into a TRUPACT-II shipping
package), and cylindrical canisters for RH-TRU. Based on a waste volume of 0.208 m3 for a
55-gallon drum, the capacity of the repository is 846,000 drum equivalents.18 Each disposal
room within the repository is nominally slated to receive 6,804 drums.
According to Revision 1 of the WIPP Transuranic Waste Baseline Inventory Report
(WTWBIR), the DOE TRU waste generator sites currently have hi inventory 73,000 m3 of
CH-TRU and 1,200 m3 of RH-TRU waste (DOE95a). Thus, the current inventory is
approximately 41% of CH-TRU capacity and 17% of RH-TRU capacity. The sites expect to
generate an additional 51,000 m3 of CH-TRU waste and 3,600 m3 of RH-TRU waste hi the
future. Since the current inventory plus the volumes of waste projected to be generated
before repository closure are less than the statutory/regulatory capacity of the repository,
DOE, for scoping purposes, scales the projected inventory so that the statutory capacity is
18 The term drum equivalents is used to reflect the fact that not all the waste is packaged in 55-gallon
drums. The dram-equivalent calculation assumes a repository volume of 176,000 m3.
4-28
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reflected in total inventory numbers. For example, since the currently anticipated RH-TRU
volumes are 4,800 m3, and the capacity as limited by DOE's agreement with the State of
New Mexico is 7,080 m3, an additional 2,280 m3 of waste are added to the anticipated RH-
TRU quantity to reach the repository limit.19 CH-TRU is treated similarly. Details are
presented in Table 4-2.
Table 4-2. Transuranic Waste Disposal Inventory for WIPP (Cubic Meters)
Waste Matrix Code Groups
Contact Handled Waste
Combustible
Filter
Graphite
Heterogenous
Inorganic Non-metal
Lead/Cadmium Metal Waste
Salt Waste
Soils
Solidified Inorganics
Solidified Organics
Uncategorized Metal
Unknown
Total CH Volumes
Remote Handled Waste
Combustible
Filter
Heterogenous
Lead/Cadmium Metal Waste
Salt Waste
Solidified Inorganics
Uncategorized Metal
Unknown
Total RH Volumes
Total TRU Waste Volumes
Stored
Volumes
7.1E+03
4.3E+02
6.7E+02
3.0E+04
1.2E+03
5.6E+01
3.3E+01
3.7E+02
1.7E+04
1.5E+03
1.2E+04
1.7E+03
7.3E+04
1.5E+01
8.9E-01
4.4E+02
O.OE+00
O.OE+00
6.1E+02
8.8E+01
1.1E+01
1.2E+03
7.4E+04
Projected
Volumes
2.7E+04
1.1E+03
4.3E+01
4.6E+03
3.2E+02
1.3E+02
6.0E+01
4.5E+02
8.0E+03
3.0E+02
8.6E+03
O.OE+00
5. IE +04
3.2E+00
2.1E+00
3.3E+03
6.0E+00
2.8E+00
1.7E+02
8.6E+01
2.4E+01
3.6E+03
5.4E+04
Anticipated
Volumes
3.4E+04
1.5E+03
7.1E+02
3.5E+04
1.5E+03
1.8E+02
9.2E+01
8.3E+02
2.5E+04
1.8E+03
2.1E+04
1.7E+03
1.2E+05
1.8E+01
3.0E+00
3.8E+03
6.0E+00
2.8E+00
7.9E+02
1.7E+02
3.5E+01
4.8E+03
1.3E+05
WIPP Disposal
Volumes
6.2E+04
2.6E+03
7.6E+02
3.9E+04
1.8E+03
3.1E+02
1.5E+02
1.3E+03
3.4E+04
2.1E+03
3.0E+04
1.7E+03
1.8E+05
2.0E+01
4.3E+00
5.9E+03
9.8E+00
4.6E+00
9.0E+02
2.3E+02
3.5E+01
7.1E+03
1.8E+05
Source: WTWBIR, Revision 1, Table 3-5
19 In its WTWBIR documentation, Hanford submitted two "suspect" RH-TRU waste streams with a volume
of 41,232 m3. Since no radionuclide information was available on these streams, they were not included in the
scale-up in Revision 1 of the WTWBIR, but it should be noted that the volume of these two streams is six times
the allowable RH-TRU capacity of the repository.
4-29
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TRU waste is a complex mixture of sludges, metals, combustibles such as paper and rags,
soils, filters, graphite, etc. As discussed above, these waste components can influence
actinide solubility, gas generation, and waste strength characteristics. Table 4-3 provides a
comparison of the relative compositions of the CH-TRU and RH-TRU wastes based on the
data in Table 4-2 (DOE95b).
Table 4-3. Estimated Composition1 of Waste Disposal Inventory at WIPP Repository
Capacity (DOE95b)
Final Waste Form
Combustible
Filter
Graphite
Heterogeneous Waste
Inorganic Non-Metal Waste
Lead/Cadmium Metal Waste
Salt Waste
Soil
Solidified Inorganics
Solidified Organics
Uncategorized Metals2
Unknown3
% Total CH
Inventory
34
1.4
<1
22
1
<1
<1
1
19
1
17
1
% Total RH
Inventory
<1
<1
0
83
0
<1
<1
0
13
0
3
1
% Total
CH-TRU and RH-TRU
Inventory
33
2
1
24
1
<1
<1
1
19
1
16
1
1 Totals may not add to 100% due to rounding.
2 Includes all metals/alloys except lead and cadmium.
3
Waste is presently uncharacterized but will be characterized prior to shipment to WIPP.
In developing the information contained in Tables 4-2 and 4-3, DOE prepared profiles for
approximately 360 waste streams at various generating sites. The profiles were then assigned
to one of approximately 130 waste matrix codes (WMC) and the WMCs were categorized
into one of thirteen Waste Matrix Code Groups (WMCG) (DOE95a). The WMC numbers
and the WMCG descriptions are shown in Table 4-4.
4-30
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Table 4-4. Waste Matrix Code Group Names
(Source: WTWBIR, Revision 1, Table 1-2)
Waste Matrix Code Group
Solidified Inorganics
Salt Waste
Solidified Organics
Soils
Uncategorized Metal (Metal Waste
Other Than Lead and/or Cadmium
Lead/Cadmium Metal
Inorganic Non-Metal Waste
Combustible
Graphite
Heterogenous
Filter
Excluded Waste12
Unknown13
: ; : \ Waste Matrix Codes
1000', 11001, 11101, 11201, 1130', 11401, 1190', 1200', 12101, 1220',
12301, 12401, 12901, 30002, 3100, 31103, 31113, 31123, 3113, 31153,
31163, 31193, 3120, 3121, 3122, 3123, 3124, 3125, 3129, 3130,
31313, 31321, 3139lor3, 3150, 3190, 39002, 6100", 61205, 61306,
6140s, 61904, 62007, 62108, 62308, 62907, 73003, 91002, 92002
30002, 3140, 3141, 3142, 3143, 3149, 39002
20001, 21001, 21 101, 2120", 2190', 22001, 2210', 22201, 22901, 29001,
30002, 3114, 3200, 3210, 3211, 3212, 3213, 3219, 3220, 3221, 3222,
3223, 3229, 3230, 3290, 39002, 6100", 61 105, 61904, 62007, 62907,
91002, 92002
4000, 4100, 4200, 4900
50009, 5100, 5110, 5190, 62007, 62208, 7000'°, 7490", 930010
50009, 5120, 5130, 62007, 62208, 7000'°, 7200, 7210, 7220, 7400",
7410", 7420", 930010
50009, 5200, 5210, 5220, 5230, 5240, 5290
50009, 5300, 5310, 5311, 5312, 5313, 5319, 5320, 5330, 5390
50009, 5340
50009, 5400, 5420, 5430, 5440, 5450, 5490, 62007, 62208, 62907
50009, 5410
5250, 5350, 6300, 6400, 7100
8000, 8100, 8200, 8900
1 Liquid waste streams are assumed to be solidified prior to sending to WIPP.
2 WMCs 3000, 3900, 9100, and 9200 are placed in "solidified inorganics," "salt waste," or "solidified
organics," depending on the information provided by the TRU waste generator/storage site.
3 paniculate waste streams are assumed to be solidified prior to sending to WIPP.
4 WMCs 6100 and 6190 are placed in "solidified organics," or "solidified inorganics," depending on the
information provided by the TRU waste generator/storage site.
5 Liquid lab pack waste is assumed to be solidified prior to sending to WIPP.
6 Solid lab packs are assumed to be solidified prior to sending to WIPP.
7 WMCs 6200 and 6290 are placed in "solidified organics," "solidified inorganics," or "heterogeneous" if the
waste stream must be solidified per the generator/storage site. They are placed hi "uncategorized metal,"
or "lead/cadmium metal waste" if they are primarily nonreactive metal contaminated with reactive metal.
Reactive waste streams must be treated prior to shipment to WIPP.
8 Waste stream is assumed to be treated prior to sending to WIPP. Volume change is provided by the TRU
waste generator/storage site.
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9 WMC 5000 is placed in "uncategorized metal," "lead/cadmium metal," "inorganic non-metal,"
"combustible," "graphite," "heterogeneous," or "filter," depending on the information provided by the
generator/storage site.
10 WMC 7000 and 9300 are placed in "uncategorized metal" or "lead/cadmium metal," depending on the
information provided by the generator/storage site.
11 WMCs 7400, 7410, 7420, and 7490 are assumed to be drained of liquid and contain only metal waste.
12 These waste streams are excluded from disposal in WIPP at this time, e.g., PCB and asbestos wastes (see
Table 3-2).
13 If adequate information is provided by the generator/storage site, these WMCs are changed. If there is not
enough information, these waste streams remain as "unknown" and are excluded from disposal in WIPP
until characterized.
Because various waste material parameters (i.e., waste components) are important to
performance assessment calculations, the WTWBIR provides estimates of the mix of
materials expected in the inventory. For example, iron and aluminum are important to assess
the amount of hydrogen gas which might be generated by anoxic corrosion if these metals are
exposed to brine. The estimated ranges for these material parameters, expressed as material
densities, are summarized in Table 4-5 for CH-TRU waste (DOE95a).
Table 4-5. WIPP CH-TRU Waste Material Parameter Disposal Inventory
(Table 5-1 from DOE95a)
Category
Inorganics
Organics
Solidified Materials
Soils
Container Materials
Materials
Iron Based
Aluminum Based
Other Metals
Other Inorganic
Cellulose
Rubber
Plastics
Inorganic
Organic
Steel
Plastic/Liners
Waste Material Density (Kg/m3)
Maximum
2.1E+03
l.OE+03
1.4E+03
2.1E+03
9.6E+02
6.8E+02
8.9E+02
2.2E+03
1.4E+03
1.6E+03
Average
8.3E+01
1.2E+01
2.7E+01
3.9E+01
1.7E+02
2. IE 01
5.3E+01
1.3E+02
8.4E+00
5.7E+00
1.4E+02
3.3E+01
Minimum
O.OE+00
O.OE+00
O.OE+00
O.OE+00
O.OE+00
O.OE+00
O.OE+00
O.OE+00
O.OE+00
O.OE+00
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Using the average values from this table, the waste material density in a drum is 550 kg/m3.
Based on the statutory waste volume, the total weight of waste in the repository would be
about 97 million kilograms (210 million pounds). The waste containers will add another 170
kg/m3 to the inventory or 30 million kilograms (66 million pounds).
4.3.4 Identification of Significant Radionuclides
In addition to information on physical and chemical parameters, the WTWBIR also includes
information on the radioactivity associated with the wastes. The estimated radionuclide
inventories in the WTWBIR, scaled to statutory capacity, are:
• CH-TRU — 3.60 million curies
• RH-TRU — 2.11 million curies
Details are included in Table 4-6.
Table 4-6. Major
(Source:
Nuclides in Disposal Radionuclide Inventory
WTWBIR, Revision 1, Table 4-2)
'; -,-.5'.." NucMiE :::':'/-:.;
Am-241
Ba-137m
Cs-137
Pu-238
Pu-239
Pu-240
Pu-241
Sr-90
U-233
Y-90
TOTAL, major nuclides
TOTAL, all nuclides
TOTAL €11-1110 (Ci)
2.23E+05
5.03E+03
5.32E+03
1.89E+06
3.85E+05
7.22E+04
1.01E+06
4.07E+03
1.38E+03
4.07E+03
3.60E+06
3.60E+06
TOTAL RH-TRU (Ci)
5.30E+02
3.10E+05
3.28E+05
3.53E+03
6.41E+03
1.74E+02
9.06E+02
6.68E+05
8.57E+02
6.68E+05
1.99E+06
2.11E+06
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Virtually all (i.e., 99.4%) of the CH-TRU radioactivity is associated with only five
nuclides—Am-241, Pu-238, Pu-239, Pu-240, and Pu-241. In the case of RH-TRU, 93.5% of
the curie inventory is attributable to four fission products (Cs-137, Sr-90, Y-90, and Ba-
137m) with half-lives of 30 years or less. Because most of the RH-TRU inventory is
composed of nuclides with short half-lives, DOE has estimated that the contribution of RH-
TRU to the total radionuclide inventory in the repository will decline from about 37%
initially to about 1% after slightly more than 200 years (DOE95b). Based on the specific
activity of the five major CH-TRU nuclides, the total quantity of these radioactive materials
in the WIPP is about 7,000 kg or about 0.005% of the total inventory mass. The total
quantity of other very long-lived uranium and thorium radionuclides is about 104,000 kg.
Accurate data on the fractional abundance of each radionuclide contained in TRU waste are
necessary because differences in solubility, mobility, and half-life determine the extent to
which specific radionuclides reach the accessible environment in a given scenario. The
behavior of uranium isotopes U-233 and U-234 provides a good example of the importance
of understanding the radionuclide composition of TRU wastes in assessing their potential
migration to the accessible environment. In the 1992 performance assessment (SAN92),
U-233 and U-234 were estimated to comprise approximately 0.06 percent of the initial
inventory, yet they accounted for about 21 percent of the projected discharge to the
accessible environment (for the E1E2 scenario at 1,000 years with fracture flow, matrix
diffusion, and no retardation). Accurate determination of the uranium inventory is thus very
important, even though its quantity is minor compared to plutonium and americium
radionuclides.
4.3.5 Determination of Actinide Solubility Limits
Actinide solubility in the Castile or Salado brines that come in contact with the waste is
generally thought to be one of the most important parameters for calculating releases to the
accessible environment (SAN92). Because acjinide solubility is not well understood, there is
considerable uncertainty in estimating the quantities of plutonium, americium, and uranium in
solution. Estimates of the solubilities of actinide species expected in TRU wastes had a
range of 13 orders of magnitude in the 1992 performance assessment (SAN92). The mean,
median, and range of values used in SAN92 were obtained by expert elicitation.
4-34
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In addition to pure solubility (where solid material is dissolved in the liquid) which can be
affected by brine salinity, pH, Eh, and the presence of chelating agents and other chemical
constituents, there are concerns and greater uncertainty about the possible concentrations of
colloidal dispersions (very fine particles in the 0.001 to 0.1 /xm diameter range that can
remain suspended in the liquid). Colloid formation was not considered in the 1992 PA
(SAN92).
To provide more defensible information, DOE has been conducting laboratory experiments
on actinide solubility and colloid formation under the Actinide Source Term Program (ASTP)
(LOS93, NOV94a, NOV94b). The ASTP has been using small-scale laboratory experiments
to develop a conceptual model of actinide solubility. DOE intends to verify this model using
large-scale tests (the Source Term Test Program-STTP) with TRU wastes. These tests are
currently in process at Los Alamos National Laboratory (LANL). Questions remain
regarding the extent to which these studies are representative of actinide mobility in TRU
wastes under disposal conditions. For the final Compliance Certification Application, DOE
is proposing to include a look-up table which will define solubilities in various environments.
The solubility (or dissolved species) model is an "equilibrium thermodynamic model based
on the Pitzer formalism for activity coefficients in concentrated electrolytes" (BYN95). The
dissolved species model is developing experimental solubility data, in brines of various
compositions and ionic strengths, for five actinide elements—americium, neptunium,
thorium, uranium and plutonium in four valence states—+III, +IV, +V, and +VI.
Ultimately, the dissolved species model is expected to provide to performance assessment the
solubility for these five actinide elements as a function of:
• oxidation state
• brine type
pH
pC02
• organic ligands
The partitioning of the actinide elements between the four possible oxidation states must also
be specified for PA. In recent modeling studies, solubility ranges of 1 to 10~10 were assigned
for all oxidation states with median values ranging from 10"7 moles per liter for +IV to 10"4
moles per liter in the +VI oxidation (SNL95). Suggested oxidation state distributions in the
4-35
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same study were:
• Americium — all +IE
• Thorium — all +IV
• Uranium — 0 to 20% +VI, balance: randomly distributed among +III, +IV
and +V
• Neptunium - randomly distributed between +IV and +V
• Plutonium — 0 to 20% +VI, balance: randomly distributed among +III,
+IV, and +V
The specification of the oxidation state distribution for each element poses some difficult
technical questions. Since the disposal room environment is expected to become anoxic early
in the life of the repository, logic would suggest that the actinide elements will exist in their
reduced oxidation states (NOV94a). However, research has shown that alpha radiolysis can
convert Am+III to Am+V (NOV94a) and the presence of carbonate stabilizes plutonium in
the +VI state (REE94). Consequently, DOE chose to use statistical sampling to characterize
the mix of oxidation states for the various actinide elements to be included in PA. The
STTP may provide additional experimental insight into these distributions (NOV94a).
Based on experimental work under the ASTP currently hi progress, DOE plans to refine the
data used in the dissolved species model by the end of the first quarter of 1996 and use these
data in the final compliance certification application (BYN95).
4.3.6 Determination of Gas Generation Rates
Volatile organic compounds (VOCs) present in TRU waste can vaporize after waste
emplacement in the disposal system and create a potential problem for compliance with
RCRA regulations. Gases other than VOCs are also expected to be generated in the waste as
a result of corrosion, microbial activity, and radiolysis. These processes are expected to
produce gases in much greater quantities than from VOCs present in the waste and represent
the principal mechanisms of concern in performance assessment.
In PA, it is necessary to evaluate the combined effect of gas generation on waste storage
room closure and brine inflow. The pressure resulting from significant gas generation could
retard the rates of both room closure and brine inflow. In the absence of any gas generation
4-36
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there would be no retardation of room closure rates or brine inflow. The determination of
the rates for room closure and brine inflow requires complex modeling with computer codes
where coupling of physical processes is difficult and use of parameters that have not been
measured on actual or, in many cases, even simulated waste.
An analysis of the combined effect of room closure and brine inflow requires an assessment
of which occurs first. If complete closure occurs before brine inflow, the enclosed space's
very low permeability and porosity could effectively minimize any future brine inflow and
mixing with waste. The amount of contaminated brine available for release by drilling would
thus be minimal. Conversely, if brine inflow occurs before complete room closure, there
could be extensive mixing of disposal system contents with brine, creating a significant
amount of contaminated brine available for release in a drilling puncture.
Gas generation is also directly related to actinide solubility, discussed in a previous section.
Preliminary work under the ASTP indicates that the presence of carbon dioxide gas (CO2)
directly affects the solubility of plutonium, uranium and other actinides under laboratory
conditions (SAN93). The applicability of this information to actual TRU wastes under
disposal conditions remains to be demonstrated. As previously mentioned, gas generation
can also impact the amount of waste spallation associated with drilling events.
Waste components will affect gas generation rates and processes. The amount of gas
generated by corrosion is directly related to the quantity and type of metal present in waste
and waste containers, the surface area of the waste, and available moisture. The amount of
gas generated by microbial activity is related to the amount of available moisture and
cellulosic material (e.g., paper, cloth, and wood). Radiolytic gas generation is a function of
the amounts of alpha radioactivity, moisture, and cellulosic material present.
The initial liquid content of the waste may be important to its gas generation characteristics
(SAN92). Table 3-1 of the WIPP WAC notes that, as a guideline, residual liquid in well-
drained containers should be restricted to approximately one volume percent of the internal
container, with the aggregate amount of residual liquid less than one volume percent of the
external container (DOE91). The residual liquid limit could be checked in three ways:
• Upon assembly of the drum by personnel at the waste generator site;
• By radiography performed on site by waste generators during the drum
certification process
• During visual examination of a container, as applicable
4-37
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While the combination of these three techniques appears adequate to meet the residual liquid
criterion, the use of one technique alone may not suffice. For example, radiography has not
been demonstrated to be a fail-safe method for detecting containers of liquids within a waste
drum. In January 1993, a full 8-ounce can of adhesive was missed by an operator
conducting Real-Time Radiography (RTR) at INEL, and later discovered during the visual
examination of the drum contents. Radiography detects movement of liquids within a
container; therefore a completely full container could be missed.
4.3.6.1 Average Stoichiometry Model
The average Stoichiometry model was used to calculate quantities of gas generated in the
1992 performance assessment (SAN92). DOE also plans to use this model for calculations in
the final Compliance Certification Application (NOW95). The average Stoichiometry model
is part of BRAGFLO—a computer code which calculates two-phase flow in the repository
Thus, brine and gas flows into and out of the repository are coupled to gas generation (i.e.,
pressure). Sufficient gas pressure can also cause fracturing of the nearby anhydrite layers
increasing their permeability. In addition, BRAGFLO uses a porosity surface developed by
the SANCHO/SANTOS computer codes to simulate room closure. This porosity surface is a
function of the amount of gas present at any point in time. In this way, gas generation is
also coupled to the geomechanical behavior of the disposal rooms.
The average Stoichiometry model considers the anoxic corrosion of ferrous materials and the
anaerobic degradation of cellulosics and rubbers, and calculates the quantity of gas generated
and the quantity of water consumed. DOE has discussed the fact that aluminum and its
alloys could behave hi a similar manner to ferrous materials, but have not included aluminum
corrosion in the model.20 The model does not include possible gas consuming reactions nor
does it address other possible gas producing mechanisms such as radiolysis.
20 Based on the information contained in Table 4-5, it can be estimated, using average values, that the
amount of hydrogen produced from ferrous metal corrosion could range from 6.9 to 9.2 x 108 moles depending
on which corrosion reaction occurs and assuming the presence of sufficient water to consume all the iron.
Similarly, the amount of hydrogen produced by the corrosion of all of the aluminum would be 1.2 x 108 moles
or 11 to 14% of the amount from iron corrosion. However, using the maximum values in Table 4-5, the
contribution from aluminum would be more than 50% of the total hydrogen. These calculations assume that 1.5
moles of hydrogen are generated for each mole of aluminum consumed.
4-38
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Two possible anoxic corrosion mechanisms are considered in the model:
Fe + 2H2O = Fe(OH)2 + H2 (1)
3Fe + 4H2O = Fe3O4 + 4H2 (2)
Equation 2 produces 33 % more hydrogen per mole of iron consumed than does equation 1.
Because uncertainty exists as to which equation prevails, DOE has chosen to treat the
stoichiometry of the reaction as an uncertain variable which is sampled over the range of
possible values for performance assessment calculations. To do this, the two equations above
are combined into an "average" equation as follows:
Fe + ((4+2x)/3)H2O = (4-x)H2 + (3x)Fe(OH)2 = ((l-xXS^eA
The values of x are assumed to be uniformly distributed between 0 and 1 for Latin
Hypercube sampling purposes in PA.
Inundated corrosion rates have been developed from laboratory corrosion studies of mild
steels for up to 24 months duration in brine solutions with the pH ranging from an initial
value of 6.7 to approximately 8.3 at the end of the tests and a nitrogen overpressure of 10 to
15 arm. The measured hydrogen production rates as a function of time were as follows:
3 months — 0.19 moles H2 per m2 steel surface per year
6 months — 0.21 moles H2 per m2 steel surface per year
12 months — 0.16 moles H2 per m2 steel surface per year
24 months — 0.10 moles H2 per m2 steel surface per year
SNL recommended that a value of 0.1 moles/m2-y (3 x 10~9 moles/m2-sec) be used as the
best estimate (i.e., median value) (BRU94).
To obtain an estimate of the maximum inundated corrosion rate, it was assumed that the
actual pH of the brines in the WIPP repository could vary from 3 to 12. Based on work by
earlier investigators cited in BRU94, SNL assumed that the anoxic corrosion rate was
essentially constant between pH 4 and 10. Outside this range the following pH dependent
changes were anticipated: at pH 3, the rate was expected to be higher by a factor of 50; at
pH 11, it would be lower by a factor of 0.05; and at pH 12, it would be lower by a factor of
0.005. In addition, it was assumed that the corrosion rate would increase with pressure and
4-39
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at lithostatic pressure the rate would be four times higher than under the experimental
conditions noted above. Consequently, the maximum rate for inundated corrosion was set at
4 x 50 x 0.1 or 20 moles/m2-y (BRU94). This is equivalent to a maximum value of 6.35 x
10'7 moles/m2-sec.
Corrosion rate data for humid environments (i.e, where the steel is exposed to water vapor)
were also developed. Based on the amount of brine present in a disposal room at any given
time as calculated by BRAGFLO, the relative amounts of steel subject to inundated corrosion
and humid corrosion are calculated. For PA purposes, it is necessary to convert these
corrosion rates to a volumetric gas generation rate (i.e., moles H2 per m3 of repository
volume per second). This requires information on the surface to volume ratio of the contents
of an average drum. To perform this conversion, SNL assumes that a drum of CH-TRU
waste has an approximate area of 4 m2 and the contents of the drum contribute an additional
2 m2 (BRU94). If one assumes that the drum and its contents have the same surface to
volume ratio (as was assumed by SNL in the past) and the surface area of the drum is
actually 4.5 m2, then, from the current average inventory data in Table 4-5, it can be
estimated that the surface area of the ferrous contents of a drum is 2.7 m2 and the total
surface area of steel per drum is 7.2 m2 which is 20% higher than the value being used in
the 1992 PA (SAN92). For microbial reactions, the following highly generalized equation is
used to calculate gas generation (BRU94):
CH2O + unknowns + microorganisms = (5/3)gas + unknowns (3)
CH2O, a simplified formula for glucose, is assumed to represent various organic materials
(cellulosics, rubbers, and plastics) present in the waste which may be subject to microbial
degradation. The actual reactions which could occur and the extent to which water is
produced or consumed are subject to a considerable amount of uncertainty. The quantity of
gas produced could vary from 0 to 1.67 moles per mole of glucose consumed depending on
which of several possible reactions is assumed to occur. Consequently, the stoichiometric
coefficient is assumed for PA to vary uniformly over this range rather than remain fixed as
shown in equation (3) (SAN92). Plastics and rubbers are expected to be generally more
resistant to microbial degradation than cellulosics (papers and rags), but may be subject to
some radiation-related degradation reducing their resistance to microbial attack. The extent
to which rubbers and plastics enter into the gas generation reactions is uncertain, but should
be addressed in PA. Similar to the treatment of anoxic corrosion, assumptions were made
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that microbial degradation in humid environments proceeded at some fraction of the
inundated rate. This fraction was assumed to vary uniformly from 0 to 0.2.
4.3.6.2 The Reaction Path Model
SNL has also developed a more sophisticated model to analyze gas generation, called the
reaction path model. This model includes treatment of oxic and anoxic corrosion of steels
including passivation and depassivation reactions, microbial degradation, radiolysis of brine,
and consumption of carbon dioxide by calcium-bearing species. Unlike the average
stoichiometry model, the reaction path model uses thermodynamic calculations to estimate
phase stability at any point in time. For example, at certain CO2 fugacities, iron carbonate
may form which prevents anoxic corrosion of the ferrous materials. If the CO2 fugacity is
reduced sufficiently, the passivating layer can decompose allowing corrosion to proceed.
While SNL has judged this to be the most comprehensive gas generation model (BRU94),
DOE decided that the average stoichiometry model will be used for compliance
demonstration. This decision is based on the position that PA has shown a low sensitivity to
gas generation and the reaction path model is "unnecessarily costly in time and resources for
PA calculations" (NOW95). The model will be retained to support calculations related to
actinide chemistry.
4.3.7 Establishing the Waste Envelope
For the Compliance Certification Application, DOE will conduct performance assessments
using the available information on waste characteristics and waste components, which must
demonstrate that the WIPP complies with §191.13 and §194.34. Many of the waste
properties will not be precisely known values which can be used for input to PA as
constants. Rather, they will be imprecisely known variables for which a range and
probability distribution function will be assigned. The PA process will define an acceptable
envelope of waste properties which will ensure compliance with the regulations on a
statistical basis. This is not to say that some individual complementary distribution functions
(CCDFs) produced from particular combinations of uncertain parameters may not exceed the
limits in §191.13, but it is required that the mean CCDF comply with §194.34(f) (i.e., there
is 95 % confidence that the mean of the population of CCDFs meets the disposal standards in
§191.13).
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Once an acceptable waste envelope has been defined through PA calculations, DOE must
have procedures in place to ensure that the actual wastes fall within this envelope. It is
conceivable that an actual waste component could lie within the range used to develop the
waste envelope, but have a different probability distribution function than was assumed in the
PA calculations used for compliance determination. Compliance might not then be
demonstrable with actual waste. To preclude this possibility, 40 CFR 194.24 contains
procedures for showing compliance at the waste envelope limits. These procedures are
discussed in Section 4.6.
4.4 METHODS FOR CHARACTERIZING WIPP WASTE INVENTORY
The DOE/CAO Quality Assurance Program Description (QAPD) (DOE95c) is the quality
management document that identifies the quality requirements applicable to WIPP waste
characterization. The QAPD establishes the minimum requirements for the development of
QA programs for the National TRU Programs Office (NTPO) and the generator sites'
QAPjPs. The DOE states that the requirements in the CAO QAPD "are based on the QA
requirements and criteria contained in 10 CFR §830.120," and that the QAPD is "consistent
with applicable Environmental Protection Agency (EPA) requirements" (DOE94d).
As mentioned previously, the controlling document for TRU waste characterization is the
TRU QAPP (DOE94b). This document outlines two approaches, one for retrievably stored
wastes and one for newly generated wastes. Both approaches are based in large part on the
waste classification system presented hi the WTWBIR.21 DOE asserts hi the WTWBIR that
the Waste Matrix Codes (WMCs) used to categorize wastes have been established based on
"grouping wastes with similar physical and chemical properties."22 In the TRU QAPP,
DOE states their rationale for using WMCs to track TRU wastes:
21 The waste classification presented in the WTWBIR was initially developed by DOE and was presented in
the DOE Waste Treatability Group Guidance (DOE95) which was prepared to meet the requirements of the
Federal Facilities Compliance Act (FFCA) of 1993. This approach was used hi DOE's Mixed Waste Inventory
Report (MWIR) (DOE93).
22 It should be noted that wastes may be categorized differently depending on the waste management
objective, e.g., for purposes of storage, transportation, or treatment. For example, wastes with the same WMC
would be stored together due to their similar physical or chemical nature. For transportation, wastes would be
grouped according to different requirements, e.g., the requirements of the TRUPACT-II content codes.
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To ensure consistency throughout the DOE complex regarding TRU waste
inventory information, TRU waste characterization information will be
correlated to the Waste Matrix Codes established by DOE as acceptable to the
WIPP facility.23
The TRU QAPP states that there are three broad groups of WMCs:
• solid process residues — 3000 series
• soils — 4000 series
• debris wastes — 5000 series
Existing wastes in these three WMCs will be considered retrievably stored wastes and will be
characterized directly. Existing wastes in the other WMCs described in Table 4-4 (WMCs
1000, 2000, 6000 and 9000) will require treatment prior to shipment to the WIPP and will
then be considered newly generated wastes. Wastes will be characterized for disposal in
accordance with the approach outlined below. The Acceptable Knowledge Guidance Manual
(DOE95c) states that
Four broad matrix parameter categories of waste are used to describe the
physical form of the waste and to determine TRU waste characterization
requirements: homogeneous solids (summary category S3000), soil/gravel
(summary category S4000), debris wastes (summary category S5000), and
special wastes (summary category S7000).
The Acceptable Knowledge Guidance Manual further states that Series 7000 (special) wastes
will be classified as RCRA hazardous in the same manner as the Series 5000 (debris) wastes
(DOE94b). For both waste types, the determination of their RCRA hazardous classification
will be made using acceptable knowledge alone, without corroborative empirical sampling
and/or analysis. A generalized flow diagram for TRU waste characterization is presented in
Figure 4-1. The specific approaches for characterizing newly generated and retrievably
stored wastes provide statistically derived means to select waste containers from all three
WMCs for verification by visual examination, and waste containers from series 3000 and
4000 WMCs for RCRA characterization.
23 Many wastes will have other identification codes that are no longer used as well as EPA derived
hazardous waste codes assigned to them. This creates considerable confusion.
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The steps in characterizing newly generated and retrievably stored waste are as follows:
• establish profiles for waste streams
• using process knowledge, assign waste containers to waste streams
• assign a waste matrix code to each waste stream
• test all waste containers using headspace gas analysis, radiography, and
radioassay
• select statistically determined number of waste containers for RCRA
characterization and/or visual examination, depending on the assigned WMC
• determine if waste is hazardous and develop a WMC description
In addition, for newly generated wastes, it is necessary to verify that the waste generating
processes have operated within the profile's established administrative controls.
The information listed above must be coordinated with a consideration of the manner in
which the waste stream is defined. The definition applicable to TRU waste is found in
DOE's Acceptable Knowledge Guidance Manual (DOE95c), and is of fundamental
importance to TRU waste characterization because "waste characterization and DQO
activities are performed on a waste stream basis" (DOE95c). For the purposes of
characterization, TRU waste streams are distinguished on the basis of three factors:
• the summary category of the waste (WMC, WMCG, etc.)
• the waste's status (newly generated or retrievably stored)
• the waste gensis (continuous process or batch)
The combination of these three factors determines the waste stream's anticipated variability
and the extent of verification required. Additionally, waste streams are identified on the
basis of their waste characterization objectives as defined by the applicable regulatory
requirements, e.g., RCRA, TRUPACT-II, etc.
DOE's TRU Waste Characterization Program currently consists of the following six
activities: radiography, radioassay, headspace sampling and analysis, solid process residues
and soils sampling and analysis, visual examination, and use of acceptable knowledge/process
knowledge. Other aspects of TRU waste characterization typically involve scientific research
(actinide solubility, etc.) to define waste characteristics (see Section 4.3.1). Radiography,
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Document changes
The waste is in
Waste Matrix Code
series 500O
Satoct waste population* and
' identify drum* for visual examination 1
and solid process residues and soils
campling and analysis
1
Perform Radiography
No
Does Waste Matrix
Code still appty to waste?
Perform Radioaseay
Perform solid process residues
and soils sampling and analysis
on selected drums
Transmrt data to DOE/CAO
Perform headspace gas
campling and analysis
i
,
Perform visual examination
on ea looted drums
Does Waste Matrix
Code still appty to waste?
Waste is excluded
from the Program
Figure 4-1. Generalized Sequence for TRU Waste Characterization
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radioassay, headspace sampling and analysis, solid process residues and soils sampling and
analysis, visual examination are summarized in the sections below, and the use of acceptable
knowledge/process knowledge is discussed in greater detail in Section 4.6.
All TRU waste generators currently perform some waste characterization activities on site,
although their capabilities vary considerably. The major TRU generator facilities are: Idaho
National Engineering Laboratory, Oak Ridge National Laboratory, the Rocky Flats
Environmental Technology Site, Savannah River Site, Hanford, Los Alamos National
Laboratory, the Nevada Test Site, Lawrence Livermore National Laboratory, Argonne
National Laboratory-East, Argonne National Laboratory-West and the Mound Plant
(DOE94d). Some of these sites have multiple facilities involved with some aspect(s) of TRU
waste generation, characterization, and storage. As indicated in Table 4-7, these sites have a
mix of equipment required to perform the analytical techniques listed above.
Table 4-7. Waste Characterization Capabilities of Ten Main TRU Waste Generators
TRU Generator Site
Current Waste Characterization Capabilities
Oak Ridge National Laboratory (ORNL)
Hanford (HANF)
Idaho National Engineering Laboratory (INEL)1
Argonne National Laboratory-East (ANL-E)
Savannah River Site (SRS)
Rocky Flats Environmental Technology Site (RFETS)
Los Alamos National Laboratory (LANL)
Lawrence Livermore National Laboratory (LLNL)
Nevada Test Site (NTS)4
Mound Plant (MOUND)5
RT RA VE SA HG2
RT RA VE3 HG3 SS3
RT RA HG VE SA SS
RAHG
RT RA
HG RT RA VE SS SA
RT RA HG VE SS SA
RA VE HG RA
RT = Radiography
RA = Radioassay
VE = Visual Examination
HG = Headspace Gas Sampling and Analysis
SA = Solid Residue Analysis
SS = Solid Residue Sampling
1 Includes Argonne National Laboratory-West (ANL-W).
2 Expected to have this capability by FY 1996.
3 Expected to have this capability by 2002.
4 NTS currently plans to use the mobile TRU characterization system being developed by LANL.
5 Mound's plans for TRU characterization are currently uncertain.
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In general, facilities that either have historically produced or currently manage plutonium or
plutonium-bearing wastes as part of their routine operations have radioassay facilities for the
purpose of nuclear accountability. These facilities can be used for waste characterization
purposes. Los Alamos National Laboratory is currently developing a mobile TRU
characterization system for use by small-quantity TRU sites (MAR95). While DOE will
require all TRU waste generator sites to be fully capable of certifying their own wastes prior
to shipment to WIPP, the specific details and logistics regarding characterization are
unavailable at this time.
4.4.1 Radioassav
Radioassay involves a variety of measurement techniques used to determine the radionuclide
content of a waste container. Typically, TRU waste generators are most interested in certain
radionuclides, specifically actinide or transuranic species. However, for purposes of
radionuclide inventory, many other radionuclides are quantified predominantly by
measurement of their gamma emission. Generally, TRU waste generators use non-
destructive techniques based on neutron or gamma measurements to quantify the
Physical measurements, i.e., inductively coupled plasma/mass spectrometry (ICPMS), are
also used, but less frequently. Passive Active Neutron (PAN) counting and Segmented
Gamma Scan (SGS) counting are two examples of systems in common use.24 PAN is used
to identify and quantify the odd- and even-numbered isotopes of plutonium by measuring
their neutron emission both spontaneously in the passive mode and in response to
bombardment within the detector, the active mode. SGS measures the photon emissions
from a waste container using a standard intrinsic germanium type of photon detection system
coupled with a transmission source, typically Se-75 for assays of weapons grade Pu-239. A
container is divided into a number of segments and each segment is assayed with the
transmission source to develop a waste drum specific photon attenuation correction factor by
segment. Next the drum is measured without the source and the radionuclides of interest are
quantified. Computer enhancement of the data provides a more complete assay of the drum's
photon emitting radionuclides (Pu-239, Am-241, etc.).
24 There are other radiometric techniques used for radioassay, such as Pulse Neutron Coincidence Counting
(PNCC) and gamma determinations using a simple unsegmented intrinsic germanium type photon detection
system. Still other methods are currently under development.
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Due to the wide variety of assay systems employed by TRU generators, concerns have been
expressed regarding the comparability of radioassay data among DOE sites, partially in
response to this, DOE recently implemented a performance demonstration program (PDF) for
radioassay techniques comparable in principle to the PDF for Headspace Gas Analysis,
described hi Section 4.4.3. PDF participants receive a "standard" waste drum with a known
activity concentration and isotopic distribution. Each participant analyzes the drum and
reports the results to the program coordinator for scoring and statistical evaluation.
Participants are required to use the same techniques for PDP samples as they use for actual
characterization of TRU wastes and are "qualified" for that specific technique or combination
thereof. Qualification is mandatory and must be maintained to enable a site to certify and
ship TRU waste to WIPP.
4.4.2 Radiography
Radiography is a nondestructive, non-intrusive qualitative technique used to identify the
contents of a waste container. Most DOE sites currently employ Real-Tune Radiography
(RTR), which uses x-rays and a video system to allow an operator to view the container's
contents in real-time. RTR's primary use is to examine and verify the physical form of the
waste and to ascertain that a container complies with the specifications of a content code or
other physical requirements. The Quality Assurance Objectives (QAOs) for radiography do
not address precision or include specific Minimum Detectable Levels (MDLs) because this
technique is primarily a qualitative determination (DOE94b). A statistically determined
subset of the waste containers examined with radiography will be verified independently by
visual examination (DOE94b). The overall approach to visual examination of waste is
presented in Figure 4-2.
While radiography is generally effective, certain materials, particularly lead liners, are not
readily penetrated by x-rays and render radiography ineffective when they are present in a
waste container. DOE has acknowledged this and states in their Waste Characterization
Methods Manual (DOE95d) that
Containers with high density waste (e.g., leaded rubber, cemented sludges) can
only be examined at then- edges. In addition to this limitation, waste
containers that are configured with a lead liner cannot be examined with
radiography
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Use radiography to
determine/verify the
Waste Matrix Code
and Waste Material
Parameter weights
\
1
Visual examination
\
1
Visually examine unopened
waste bags/packages
Perform a limited visual
examination through the
unopened bags/packages
Confirm radiography-
indicated Waste Matrix
Code and Waste Material
Parameter weights
Can Waste
Matrix Code (DOE 1994d)
and Waste Material Parameter
weights (Table 10-1) be
determined without opening
bags/packages?
Perform a full
visual examination
Open bags/packages
Determine Waste Matrix
Code and Waste Material
Parameter weights
Based on the results of visual
examination calculate the
percentage of waste containers
with incorrectly assigned Waste
Matrix Codes
Figure 4-2. Programmatic Approach to Visual Characterization of TRU Waste
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As discussed in Section 4.3.6, small containers completely full of liquid intermingled with
other waste in a drum can appear to be empty due to the lack of visible fluid movement upon
agitation, and may be missed by operators. Radiography has historically been performed
manually, which is tedious and labor intensive. However, DOE has been investigating the
feasibility of digitizing the current analog information obtained with RTR and hopes to
realize sufficient gains in efficiency to allow installation of an automated system at INEL and
possibly at other sites. DOE has made the point that there is "no equivalent or associated
method found in EPA sampling and analysis guidance documents." There are other
industries that use radiography and may have protocols applicable to DOE. DOE further
states in their Waste Characterization Methods Manual that:
Standardized training requirements for radiography operators are based on
existing industry standard training requirements and comply with the training
and qualification requirements of ASME NQA-1, Element 2, except for
Supplement 2S-2 (DOE95d).
There is no DOE-wide formal certification or accreditation process for radiography operators
and each site specifies how it will achieve the training requirements and QAOs presented in
the TRU QAPP in their QAPjP.
4.4.3 Headspace25 Sampling and Analysis
Headspace sampling and analysis are the determination of the chemical composition and
concentration of flammable gases, volatile organic compounds, and other gases contained in
the void volumes of waste containers. These compounds are determined by gas
chromatography and/or gas chromatography-mass spectrometry. TRU wastes are typically
packaged in 208 liter (55 gallon) drums. The drums contain 90 mil polyethylene liners, and
inside each liner is a 208 liter polyethylene bag that may contain many other smaller bags.
Sampling within a waste container can occur in three general areas: in the innermost layer
of confinement, i.e., any of the small bags within the drum's interior; in the spaces within
the drum liner; and under the drum lid, in the space between the drum lid and the sealed
drum liner. The 3-year-old WIPP Performance Demonstration Program for Headspace Gas
25 The term "headspace gas" should be interpreted to mean hydrogen, methane, and the volatile organic
compounds that exist within a layer of confinement in a TRU waste container (DOE945).
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Analysis is detailed in DOE92. This program is used to qualify DOE TRU generators to
certify TRU waste containers for shipment to WIPP. Once a participant is qualified using a
technique(s), the participant may characterize waste containers for shipment to WIPP using
only that same analytical technique(s) used to analyze the PDP samples, participation is
mandatory and blind samples are distributed to all participants annually.
4.4.4 Solid Process Residues and Soils Sampling and Analysis
Solid process residue and soil sampling and analysis are used to determine the hazardous
constituents in TRU wastes classified as solid process residues and soils (Series 3000 and
4000 WMCs). Sampling procedures are based on methods found in EPA's SW-846 (EPA86)
and are detailed in the Methods Manual (DOE95d). The analytical procedures to be used are
also based on SW-846, but were modified by Los Alamos National Laboratory for this
purpose. A facility for these analyses is presently operational at Oak Ridge National
Laboratory. The DOE intends to use sampling and analysis primarily to verify
characterizations made using process knowledge.
4.4.5 Visual Examination
Visual examination is the characterization of the contents of a waste container by physical
removal, inspection, and sorting for the purpose of establishing or verifying that the correct
waste codes have been assigned. In this time-consuming, hands-on process, the contents of a
drum are unpacked, examined, segregated if necessary, and repackaged. Several TRU
generators have modified facilities that can be used for this purpose. However, it is not clear
whether DOE will require each TRU waste generator to have this capability on site or if
certain sites would be designated to perform this function for others. Argonne National
Laboratory-West has a waste characterization chamber designed for visual examination of
waste containers. DOE considers visual examination to be a means of verifying assumptions
made using process knowledge, e.g., correct waste code assignment and absence of non-
conforming items (residual liquids, compressed gases, etc.). For newly generated wastes,
DOE intends to use process knowledge and prospective documentation of each waste
container's contents to ensure each container's compliance. For all TRU wastes (newly
generated and retrievably stored), DOE says that
As a QC check, a statistically significant portion of the certified waste
containers must be opened and visually examined. (DOE94b)
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The actual number of containers examined must be empirically derived by each site annually,
and DOE asserts that
The number of waste containers requiring visual examination will ensure that
the Program is 80-percent confident that, if the indicated number of waste
containers is examined, the UCL^ of the miscertification percentage will be
less than 14 percent, (i.e., there is only a 10-percent chance that the
miscertification rate is greater than 14 percent). (DOE94b)
4.4.6 Use of Acceptable Knowledge/Process Knowledge26
Each of the above techniques is intended to complement the waste characterization data
generated using process knowledge. DOE intends to rely heavily on process knowledge for
most WMCs and to use it as the primary means of waste characterization for newly
generated waste and retrievably stored WMC 5000 wastes (DOE94b). DOE anticipates that
retrievably stored waste will require more frequent verification by empirical techniques to
certify wastes in accordance with all applicable requirements. Because process knowledge is
such an important element in waste characterization it is discussed in detail in Section 4.5
below.
4.5 USE OF PROCESS KNOWLEDGE (ACCEPTABLE KNOWLEDGE) TO
CHARACTERIZE TRU WASTES
4.5.1 Definition and Regulatory Precedent For the Use of Process Knowledge (Acceptable
Knowledge)
The DOE recently released guidance to address the use of acceptable knowledge/process
knowledge for the characterization of TRU wastes (DOE95c). This guidance document
provides the following:
• definitions for acceptable knowledge and process knowledge
• guidance to distinguish types of waste streams for waste characterization
purposes
26 The term process knowledge has historically been used to refer to what DOE currently calls acceptable
knowledge. As defined by DOE (DOE95e) and discussed in the text, acceptable knowledge is a broad category
of types of information that includes process knowledge.
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• classes of acceptable knowledge
• Quality Assurance requirements for the use of acceptable knowledge to
characterize TRU wastes
• specific requirements for acceptable knowledge documentation
The document summarizes DOE's approach to the use of process knowledge for
characterizing TRU wastes that previously was scattered among many DOE and CAO
documents (DOE94b, DOE94e, DOE95c). In this document, DOE has followed EPA's
approach of defining process knowledge as a subset of acceptable knowledge (EPA92). DOE
defines acceptable knowledge as follows:
Acceptable knowledge includes process knowledge and results from previous testing,
sampling, and analysis associated with the waste. Acceptable knowledge includes
information regarding the raw materials used in a process or operation, process
description, products produced, and associated wastes. Acceptable knowledge
documentation may include the site history and mission, site-specific processes or
operations, administrative building controls, and all previous and current activities
that generate a specific waste.
DOE also states that-
Acceptable knowledge refers to information that can be used for waste
characterization in lieu of waste sampling and analysis conducted in accordance with
the requirements specified hi the Transuranic Waste Characterization Quality
Assurance Program Plan, and may include process knowledge and the results of
previous surrogate waste sampling and analysis.
DOE defines process knowledge as follows:
Process knowledge is a term used by the EPA to refer to detailed information on a
waste that is obtained from existing published or documented waste analysis data or
studies conducted on hazardous wastes generated by process [s] similar to that which
generated the waste. Process knowledge describes the process or operation that
generated the waste that is being characterized. Process knowledge is used to identify
specific constituents hi a waste stream and the method (or process) by which the
constituents are used that created the final waste.
The precedent for the use of waste-related information hi waste characterization originates in
the Resource Conservation and Recovery Act (RCRA). Under RCRA, a waste generator is
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allowed to use "acceptable knowledge" to determine whether a waste is hazardous (EPA94).
As stated above, process knowledge is one form of "acceptable knowledge." DOE has
determined that "when used in conjunction with other waste characterization techniques"
acceptable knowledge "is appropriate to obtain the required TRU waste characterization
information" (DOE95c). This information encompasses many aspects of TRU waste,
including WMC, physical form, and assignment of a waste container to a specific waste
stream. This information will be required to determine compliance with the acceptance
criteria from the WIPP WAC, TRUPACT-II, and the TRU QAPP.
Historical definitions of process knowledge within the EPA-regulated community of RCRA
waste generators typically include two important aspects:
• they were used solely for the purpose of determining that a waste is hazardous
under RCRA; and
• they focused on engineering assessments of waste streams where waste
characterizations were based on computational methodologies that were
documented, such as mass balance or process engineering diagrams.
While DOE's definition includes these aspects and others, it is not clear that DOE's use of
process knowledge is completely consistent with RCRA. In the TRU QAPP (DOE94b) and
the Transuranic Waste Characterization Acceptable Knowledge Guidance Manual (DOE95c),
DOE outlines the main purposes for the use of process knowledge, including:
• sorting newly generated and retrievably stored waste containers into waste
streams;
• estimating the volume and weight of a waste container's contents;
• determining if WMC Series 3000 & 4000 wastes exhibit toxicity characteristics
as specified in 40 CFR part 261, Subpart C, in conjunction with empirical
sampling and analysis;
• determining if WMC 5000 Series wastes are RCRA hazardous in the absence
of empirical sampling and analysis;
• selecting the appropriate method to quantify a waste drum's radionuclide
content; and
• describing waste stream continuous processes and changes over time.
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4.5.2 Using Process Knowledge for Waste Characterization
The credibility of using process knowledge ultimately rests upon the user's ability to provide
the appropriate support documentation. This documentation must demonstrate that the waste
producing process was adequately controlled during waste generation to allow the use of
information as opposed to empirical investigation. The DOE has proposed eight classes of
acceptable knowledge (DOE95c). These are summarized in Table 4-8.
For newly generated and retrievably stored wastes, DOE plans to assign waste containers to
a waste stream based on process knowledge after first developing a profile for each waste
stream (DOE94b). DOE describes this approach in the TRU QAPP. Waste stream profiling
assumes that the waste-generating process is a well-defined and controlled process that can be
supported by sufficient documentation, and that the documentation is available and amenable
to direct inspection.
Table 4-8. Classes and Examples of Acceptable Knowledge
0ass ;pf Acceptable Knowledge
Examples of Supporting; Information
waste generating process information
engineering and design information
supporting data
supplemental data
expert knowledge
standard industry practice information
compliance program information
program management information
process flow diagrams, documented inputs/outputs, process
controls, operating procedures
piping and glove box designs, equipment and holding tank
specifications
surrogate waste sampling and analysis data, comparable waste
stream analytical data
data obtained from research and development operations,
effluent monitoring data, product quality control data
personnel interviews, site inspections, test or research plans
vendor information, material safety data sheets, common
industrial operations or treatment practices
RCRA permits, safety analysis reports, chemical inventory
databases
Quality Assurance Plans, procurement documents, operating
procedures, waste certification procedures
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Important aspects of waste stream profiling include consideration of the following:
• whether multiple profiles are required for complex waste streams
• profile's responsiveness to changes in the waste producing process(es)
• quantification of the uncertainty associated with each part of the profile
• how each stream's profile would be determined, i.e., using average
concentration values of specific constituents, or by establishing a range of
acceptable concentrations
• reconciliation of a waste stream's profile with an out-of-specification analysis
of a specific drum originating in the stream
• protocol required when a waste stream was found to be outside of the profile
It should be noted that much of the waste for which DOE uses acceptable knowledge/process
knowledge as the main waste characterization tool originates from non-routine types of
activities that are not typically understood to be controlled processes, with well-defined feed
materials, intermediate products and outputs. Example are wastes from unscheduled
maintenance and the cleanup of chemical or radioactive spills. Acceptable knowledge/
process knowledge may be a poor choice as a waste characterization tool for these and other
similar types of waste.
4.5.3 Use of Acceptable Knowledge/Process Knowledge for TRU Inventory
The use of acceptable knowledge/process knowledge to characterize TRU wastes is
advantageous for several reasons:
• to minimize worker radiation exposure;
• the physical nature of many wastes (i.e., WMC Series 5000 and 7000 wastes)
does not lend itself to conventional SW 846 type analytical procedures; and
• many historical wastes were generated prior to the establishment of RCRA,
and are inadequately characterized according to current standards.
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As discussed earlier, DOE currently details its inventory of current and anticipated TRU
waste in the WTWBIR. The WTWBIR combines information from the following two
documents:
• Integrated Data Base for 1993: U.S. Spent Fuel and Radioactive Waste Inventories,
Projections, and Characteristics, DOE/RW-0006, Rev. 9, April 1994; and
• U.S. Department of Energy Distribute of the Phase II Mixed Waste Inventory Report,
May 1994 (MWIR).
The WTWBIR is currently considered the best source for information on DOE's inventory of
TRU waste. Process knowledge was used to generate much of the information in this
document particularly as the basis for calculating waste volumes and other data. Process
knowledge should be used with caution because DOE TRU waste generators exhibit great
diversity with respect to waste generation and handling. Additionally, uncertainty estimates
associated with process knowledge data and their application are not provided and it is
unclear that DOE has sufficiently quantified/evaluated these.
4.5.4 Evaluating the Use of Process Knowledge
The use of process knowledge as a predictive tool for TRU waste characterization has not
undergone rigorous regulatory scrutiny. Due to the nature of chemical analyses and the
complexity of assigning hazardous waste codes, it is important to assess the appropriateness
of comparing waste characterizations made with process knowledge to those made with
empirical sampling and analysis.27 For the purpose of this report, such comparisons have
been made and one is discussed below (EPA95).
DOE conducted a 2-year investigation of the correlation between process knowledge and
empirical sampling and analysis. This study was completed in 1985 and involved a total of
242 containers of TRU waste, which ranged from new (6 months old) to older waste (12
27 For example, certain waste streams are classified as hazardous solely by virtue of the presence of a
specific chemical(s) within the waste generating process (process knowledge), regardless of concentration. For
such listed wastes, the inability of a chemical analysis to detect the listed waste does not affect the waste
stream's classification as hazardous (EPA94). Waste streams are often assigned waste codes for characteristics
wastes (D Codes) in a conservative manner for the purpose of storage, meaning that if a waste generator thinks
there is a reasonable probability the waste could exhibit a specific D Code (process knowledge), it is assigned.
However, upon empirical testing, many wastes would not actually test hazardous for all of the D Codes they
had been assigned. In both of these examples, the comparison between the waste codes assigned using process
knowledge and empirical sampling and analysis is inappropriate.
4-57
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years old in 1983). Of these, 199 drums and 10 boxes were generated at the Rocky Flats
Plant and 33 drums were generated at Los Alamos National Laboratory. All containers were
initially shipped to the Idaho National Engineering Laboratory (INEL), where they were
assayed nondestructively using RTR. The study's objective was to collect information on gas
generation, evaluate various venting devices, examine waste for compliance with the WIPP-
WAC and evaluate the adequacy of nondestructive examination as a certification technique.
The two-volume document TRU Sampling Program: Volume I—Waste Sampling and Volume
II—Gas Generation Studies (CLE85) describes the study in detail and provides the
investigation' s results.
The waste containers had initially been "characterized" at the generator facility (RFP or
LANL) by the assignment of a Waste Content Code28 (see Appendix A to CLE85) prior to
shipment to INEL. At INEL, each drum was examined using real-time radiography and
radioassay by passive-active neutron counting and the results were recorded. The drums
were then shipped to the Rocky Flats Plant, where each drum was completely dismantled
within a hot cell for visual examination. The contents were emptied, weighed, and analyzed
by radioassay when appropriate, and all results were recorded. This study's main purpose
was to determine the adequacy of RTR as a nondestructive characterization technique.
However, it also provided an opportunity to evaluate the use of process knowledge as a
predictive tool. By comparing the content code assigned by the generator using process
knowledge against the "proper" content code assigned after complete hands-on examination
of the waste container (the equivalent of sampling and analysis), process knowledge can be
evaluated as a tool for assigning the correct content code. Toward this end, the data from
this investigation were analyzed statistically and the results are described below.
The Kappa statistic was used to evaluate how well process knowledge was able to classify
waste by content code compared to how well the codes would be expected to have been
assigned by chance alone (EPA95). In summary, process knowledge assigned content codes
much better than would be expected by chance alone, indicating that for these waste
containers, process knowledge was effective as a predictive tool for waste classification. It
should be noted that DOE's proposed use of process knowledge may not lend itself to this
type of verification, in large part because problematic sample matrices do not permit
comparisons to be made with sampling and analysis results. DOE has recognized this
28 The Waste Content Codes used for this study were developed prior to TRUCON Codes. TRUCON
Codes were intended to include all aspects of waste covered by the Waste Content Codes.
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problem with debris wastes (WMC 5000 series) where process knowledge is the preferred
waste characterization technique.
In evaluating the CLE85 study, three caveats should be noted -
1) The waste containers used in the study were not statistically selected and therefore were
not necessarily representative of TRU waste, thus limiting the study's applicability.
2) Production and waste handling practices, documentation protocols, assay methods, etc.,
vary among TRU generators. Because of the lack of established, auditable, uniform criteria
for waste characterization by all TRU generators, questions exist regarding this study's
applicability. Caution must be exercised in applying conclusions to TRU waste generators or
specific waste streams other than those used in this study which originated primarily from
Rocky Flats and Los Alamos.
3) This analysis provides information on the ability of process knowledge to assign a content
code; no conclusions can be drawn about the ability of process knowledge to provide other
important, detailed information (e.g., isotopic distribution, amount of free liquids, gas
generation rates). This is particularly true for retrievably stored, older waste, where existing
information is sparse.
The study discussed here is the only documented evaluation of the use of process knowledge
available at this time. However, additional information exists at INEL, Hanford, Oak Ridge
National Laboratory, and the Savannah River Site, where DOE contractors have been
attempting to verify waste content codes assigned by process knowledge using other
techniques. At INEL alone, DOE has performed radiography and radioassay on
approximately 30,000 drums of waste to date, some percentage of which have also been
visually examined. The information is not yet available so it is not known what level of
documentation exists for these examinations or if other formal comparisons have been made.
This information could be very useful to a more comprehensive evaluation of the use of
process knowledge as a predictive tool.
4.6 TECHNICAL RATIONALE FOR WASTE CHARACTERIZATION PROVISION OF
40 CFR part 194
4.6.1 General Information on Waste
Section 194.24(a) of the rule requires that DOE provide information on the chemical,
radiological, and physical composition of the waste scheduled for disposal at the WIPP
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including both existing and, to the extent practicable, to-be-generated waste. This description
can be based on assays, non-destructive examination, process knowledge and any other
appropriate evaluation techniques. This information is needed to anticipate the behavior of
the waste in the disposal system.
Description of the radiological composition requires, for each radionuclide present or
expected, an estimate of the quantity of radioactivity (curies) at the time of disposal (i.e.,
when the disposal system is sealed). This could involve setting an upper limit for each
nuclide. Demonstration that the waste meets the TRU criterion of 100 nCi/g is also
required. In addition, information on the expected drum-to-drum variation in radioactivity
levels may be required to model cuttings releases associated with drilling events.
Description of the chemical composition would involve documentation of components which
might affect waste containment by affecting waste solubility, colloid formation, gas
generation, or gas consumption, inter alia. As has been discussed previously, solubility can
be affected by the quantity of organic ligands, the quantity of CO2-forming constituents, and
the quantity of waste constituents which can alter the pH of any intruding brines. To
characterize gas generation potential, it is necessary to know the quantities of iron and
aluminum alloys, the quantities of combustibles, plastics and rubber, and the quantity of
water initially present in the waste.
Description of the physical characteristics of the waste would include information on surface-
to-volume ratios of corroding metals, waste density and porosity, waste permeability, weight
or volume mix of waste forms such as sludges, metals, paper, rags, etc.
4.6.2 Documentation of Waste Characteristics
§194.24(b)(l) further requires that DOE submit documentation substantiating that all waste
characteristics which influence the containment of wastes in the disposal system have been
identified and assessed for their impact on disposal system performance. The rule lists, but
does not limit the characteristics to such items as solubility, gas generation, ability to form
stable colloids, shear strength, and compactability as examples of waste characteristics which
must be assessed.
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The waste shear strength (shear resistance) used in modeling borehole wall erosion during
drilling events was originally deduced by SNL investigators from seabed data, which showed
that the shear resistance of such materials was between 1 and 5 Pa, a range quoted to be
several orders of magnitude lower than macroscopic soil shear strength (PAR70, BER95).
For the 1992 WIPP PA, SNL assigned a range of 0.1 to 10 Pa and a median value of 1 Pa
for the shear resistance based on the assumption that the waste would behave similarly to
montmorillonite clay (BER95, SAR73). However, this parameter was not sampled over the
assigned range in the PA calculations; rather the median value of 1 Pa was used in the
CUTTINGS model.
If the flow of the drilling mud between the drill collar and the borehole wall is turbulent
rather than laminar, an additional waste characteristic—the surface roughness—is required to
calculate the shear stress acting on the waste. In the 1992 PA, the range of expected surface
roughness was set at 0.025 to 0.04 m, with a median value of 0.01 m (SAN92). The
absolute surface roughness values chosen for PA exceeded those of very rough concrete or
riveted steel piping (BER94, STR75).
4.6.3 Documentation of Waste Components
Section 194.24(b)(2) requires DOE to submit documentation substantiating that all waste
components which influence the waste characteristics described above in 4.6.2 have been
identified, and their impact on disposal system performance assessed. The waste components
to be evaluated include, inter alia, metals, cellulosics, chelating agents, water, and total
activity (in curies) for each radionuclide present in the waste. Other waste components not
specifically listed in the rule which may need evaluation include waste mix (by weight,
volume, and/or density), quantities of rubber and plastics, quantities of pH altering
constituents, quantities of CO2-forniing and -consuming species, and container-to-container
variability in radioactivity level. A summary of waste characteristics likely to be used in
performance assessment and the waste components which influence them is presented in
Table 4-9. It should be noted that, hi many cases, there is no single companion waste
component for a waste characteristic. This is because the characteristics are in many cases
determined by laboratory experiments which cannot be directly related to on-going
measurements of the waste.
4-61
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Table 4-9. Summary of Waste Characteristics and Waste Components Likely
to be Used in WIPP Performance Assessment
Waste Characteristic
Waste Component Influencing Waste
Characteristic
ROOM CLOSURE
Shear modulus
Bulk unloading modulus
Yield function constants
Volumetric strain vs. pressure
Initial waste density
Waste mix
Waste mix
CUTTINGS
Shear strength
Absolute surface roughness
Permeability
particle diameter
particle density
Tensile strength
BRAGFLO - Flow
Pore shape distribution parameter(s)
Residual saturations - liquid and gas
Threshold displacement pressure
Intrinsic permeability
Initial porosity
Waste mix
BRAGFLO - Gas Generation
Anoxic corrosion rates (humid and inundated)
Microbial degradation rates (humid and inundated)
Equivalent drum surface area
Number of drums per disposal room
Radiolysis rate
Quantity of iron (and aluminum)
Quantity of cellulosics
Quantity of plastics and rubbers
Quantity of electron acceptors (oxidants) such as
SO42- and NO3''
Quantities of nutrients (P and N)
Surface to volume ratio for iron (and aluminum)
Quantity of alpha emitters
Initial water content1
Quantity of cellulosics
ACTESTDE MOBILITY
Solubility - pCO2
Solubility - pH
Solubility - complexing agents
Solubility brine concentration
Solubility actinide oxidation states
Colloid concentration(s)
Quantities of CO2-forming and CO2-consuming
species
Quantities of acid and base formers
Quantities of complexing agents
SOURCE TERM
Radioactivity
Actinide concentration
Quantity of curies for each radionuclide
Drum-to-drum curie distribution
Quantity of each actinide
1 Influences all gas generation mechanisms.
4-62
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In addition to identifying and assessing the impact on disposal system performance of all
waste components which influence waste characteristics, DOE is required under
§194.24(b)(3) to substantiate any decision to exclude consideration of any waste characteristic
or waste component because such characteristic or component is not expected to significantly
influence the containment of the waste in the disposal system.
4.6.4 Limits on Waste Components
DOE is required to set limits on all significant waste components and show that the WIPP
complies with §194.34 and §194.55 based on these limits (§194.24(c)). In doing this, DOE
must describe the basis for setting these limits and demonstrate that, when all of these waste
component parameters are set at their limit, the mean CCDF obtained will meet the
containment requirements of 40 CFR part 191.13 at the 95% confidence limit.
As discussed previously in Sections 4.3.1.1 and 4.3.5, actinide solubility depends on various
factors including pH, pCO2, and presence of organic ligands. If the quantities of pH altering
species, CO2-forming and -consuming species and organic ligands in the waste are
determined to be important, DOE is required to set limits on these waste components and
demonstrate that the mean CCDF obtained when these components are set at the conservative
limits meets the requirements of §194.34. For example, CO2 tends to stabilize plutonium in
the + VI valence state which has high solubility, but CO2 can be removed from the system by
reaction with lime or calcium hydroxide. Thus, the conservative limits would be those
quantities of materials which produce the maximum amount of C02 and result in the least
CO2 removal in this specific example with respect to plutonium.
Once the acceptable limits on the waste components have been set, DOE must establish a
system of controls which assures that the waste actually emplaced hi the WIPP will fall
within these limits. Elements of this system of controls include measurement, sampling,
chain of custody, and other documentation. If, as discussed hi Section 4.6.1, DOE sets an
upper limit on the quantities of each radionuclide, then it will be necessary to show during
disposal, that this limit will not be exceeded taking into account uncertainties in the
measurements of the curie-content of the waste at the various generator sites.
4-63
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4.6.5 Quality Assurance
As discussed previously in Section 4.3.7, the components of actual waste may differ
significantly from the components that were assumed in developing the waste characteristics
for the compliance application. This is especially true, since only about 40% of the total
CH-TRU waste has been generated to date. The provisions of §194.24 were developed to
ensure that the repository will remain in compliance as long as the waste emplaced is within
the established limits. EPA believes that the proposed procedure for bounding the waste
characteristics is not overly prescriptive and can be addressed within the sensitivity analysis
framework which is an integral part of performance assessment. To enhance confidence in
the waste characterization process, all activities and assumptions are subject to the quality
assurance requirements of §194.22. Use of process knowledge to quantify waste components
is specifically subject to these quality assurance requirements (§194.24(c)). EPA is
empowered to use audits and inspections to ensure that the quality assurance requirements are
met (§194.24(h)).
4.7 REFERENCES
ANL94 Argonne National Laboratory-East Environment and Waste Management
Division, "Draft Site Treatment Plan for Argonne National Laboratory-East,"
August 1994.
BER94 Berglund, J.W., "Direct Releases Due to Drilling," New Mexico Engineering
Research Institute, Presentation at Joint DOE/EPA Technical Exchange,
February 23, 1994.
BER95 Berglund, J.W., "Theoretical Reference Manual for CUTTINGS_S: A Code
for Computing the Quantity of Wastes Brought to the Ground Surface when a
Waste Disposal Room of a Radioactive Waste Repository is Inadvertently
Penetrated by an Exploratory Borehole," DRAFT, New Mexico Engineering
Research Institute, SAND95-XXXX, 1995.
BRU94 Brush, L., "Position Paper on Gas Generation in the Waste Isolation Pilot
Plant" (Draft), Sandia National Laboratories, November 15, 1994.
BYN95 Bynum, V., C. Novak, and P. Vaughn, "Actinide Source Term Program:
Performance Assessment and Experimental Perspectives," Sandia National
Laboratories, presented at Joint DOE/EPA Technical Exchange, August 30-31,
1995.
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CLE85 Clements, T., and D. Kudera, "TRU Sampling Program: Volume I—Waste
Sampling and Volume II—Gas Generation Studies," Idaho National
Engineering Laboratory, EGG-WM-6503, September 1985.
DOE87 U.S. Department of Energy, "Assessment of Allowable Transuranic Activity
Levels for WIPP Wastes," DOE/WIPP 87-014 (formerly WTSD-TME-062),
December 1987.
DOE91 U.S. Department of Energy, "Waste Acceptance Criteria for the Waste
Isolation Pilot Plant," DOE/WIPP-069, Revision 4, 1991.
DOE92 U.S. Department of Energy, "Performance Demonstration Program for the
WIPP Experimental-Waste Characterization Program," DOE/WIPP 91-016,
Revision 2, February 1992.
DOE93a U.S. Department of Energy, "Comparative Study of Waste Isolation Pilot
Plant (WIPP) Transportation Alternatives," DOE/WIPP 93-058, February
1994.
DOE93b U.S. Department of Energy, "Resource Conservation and Recovery Act part B
Permit Application," DOE/WIPP 91-055, Revision 3, Vol. I, January 1993.
DOE94a U.S. Department of Energy, "National Summary Report of Draft Site
Treatment Plans," Final Draft, Volumes I and II, November 14, 1994.
DOE94b U.S. Department of Energy, "TRU Waste Characterization Quality Assurance
Program Plan," CAO-94-1010 Draft, Revision B., July 8, 1994.
DOE94c U.S. Department of Energy, "Distribute of the Phase II Mixed Waste
Inventory Report Data," memo from Patty Bubar to distribution, May 17,
1994.
DOE94d U.S. Department of Energy, "Waste Isolation Pilot Plant Transuranic Waste
Baseline Inventory Report," Carlsbad Area Office, CAO-94-1005, Revision 0,
June 1994.
DOE94e U.S. Department of Energy,"Quality Assurance Program Description,"
Carlsbad Area Office, CAO-94-1012, Revision 0, June 1994.
DOE95a U.S. Department of Energy, "Waste Isolation Pilot Plant Transuranic Waste
Baseline Inventory Report," CAO-94-1005, Revision 1, February 1995.
DOE95b U.S. Department of Energy, "Remote-Handled Transuranic Waste Study,"
DOE/CAO-DRAFT-10905, July 1995.
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DOE95c U.S. Department of Energy,"Predecisional Draft Transuranic Waste
Characterization Acceptable Knowledge Guidance Document," Carlsbad Area
Office, (not numbered), August 1995.
DOE95d U.S. Department of Energy, "Transuranic Waste Characterization Program
Sampling and Analysis Methods Manual," DOE/WIPP-91-043, May 1995.
EPA86 U.S. Environmental Protection Agency, "Test Methods for Evaluating Solid
Waste (SW846): Physical/Chemical Methods," Third Editiion and Revision,
1986.
EPA92 U.S. Environmental Protection Agency,"No Migration Variances to the
Hazardous Land Disposal Prohibitions: A Guidance Manual for Petitioners,"
Draft, EPA 530-R-92-023, July 1992.
EPA94 U.S. Environmental Protection Agency, "Waste Analysis at Facilities That
Generate, Treat, Store, and Dispose of Hazardous Wastes - A Guidance
Manual," OSWER 9938.4-03, April 1994.
EPA95 U.S. Environmental Protection Agency, "Criteria for the Certification and
Determination of the Waste Isolation Pilot Plant's Compliance with 40 CFR
part 191: Background Information for the Proposed 40 CFR part 194," EPA
402-R-95-002, January 1995.
IDA94 Idaho National Laboratory, "Idaho National Engineering Laboratory Draft Site
Treatment Plan," DOE/ID-100453, August 31, 1994.
LAB95 Labreche, D.A. et al., "A Sensitivity Analysis of the WIPP Disposal Room
Model: Phase I," SAND94-0890, Sandia National Laboratories, July 1995.
LOS93 Los Alamos National Laboratory, "Test Plan for Actinide Source-Term Waste
Test Program (STTP)," Document No. CLSI-STP-SOP5-012/0, May 1993.
MAR95 Personal Communication, P. Kelly, S. Cohen and Associates, Inc., with M.
Martin, Lockeed Idaho Technical Company, August 31, 1995.
NOV94a Novak, C.F. et al., "Actinide Source Term Program Position Paper," Revision
1, Sandia National Laboratories, November 15, 1994.
NOV94b Novak, C.F., editor, "Actinide Chemistry Research Supporting the Waste
Isolation Pilot Plant (WIPP): FY94 Results," Sandia National Laboratories,
SAND94-2274, August 1995.
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NOW95 Nowak, J. and P. Vaughn, "SNL Gas Generation Studies: Performance
Assessment and Experimental Perspectives," Sandia National Laboratories,
presented at Joint DOE/EPA Technical Exchange, August 30-31, 1995.
NTS94 "Nevada Test Site Draft Site Treatment Plan," Revision 0, U.S. Department of
Energy Nevada Operations Office, August 1994.
OAK94 Oak Ridge Operations, "Draft Site Treatment Plan for Mixed Waste on the
U.S. Department of Energy Oak Ridge Reservation," U.S. Department of
Energy, DOE-OR/2016, August 1994.
PAR70 Parthenaides, E., and R.E. Paaswell, "Erodibility of Channels with Cohesive
Boundary," in Proceedings of the American Society of Civil Engineers,
Journal of Hydraulics Division. 99:555-558, 1970.
REE94 Reed, D.R., et al.,"Stability of Plutonium (VI) in Selected WIPP Brines,"
Sandia National Laboratories, SAND 93-7114C, 1994.
RFP94 Rocky Flats Field Office, "Rocky Flats Environmental Technology Site Draft
Site Treatment Plan," Compliance Plan Volume, Revision 1.5, U.S.
Department of Energy, August 31, 1994.
SAN92 Sandia National Laboratories, "Preliminary Performance Assessment for the
Waste Isolation Pilot Plant," Sandia WIPP Project, SAND92-0700, December
1992.
SAN93 Sandia National Laboratories, "Technical Requirements for the Actinide
Source-Term Waste Test Program," SAND91-2111, October 1993.
SAR73 Sargunam, A. et al., "Physico-Chemical Factors in Erosion of Cohesive
Soils," in Proceedings of the American Society of Civil Engineers, Journal of
the Hydraulics Division, vol. 99, No. HY3:555-558, March 1973.
STR75 Streeter, V.L., and E.B. Wylie, Fluid Mechanics. 6th ed. McGraw-Hill Book
Co., New York, NY, 1975.
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5. Future State Assumptions
5.1 INTRODUCTION
A framework, or set of bounding assumptions, is required in order to limit speculation in
performance assessments and compliance assessments about how changes that occur over
time may effect the WIPP disposal system. The "Future State Assumptions" are designed to
establish that framework by providing guidance on how to treat future uncertainties such as
changes in demographics, changes in human physiology, changes in technology, advances in
medical science. §194.25 states:
(a) Unless otherwise specified in this part or in the disposal regulations, performance
, assessments and compliance assessments conducted pursuant the provisions of this
part to demonstrate compliance with §191.13, §191.15 and part 191, subpart C shall
assume that characteristics of the future remain what they are at the time the
compliance application is prepared, provided that such characteristics are not related
to hydrogeologic, geologic or climatic conditions.
(b) In considering future states pursuant to this section, the Department shall
document hi any compliance application, to the extent practicable, effects of potential
future hydrogeologic, geologic and climatic conditions on the disposal system over the
regulatory tune frame. Such documentation shall be part of the activities undertaken
pursuant to § 194.14, Content of compliance certification application; § 194.32,
Scope of performance assessments; and § 194.54, Scope of compliance assessments.
(1) In considering the effects of hydrogeologic conditions on the disposal system, the
Department shall document in any compliance application, to the extent practicable,
the effects of potential changes to hydrogeologic conditions.
(2) In considering the effects of geologic conditions on the disposal system, the
Department shall document hi any compliance application, to the extent practicable,
the effects of potential changes to geologic conditions, including, but not limited to:
dissolution; near surface geomorphic features and processes; and related subsidence in
the geologic units of the disposal system.
(3) In considering the effects of climatic conditions on the disposal system, the
Department shall document in any compliance application, to extent practicable, the
effects of potential changes to future climate cycles of increased precipitation (as
compared to present conditions).
The final rule requires that performance assessments and compliance assessments shall
include dynamic analyses of geologic, hydrogeologic and climatic processes and events that
will evolve over the 10,000-year regulatory tune frame. All other present day conditions
5-1
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will be assumed to exist in their present state for the entire 10,000-year regulatory time
frame. These latter requirements apply to the future demographic and physiologic state of
mankind, among all other assumptions necessitated by performance assessments and
compliance assessments. Predicting the manner in which society will change would
necessitate predicting the effect of historical, economic and political forces which typically
bring about societal and demographic change. The speculative nature of such predictions
precludes the development of an acceptable methodology for inclusion in the final rule that
could make reliable predictions of the future state of society, science, languages or other
characteristics of future mankind. For example, suppose that we know the current population
density around the WIPP and we know that the population has grown at a certain rate over
the last three decades (a very short time-frame when compared to the WIPP regulatory time-
frame of 10,000 years). In addressing the future, one could extrapolate future population
density based on the historical growth rate which might, over a 10,000-year regulatory
period, result in unreasonable values, or one could assume that the population density
remains constant (i.e., the status quo). The latter approach is to be used for future states. It
is inappropriate to make long term predictions based on short term data. As hi the
population example above, certain activities are influenced by complex and interrelated forces
(economics, government policy, etc.) and therefore cannot be predicted with reasonable
accuracy.
Effects of climatic change are to be considered because they are reasonably predictable from
the geologic record of the last several thousand years. However, there is no need for
speculation on the possible secondary effects from climate change (e.g., increased
precipitation which could allow irrigated agriculture near the site and change human
economic activity and lifestyle) as these would be driven by the complex interrelated forces
listed above. A similarly convincing basis exists as well for hydrologic and geologic
conditions. This chapter will examine the body of scientific knowledge which may be used
to extrapolate hydrologic, geologic and climate conditions into the 10,000 year regulatory
time frame.
5.2 LAND USE AROUND THE WIPP SITE
Analyses of the WIPP's long-term performance will have to establish population and land use
characteristics. The information provided below demonstrates how certain aspects of
present-day demographics can be established and described for use in these analyses.
5-2
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Population
The 1979 resident population within fifty miles of the WIPP Site is shown in Table 5-1.
These data were estimated for the 1980 WIPP Final Environmental Impact Statement (FEIS)
(ALA79, DOE80). An examination of the US Census data indicated that the total Eddy
County and Lea County population grew from 90,363 to 104,370 between 1970 and 1990
(DOC90). This is an average of about 0.7% per year. The 1990 Census data examined
were not detailed enough to determine if there were changes in the population within 10
miles of the WIPP Site since 1979.
Table 5-1. 1979 Resident Population Within 50 Miles of the Sitea-b
Sector0
North
North-northeast
Northeast
East-northeast
East
East-Southeast
Southeast
South-Southeast
South
South-southwest
Southwest
West-southwest
West
West-northwest
Northwest
North-northwest
Radius Total
Cumulative Total
0-5
0
0
0
0
0
0
0
0
0
6
0
0
0
0
0
0
6
6
5-10
0
0
0
0
0
• o
0
0
0
0
0
0
0
10
0
0
10
16
10-20
35
25
0
15
5
5
5
0
5
5
55
1,750
70
5
30
15
2,025
2,040
20-30
25
5
25
70
15
10
15
25
15
30
30
200
31,780
190
20
5
32,460
34,500
30-40
175
55
75
205
3,240
3,035
25
10
55
90
10
50
40
55
65
220
7,440*
41,940
40-50
25
5,610
8,660
33,200
155
295
30
40
15
15
45
65
35
50
12,055
10
60,305
102,245
Total
260
5,695
8,760
33,490
3,415
3,345
75
75
95*
145
140
2,065
31,925
310
12,170
280*
102,245
Population estimated by Adcock and Associates (1977-1979).
Figures for all areas beyond the 10 mile radius have been rounded to the nearest five.
Distance from site (miles).
The totals for this column and these rows are not in agreement with die numbers shown due to errors
in the original data source.
5-3
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Land Ownership and Use
Land in the vicinity of the WIPP is primarily owned by the Bureau of Land Management
(BLM). Figure 5.1, included here from the WIPP FEIS (DOE80), illustrates land ownership
for 1980. Since 1980, the following changes occurred in the Township 22 South Range 31
East (where the WIPP site is located): 1) the state-owned sections 16 and 32 were
transferred to BLM in exchange for federal land elsewhere, and 2) the 4 by 4-mile area in
the southwestern portion of the township was withdrawn for the WIPP site (see attached
Figure 5.2 from DOE93).
At the WIPP, within its 16 mi2 site, DOE presently controls 35 acres around the site's shafts
and buildings (see Figure 5.2). Neither trespassing nor non-project uses are permitted in this
area. Not shown in the figure is a 300-acre area around which DOE eventually plans to
erect a five-strand barbed wire fence to prevent access to the area which overlies the
repository footprint. DOE is presently considering extending the no-trespassing area to a
total of 1,454 acres (see Figure 5.3), although grazing would be permitted on the newly
added 1,154 acres.
Grazing and recreational uses - hunting, trapping, and off-road vehicle use - will be
permitted on the remainder of the 16 mi2 site. No surface or subsurface mining or
exploration, nor water well drilling, will be permitted anywhere on the site except for two
existing gas leases in section 31 at the southwest corner of the site. Land use immediately
off the WIPP site allows for grazing, oil and gas exploration and production, extraction of
sand, gravel, and caliche from surface pits, and recreational use. Figure 5.3 from the FEIS
shows the location of some of these various activities. There is also extensive potash mining
to the west, north, and northwest of the WIPP site. Figure 5.3 illustrates the locations of
active potash mines relative to the WIPP site. Non-resident occupational employment within
ten miles of the WIPP site is estimated to be 360 potash workers per shift (three mines) and
twenty-four workers on cattle leases (ALA79, DOE90). Oil industry employment has not
been estimated, but in 1990-1992 there were seventy-five oil and gas wells drilled within two
miles of the WIPP site boundary (SIL94). A large agricultural area supported by irrigation
is located along the Pecos River, which is 15-20 miles to the west and southwest of the
WIPP (see Figure 5.3). Because of a scarcity of water in the fully allocated Pecos River
Basin, the irrigated area is not likely to increase significantly.
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; m **•»
NmMhxwo I '
10 Mites
Built-up areas: towns with more than 5,000 inhabitants
State lands
Private lands
Public lands
Boundary between BLM administrative districts
Figure 5.1 Land Ownership Within 30 miles of the WIPP Site
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N
t
34
10
35
T21S
T22S
11
36
31
12 7
(T) Land Withdrawl
Boundary ^
32
33
34
u
o
o
<
10
35
11
36
12
31
.-o
OS
(56
15
14
13
18
17
^) Property
; Protection \ 5
--~ Area
14
18
22
27
10
15
23
34 35
11
14
24
19
26 25
30
20
21
22
23
29
36
31
32
13
17
27
26
34
35
Proposed DOE
^Off-limits Area
-. X
4
16
10
15
14
o
0
24 >-
-o
UJ
o
a 19
«j
25
30
36
Ci
12
13
18
1. Major WIPP facility enclosed by an 8-foot, chain link fence and barbed wire outris"ers
OO
2. Can be posted against trespass
3. Will be posted by "Information Signs" only
Figure 5.2 WIPP Land Withdrawal Area and Surroundings
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10 Miles
Irrigated
farmland
Grazing rangeland
General areas of
gas production
General areas of
oil production
Location of
Potash Mines
Sources: Bureau of Reclamation Land Use Map. 1968; New Mexico
Bureau of Mines and Mineral Resources, New Mexico
Energy Resources Map. 1974; Adcock and Associates, field.
surveys, 1978.
Figure 5.3 Land Use Within 30 Mile Radius of the WIPP Site
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There are several highways near the WIPP site. US 62/180, connecting Carlsbad and
Hobbs, is a four-lane divided highway located about nine miles north of the site which
carries an average daily traffic flow of 1,850 vehicles (this and following averages were
recorded in 1978). NM 128, about three miles south of the site, connects Carlsbad to Jal,
New Mexico. This two-lane paved road conducts an average daily traffic flow of 220
vehicles. NM 31, about eight miles west of the site, connects NM 128 and US 62/180 and
averages 510 vehicles per day. Numerous dirt roads in the area are maintained for ranching,
pipeline maintenance, and oil and gas site access. In addition, there are now paved north
and south access roads to the WIPP site from US 62/180 and NM 128, respectively Most
daily commuter traffic to the WIPP site uses the South Access Road. Present plans indicate
all waste shipments arriving by truck will use the North Access Road. There is also a
railroad spur connecting the WIPP site, but there are presently no plans for its use.
5.3 FUTURE STATES OF CLIMATE
Geologic, hydrologic, or climatic conditions are the only assumptions required by §194.25 to
be predicted into the future. This section explicates part of the scientific record which can
provide a basis for informed prediction of these three categories of events.
General State of Knowledge
Paleoclimatic data from southeastern New Mexico and the surrounding area indicate that the
wettest and coolest Quaternary climate at the site can be represented by the last glacial
maximum, when mean annual precipitation was approximately twice that of the present
(SWI94). These data indicate that the hottest and driest climates have been similar to those
of the present. The report also states that "the regularity of global glacial cycles during the
late Pleistocene confirms that the climate of the last glacial maximum is suitable., for use as a
cooler and wetter bound for variability during the next 10,000 years."
Mean annual precipitation at the WIPP has been estimated to be between 28 and 34 cm/yr
(HUN85). Geologic data from southeastern New Mexico and the surrounding region show
repeated alternations of wetter and drier climates throughout the Pleistocene, corresponding
to global cycles of glaciation and deglaciation. Data from plant and animal remains and
paleo-lake levels permit quantitative climate reconstructions for the region only for the last
glacial cycle, and confirm the interpretation that conditions were coolest and wettest during
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glacial maximums (SWI93). Mean annual precipitation 22,000 to 18,000 years ago, when
the last North American ice sheet reached its southern limit roughly 1500 km north of the
WIPP, was approximately twice that of the present (SWI94).
The following text is quoted from SWI94, "Incorporating Long-Term climate Change in
Performance Assessment for the Waste Isolation Pilot Plant." It provides a general summary
of the investigations which provide the basis for the 1992 WIPP PA assumptions regarding
the potential for the range of future climatic extremes (SAN92).
Glacial periodicities have been stable for the last 800,000 years (MIL41, HAY76,
IMB84, IMB85). Barring anthropogenic changes in the Earth's climate, relatively
simple modeling of climatic responses to earth's orbital changes suggest that the next
glacial maximum will occur in approximately 60,000 years (IMB80). The extent to
which unprecedented anthropogenic climate changes may alter this conclusion is
uncertain, but presently available models of climatic response to an enhance
greenhouse effect (MIT89, HOU90) do not predict changes of a larger magnitude than
those of the Pleistocene. Furthermore, published models do not suggest significant
increases in precipitation in southeastern New Mexico following global warming
(WAS84, WIL87, SCH87, HOU90). Even allowing for anthropogenic change,
climate variability at the WIPP can be bounded by Pleistocene extremes (SWI93).
The estimated mean annual precipitation at the WIPP during the late Pleistocene and
Holocene is shown on Figure 5.4.
SWI93 draws the following three conclusions regarding climatic trends. First, maximum
precipitation in southeastern New Mexico in the past coincided with the maximum advance of
the North American ice .sheet. (Minimum precipitation occurred after the ice sheet had
retreated to its present limits.) Second, past maximum long-term average precipitation levels
were roughly twice present levels. Minimum levels may have been 90% of present levels.
Third, short-term fluctuations in precipitation have occurred during both the glacial
maximum and the present, relatively dry, interglacial period, but fluctuations during the
present interglacial period have not exceeded the upper limits of the glacial maximum.
SWI93 also states: "It would be unrealistic to attempt a direct extrapolation of precipitation
(a figure is referenced} into the future. Too little is known about the relatively short-term
behavior of global circulation patterns, and it is at present impossible to predict the
probability of a recurrence of a wetter climate such as that of approximately 1000 years ago.
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Estimated Average Annual Precipitation
c
_o
03
"g.
"o
03
c
c
0)
CD
03
k—
CD
<
100
90
80
70
60
50
40
20
10
0
30
Time of Maximum Advance
of Late Wisconsin Ice Sheet
J_
I
I
20 10
Thousands of Years before Present
0
Figure 5.4 Estimated Mean Annual Precipitation Rate at the WIPP During the Late
Pleistocene and Holocene (SWI93)
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The long-term stability of the patterns of glaciation and deglaciation, however, do permit the
conclusion that future climatic extremes are unlikely to exceed those of the late Pleistocene.
Furthermore, the periodicity of glacial events suggests that a return to full glacial conditions
is highly unlikely within the next 10,000 years."
Glaciation
Southeastern New Mexico is far from any region where extensive Pleistocene continental
glaciation occurred. It is highly improbable, even in the event that global "icehouse"
conditions developed, that the WIPP site would be affected by continental glaciers. In turn,
the probability of glaciers and glacial erosion directly affecting the WIPP site is extremely
small. Alpine glaciation, however, was quite extensive in northern New Mexico during the
Pleistocene (CHR87). Evidence for alpine glaciers extends down to elevations of at least
8000 feet, possibly less. These glaciers, their deposits, and meltwaters affected and continue
to influence the regional hydrology and Quaternary stratigraphy of this region. In the event
of a major climatic change, alpine glaciation might be possible in the Guadalupe Mountains
which have maximum elevations greater than 8000 feet (Guadalupe Peak is 8751 feet high).
5.4 GEOLOGIC FUTURE STATES
Sea-level fluctuations and hurricanes/seiches/tsunamis
The likelihood of sea-level fluctuations and hurricanes/seiches/tsunamis can be considered to
be small due to the elevation (-3,300-3,500 feet) and landlocked position of the WIPP site.
These conditions ensure that it will neither be inundated in the event of a eustatic sea-level
rise (not even a rise of unprecedented scale) nor will it be affected by any catastrophic ocean
current.
Regional uplift and subsidence
The Rio Grande Rift, an elongate, fault-bounded, extensional feature that extends roughly
north-south across central New Mexico, began to form about 30 million years ago. One
segment of the Rio Grande Rift, called the Tularosa Basin, is located within 150 km of the
WIPP site. Extension continues within the rift today, as expressed by numerous active fault
scarps that cut through Quaternary deposits (BLA76; CHA79; OLD89). There is some
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suggestion, based on the timing of development of structural features across the region, that
extension may be slowly propagating eastward, toward the WIPP site. While long-term,
rates of regional uplift/subsidence may not increase drastically over the next 10,000 years,
detailed studies of Quaternary seismic activity and co-seismic fault slip should be done in the
vicinity of the WIPP site in order to substantiate the claim that episodic, short-term
uplift/subsidence will not compromise the integrity of the repository. This will largely entail
detailed field mapping and analyses of Quaternary stratigraphy and structural features.
Landslides
The Pecos River, the largest river in the vicinity of the WIPP site, flows within 12-15 miles
of the site. The difference in elevation between the Pecos River and the WIPP site is
approximately 400 feet. It should be feasible to determine whether inundation of the WIPP
site is possible in the event that a landslide dams or diverts the river at any of several
different points along the river. A worst-case scenario might involve: 1) a landslide that
dams the river where it is narrow, slightly downstream from, yet close to, the WIPP; 2) the
landslide occurs during the annual peak discharge of the Pecos River; 3) the landslide and
damming of the river occur at the same time as major flooding in southeastern New Mexico;
and 4) the landslide occurs at a place along the river where the adjacent topography is such
that flood waters are preferentially funneled toward the WIPP. Examination of aerial
photographs and land-based studies might allow determination of the frequency of landslide
events along the banks of the Pecos River and whether or not the (paleo)Pecos River ever
flooded areas far beyond its "historical" flood plain. Furthermore, under pluvial conditions,
the Pecos River may not maintain its current course in the future. Cores and/or trenches
through Quaternary alluvium also might help with determining the frequency and magnitude
of flooding events near the WIPP site.
Seismic activity (and faulting)
The main concerns regarding seismic activity and co-seismic slip or initiation of faults are:
1) whether the fault cuts through the repository and cumulative slip along the fault then
brings radioactive waste into physical contact with circulating ground water; or 2) whether
seismic events cause permeable faults or fracture zones to develop which lead to hydrologic
communication between waste in the repository and circulating ground-water.
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Given the tectonic setting of the WIPP site and the proposed depth of the repository, it is
highly improbable that cumulative slip on a hypothetical fault that cuts through the repository
could bring waste into contact with aquifers that sandwich the repository within 10,000
years. Thus, it seems reasonable to eliminate this from geologic scenarios, although as
mentioned above under "Regional uplift and subsidence," detailed studies of the neotectonic
activity around the WIPP site seem warranted.
New faults or fractures also could breach the repository and allow circulating ground waters
to move through the repository and transport radionuclides to the surface or to shallow
aquifers.
Volcanic and magmatic activity
Given the following, however, it seems reasonable that the inception of volcanic or magmatic
activity is unlikely over the next 10,000 years and need not be considered in geologic
scenarios: 1) the WIPP site is located presently in a relatively tectonically quiescent setting;
2) the WIPP site probably will remain tectonically quiescent over the long-term, based on
present and projected vectors of motion for the North American plate and adjacent plates;
and 3) volcanic/magmatic activity is typically associated with active, plate-margin settings
(either extensional, compressional, or strike-slip settings), while mid-plate
volcanism/magmatism is much less common and is related to deep-seated, mantle processes.
The studies outlined previously under Regional uplift and subsidence would enhance
confidence in the conclusion that volcanic and magmatic activity need not be considered as
geologic scenarios.
Earthquakes produced by subsurface fluid injection/removal
Injection or removal of subsurface fluids during recovery of hydrocarbons is known to
produce earthquakes. Earthquakes with magnitudes as high as 5.0 on the Richter scale, but
generally between 1.0 to 3.0, are known from some areas (HOL68). Earthquakes of this
type have been recorded in the Permian Basin (largely anecdotal evidence). Hazard
assessment for the WIPP site should include the possibility of human-induced earthquakes as
there are productive oil fields near the WIPP site that are currently under waterflood. Even
small scale earthquakes could affect the integrity of seals.
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5.5 HYDROLOGIC FUTURE STATES
The hydrologic properties of the geologic strata within the disposal system can be changed
due to the occurrence of natural processes and events. As an example, dissolution may, in
the future, affect the hydrologic properties of the Culebra dolomite layer of the Rustler
Formation. The presence and degree of fracturing hi the Culebra dolomite is thought to be
directly related to the amount of dissolution of halite occurring below the Culebra (SNY85).
As the magnitude of fracturing and development of secondary porosity increases, the Culebra
transmissivity generally increases (CHA85). Based upon observations of outcrops, core, and
detailed shaft mapping, the Culebra can be characterized as a fractured medium, at least
locally, at the WIPP site (CHA84; HOL84). Aquifer tests also indicate responses
characteristic of a fractured media (BEA87).
Dissolution within the Rustler Formation is observed both at the surface within Nash Draw,
and in the subsurface at the WIPP site. Nash Draw, located immediately west of the WIPP
site, is a depression resulting from both dissolution and erosion (BAC81). In Nash Draw,
members of the Rustler are actively undergoing dissolution and locally contain caves, sinks,
and tunnels typical of karst morphology in evaporitic terrain (HAU87).
BAC80 identified three types of dissolution occurring in the Delaware Basin: local
dissolution, regional dissolution, and deep-seated dissolution. Local dissolution is the near-
surface dissolution where surface or ground waters penetrate soluble strata though joints or
fractures, causing local dissolution and possible collapse and fill, as well as dissolution
features such as shallow caves above the regional water table. Regional dissolution occurs
when chemically unsaturated water penetrates to permeable beds, where it migrates laterally,
dissolving the soluble units which it contacts. On a regional scale, the consequence of such
dissolution appears to be removal of highly soluble rock types, such as halite, combined with
displacement and fracturing of adjacent rocks. Deep-seated dissolution occurs well below the
water table, forming caverns within the rock.
At the WIPP site, regional dissolution is thought to have occurred within the Rustler
Formation in the past (SNY85). However, there is some controversy as to whether this
dissolution front is still active. BAC85 feels that most of the dissolution in the Rustler
predates or occurred during a much more humid time in southeastern New Mexico over
500,000 years before present. BAC85 does suggest, however, that dissolution is still active
in Nash Draw in areas very close to Livingston Ridge.
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In the Rustler Formation at the WIPP site, most investigators feel that a westward increase in
regional dissolution is reflected by a decrease in the number and thickness of halite beds and
subsequent thinning of the Rustler Formation (HAU87). The stratigraphic level of the first
occurrence of salt is in the upper Rustler along the eastern margin of the WIPP site, and
progressively moves down-section through the Rustler as one moves west. As the bedded
halites are dissolved, insoluble residues remain, forming beds of mudstones, siltstones, and
chaotic breccia with a clay matrix (HAU87). Halite beds in the non-dolomitic members tend
to be thin and grade westward into the residuum. Although most investigators concur with
the premise that a dissolution front exists in the Rustler Formation at the WIPP site (COO71,
POW78, MER83, CHA84, SNY85), there are some investigators who oppose this concept
and believe that the westward decrease in halite within the Rustler represents depositional
limits (LAM83, HOL84). HOL84 reported that, in their detailed mapping of the Rustler in
the waste-handling shaft, no post-depositional dissolution features were identified.
Whether or not the dissolution front hypothesis is correct, there are general trends associated
with the presence or lack of bedded halite within the Rustler Formation. As the presence of
bedded halite within the Rustler increases, so does the thickness of the formation. Generally,
as the amount of halite hi the Rustler decreases, the transmissivity of the dolomitic members
increases (HAU87), presumably from increased fracturing of the units as a result of halite
removal and subsequent foundering and collapse of the more competent dolomite beds. In
parts of Nash Draw, hydraulic potentials hi the Magenta and Culebra are essentially the same
(i.e. no vertical movement up or down). As one moves eastward onto the Livingston Ridge
surface, the difference in hydraulic potentials between the two units increases. This could
represent the increase hi the effectiveness of the Tamarisk Member as a confining unit (or
aquitard) with decreased halite removal (HAU87).
5.6 REFERENCES
ALA79 Adcock, Larry and Associates, "Demographic Analyses Utilizing Data
Obtained from Various Sources," 1979.
BAC80 Bachman, G.O., "Regional Geology and Cenozoic History of the Pecos
Region, Southeastern New Mexico," U.S. Geological Survey Open-File Report
D-77-946, Albuquerque, NM, 1980.
BAC81 Bachman, G.O., "Geology of Nash Draw, Eddy County, New Mexico," U.S.
Geological Survey Open File Report 81-31. Denver, CO, 1981.
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BAC85 Bachman, G.O., "Assessment of Near-Surface Dissolution at and near the
Waste Isolation Pilot Plant (WIPP), Southeastern New Mexico," Sandia
National Laboratories, Albuquerque, NM, SAND84-7178, 1985.
BEA87 Beauheim, R.L., "Interpretations of Single-Well Hydraulic Tests Conducted at
and Near the Waste Isolation Pilot Plant (WIPP) Site, 1983-1987," Sandia
National Laboratories, SAND87-0039, 169p., 1987.
BLA76 Black, B.A., "Tectonics of the northern and eastern parts of the Otero
Platform, Otero and Chaves Counties, New Mexico," in Woodward, L.A.,
and Northrop, S.A., eds., Tectonics and Mineral Resources of Southwestern
North America: New Mexico Geological Society Special Publication No. 6,
p. 39-45, 1976.
CHA79 Chapin, C.E., "Evolution of the Rio Grande rift - a summary," in, Riecker,
R.E., ed., Rio Grande Rift - Tectonics and Magmatism: American
Geophysical Union, p. 1-5, 1979.
CHA84 Chaturvedi, L. and K. Rehfeldt, "Groundwater Occurrence and the Dissolution
of Salt at the WIPP Radioactive Waste Repository Site," American
Geophysical Union, EOS, p. 457-459, July 3, 1984.
CHA85 Chaturvedi, L. and J.K. Channell, "The Rustler Formation as a Transport
Medium For Contaminated Groundwater," New Mexico Environmental
Evaluation Group, EEG-32, 85 p., 19985.
COO71 Cooper, J.B. and V.M. Glanzman, "Geohydrology of the Project Gnome Site.
Eddy County, New Mexico," U.S. Geological Survey Professional Paper 712-
A, 24p., 1971.
DOC90 US Department of Commerce, "County Business Patterns New Mexico -
1990" and "General Characteristics of Persons, Outside Metropolitan Areas -
New Mexico, 1990," Bureau of the Census.
DOE80 US Department of Energy, "Final Environmental Impact Statement Waste
Isolation Pilot Plant," DOE/EIS-0026, October 1980.
DOE90 US Department of Energy, "Final Supplemental Environmental Impact
Statement, Waste Isolation Pilot Plant," DOE/EIS-0026-FS, January 1990.
DOE93 US Department of Energy, "Draft Waste Isolation Pilot Plant Land
Management Plan," DOE/WIPP 93-004, undated.
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HAU87 Haug, A., V.A. Kelley, A.M. LaVenue, and J.F. Pickens, "Modeling of
Ground-Water Flow in the Culebra Dolomite at the Waste Isolation Pilot Plant
(WIPP) Site: Interim Report," Sandia National Laboratories, Contractor
Report SAND86-7167, 1987.
HAY76 Hays, J.D., J. Imbrie, and N.J. Shackleton, "Variations in the Earth's Orbit:
Pacemaker of the Ice Ages," Science 194:1121-1132, 1976.
HOL68 Hollister, J.C., and R.J. Weimer, eds., "Geophysical and geological studies of
the relationships between the Denver Earthquakes and the Rocky Mountain
Arsenal Well, Part A," Colorado School of Mines Quarterly, v. 63, no. 1, 270
p., 1968.
HOL84 Holt, R.M. and D.W. Powers, "Geotechnical Activities in the Waste Handling
Shaft, Waste Isolation Pilot Plant (WIPP) Project, Southeastern New Mexico,"
U.S. Department of Energy, WTSD-TME-038, 1984.
HOU90 Houghton, J.T., GJ. Jenkins, and J.J. Ephraums, "Climate Change: The
IPCC Scientific Assessment," New York, NY: Cambridge University Press,
1990,
HUN85 Hunter, R.L., "A Regional Water Balance for the Waste Isolation Pilot Plant
(WIPP) Site and Surrounding Area," SAND84-2233, Albuquerque, NM:
Sandia National Laboratories, 1985.
IMB80 Imbrie, J., and J.Z. Imbrie, "Modeling the Climatic Response to Orbital
Variations," Science 207:943-953, 1980.
IMB84 Imbrie, J., J.D. Hays, D.G. Martinson, A. Mclntyer, A.C. Mix, J.J. Merely,
N. G. Pisias, W.L. Prell, and J.J. Shackleton, "The Orbital Theory of
Pleistocene Climate: Support from a Revised Chronology of the Marine 618O
Record, pp. 269-305, Part 1," In: Milankovitch and Climate, Proceedings of
the NATO Advanced Research Workshop on Milankovitch, Palisades, NY,
November 30-December 4, 1982. A.L. Berber, J. Imbrie, J. Hays, G. Kukla,
and B. Saltzman (eds.), D. Reidel Publishing Co., Boston, MA, 1984.
IMB85 Imbrie, J., "A Theoretical Framework for the Pleistocene Ice Ages." Journal
of the Geological Society, 142:417-432, 1985.
LAM83 Lambert, S.J., "Dissolution of Evaporites in and Around the Delaware Basin,
Southeastern New Mexico and West Texas," Sandia National Laboratories,
Albuquerque, NM, SAND82-0461, 1983.
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MER83 Mercer, J.W., "Geohydrology of the Proposed Waste Isolation Pilot Plant
Site, Los Medanos Area, Southeastern New Mexico," U.S. Geological
Survey, Water-Resources Investigations 83-4016, 113 p., 1983.
MIL41 Milankovitch, M.M., "Canon of Isolation and the Ice-Age Problem,"
Koniglich Serbische Akademie, Beograd. (English translation by the Israel
Program for Scientific Translations), 1941, US Department of Commerce and
National Science Foundation. (Available from National Technical Information
Service, Springfield, VA, Order No. TT6751411012)
MIT89 Mitchell, J.F.B., "The 'Greenhouse' Effect and Climate Change," Review of
Geophysics, 27:115-139, 1989.
OLD89 Oldow, J.S., A.W Bally, H.G. Ave Lallemant, and W.P. Leeman,
"Phanerozoic evolution of the North American Cordillera;United States and
Canada," in Bally, A.W., and A.R. Palmer, eds., The Geology of North
America; An Overview: Boulder, Colorado, Geological Society of America,
The Geology of North America, v. A, p. 139-232, 1989.
POW78 Powers, D.W., S.J. Lambert, S.E. Schaffer, L.R. Hill, and W.D. Weart,
eds., "Geologic Characterization Report, Waste Isolation Pilot Plant (WIPP)
Site, Southeastern New Mexico (Volume I)," Sandia National Laboratories,
Albuquerque, NM, SAND78-1596, 1978.
SAN92 Sandia National Laboratories, "Preliminary Performance Assessment for the
Waste isolation Pilot Plant, December 1992," Sandia WIPP Project, SAND92-
0700, vols. 1-5, 1992.
SCH87 Schlesinger, M.E., and J.F.B. Mitchell, "Climate and Model Simulations of
the Equilibrium Climatic Response to Increased Carbon Dioxide," Reviews of
Geophysics, 25:760-798, 1987.
SIL94 Silva, M.K., "Implications of the Presence of Petroleum Resources on the
Integrity of the WIPP," EEG-551, June 1994.
SNY85 Snyder, R.P., "Dissolution of Halite and Gypsum, and Hydration of Anhydrite
to Gypsum, Rustler Formation, hi the Vicinity of the Waste Isolation Pilot
Plant, Southeastern New Mexico," U.S. Geological Survey Open File Report
85-229, Denver, CO, 1985.
SWI93 Swift, Peter N., "Long-term Climate Variability at the Waste Isolation Pilot
Plant, Southeastern New Mexico, US," SAND91-7055J, Environmental
Management, 17:83-97, 1993.
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SWI94 Swift, Peter N., B.L. Baker, Kathy Economy, J.W. Garner, J.C. Helton,
D.K. Rudeen, "Incorporating Long-Term Climate Change in Performance
Assessment for the Waste Isolation Pilot Plant," SAND93-2266, Sandia
National Laboratories, Albuquerque, NM, 1994.
WAS84 Washington, W.M., and G.A. Meehl, "Seasonal Cycle Experiment on the
Climate Sensitivity Due to Doubling of CO2 with an Atmospheric General
Circulation Model Coupled to a Simple Mixed-Layer Ocean Model," Journal
of Geophysical Research, 89:9475-9503, 1984.
WIL87 Wilson, C.A., and J.F.B. Mitchell, "A Doubled CO2 Climate Sensitivity
Experiment with a Global Climate Model Including a Simple Ocean," Journal
of Geophysical Research, 92:13, 315-13, 343, 1987.1, Part A," EPA/540/1-
89-002, Environmental Protection Agency, 1989.
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6. Use of Expert Judgment
6.1 INTRODUCTION
6.1.1 Background
In 40 CFR part 194, EPA states that expert judgment should only be permitted in situations
where data are not reasonably obtainable by collection or experimentation. EPA requires
that compliance applications clearly identify all instances in which judgment is used and the
experts involved. Documentation must be included which describes the process for expert
judgment elicitation, the results of expert elicitation, and the reasoning behind those results.
Documentation of interviews used to elicit judgments from experts, deliberations and formal
interactions among experts, background information provided to experts, and the questions or
issues presented for elicitation of expert judgment are also requested.
Although the Agency has not specified any particular methods for expert judgment
elicitation, the Agency has included some restrictions and guidelines for the selection of
individuals as experts in the 40 CFR part 194 criteria. These include prohibitions on:
selecting individuals who are members of the team of investigators requesting the judgment
or the team of investigators who will use the judgment; selecting individuals who maintain a
supervisory role or who are supervised by those who will utilize the judgment; and selecting
a membership of which less than two-thirds consists of individuals who are not employed
directly or indirectly by DOE (unless it can be shown that this is impracticable because of a
lack or unavailability of qualified independent experts, in which case at least one-third of the
membership must be non-DOE personnel). Compliance certification applications must
provide information which demonstrates that the expertise of any individuals and the panel,
as a whole, involved in expert judgment is consistent with the level of knowledge required by
the questions or issue presented to that individual and the panel.
Additionally, EPA requires that at least five individuals be used in any expert elicitation
process, unless a lack or unavailability of experts can be demonstrated and documented.
Also, any compliance certification application should include a discussion explaining the
relationship between the information presented, the questions asked, the judgment of any
expert panel or individual, and the purpose for which the expert judgment is being used.
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EPA requires that a minimum of five persons form an expert panel so that the elicited results
are representative of diverse viewpoint. This should result in a more informed and objective
process. However, an expert elicitation could be conducted with fewer than five individuals
in the event that there is a lack or unavailability of potential experts, provided that a rational
is stated. §194.26 of the final rule states this restriction:
At least five individuals shall be used in any expert elicitation process, unless there is
a lack or unavailability of experts and a documented rational is provided that explains
why fewer than five individuals were selected.
It is essential that any expert panel member should be free from conflict of interest.
Accordingly, two-thirds of the members of any panel should not be employed by DOE,
directly or indirectly. This restriction does not extend to those persons who receive funding
from the Department in those instances in which such funding is for activities not related to
WIPP (such as university professors). Expert panels may include persons employed by the
State of New Mexico Environmental Evaluation Group assuming that their expertise can be
demonstrated to be adequate for the elicitation. Compliance applications must demonstrate
this expertise, and EPA's judgment on the adequacy of this demonstration will be used in
making the decision on the issuance of certification.
Finally, EPA proposes that the elicitation process provide the public an opportunity for
presentation of scientific and technical views to the experts.
The Background Information Document (BID) for the 1993 amendments to 40 CFR part 191
notes diat "It is generally accepted that the use of expert judgment is required in the process
of evaluating the long-term containment potential of a geologic waste disposal facility"
(EPA93). In this context, the term "expert judgment" refers to a very structured, formalized
process involving panels of experts. However, expert judgment may also be applied by an
individual charged with making a determination on a given situation.
DOE is using both expert panel and individual investigator judgment to support the WIPP
performance assessment. In some instances, an expert panel may be convened and opinion
elicited using a highly structured, formal approach. In other cases, a single principal
investigator may be asked to supply an estimate of a parameter where a limited amount of
experimental data is available and also provide an estimated probability distribution function
for that parameter. The principal investigator may also be asked to define the probability
distribution function for a parameter where considerable experimental data are available, but
which still must be interpreted.
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6.1.2 NRC Publications on the Expert Judgment Process
NUREG/CR-5424 (NRC91) notes that the process by which expert judgment is elicited will
vary depending on the particular situation. NUREG/CR-5424 lists the following factors that
may affect how the judgment can best be gathered:
• The type of information needed from the experts (answers only or ancillary
expert data)
• The form in which the expert's answers are needed for input into a model
• The number of experts available
• The interaction desired among the experts
• The difficulty of setting up the problems
• The amount of time and study needed by the experts to provide judgments
• The time and resources available to the study
• The methodological preferences of the interviewer or knowledge engineer,
analyst, funder, and experts
Among the ways in which elicitation processes may differ include the degree to which the
experts interact, the structure imposed on the process, the number of meetings, whether the
expert's reasoning is requested or not, whether the expert judgment undergoes some
translation in a model and is returned to the experts for the next step, and whether all or
some of the elicitation is conducted in person, by mail, or by telephone.
NUREG/CR-5424 states that despite these variations, there are only three basic elicitation
situations and a general sequence of steps. The three basic situations are as follows:
Individual interviews - where one expert is interviewed in a private, usually face-to-
face situation, by an interviewer or knowledge engineer (a person who, hi addition to
interviewing, represents and enters the expert knowledge into a computer system).
This situation permits obtaining in-depth data from experts, such as on their means of
solving the problem, without distracting or influencing them with other experts.
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Interactive groups - where the experts are in a face-to-face situation with both one
another and a session moderator when they give their opinion. The degree of
structure may vary from totally unstructured to carefully choreographed as to when
the experts present their views and when there is open discussion.
Delphi where the experts give their judgments to a moderator, in isolation from one
another. The moderator makes the judgments anonymous, redistributes them to the
experts, and allows them to revise their previous judgments. If desired, the iterations
can be continued to the point where consensus is achieved. This process is intended
to counter some of the biasing effects of interaction.
The general sequence of steps in the elicitation process are described below (NRC91):
1. Selection of the question areas and particular questions
2. Refining of the questions
3. Selection and motivation of the experts
4. Selection of the components (building blocks) of elicitation
5. Designing and tailoring of the components of elicitation to fit the application
6. Practicing the elicitation and training the in-house personnel
7. Eliciting and documenting expert judgments (answers, and/or ancillary
information)
NUREG/CR-5411,"Elicitation and Use of Expert Judgement in Performance Assessment for
High-Level Radioactive Waste Repositories" (NRC90b), indicates five areas of performance
assessment of high-level waste repositories for which the benefits of a formal expert
judgment process may be warranted:
-scenario development and screening
-model development
-parameter estimation
-data collection and experimentation (information gathering)
-strategic repository decisions
40 CFR part 194 does not require that expert judgment be applied to any one area hi
particular, but leaves this choice up to DOE, subject to the restrictions in the final rule.
A well-documented application of the formal use of expert judgment is the U.S. Nuclear
Regulatory Commission's study, NUREG/CR-1150, "Severe Accident Risks: An Assessment
for Five U.S. Nuclear Power Plants" (NRC90). That study was undertaken to provide a risk
perspective for the radioactive release resulting from a core meltdown (see section 6.2.1.1
for further discussion of NUREG/CR-1150).
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6.2 EXAMPLES OF THE USE OF EXPERT JUDGMENT AT FACILITIES OTHER
THAN THE WIPP
Expert judgment has been used in various scientific forums not related to the WIPP. Four
reports on use of expert judgment are reviewed in this section: the NRC's application hi
severe accident risk assessment; the Electric Power Research Institute's use in a probabilistic
seismic hazard analysis; the United Kingdom's use in risk assessment of radioactive waste
disposal; and the European Space Agency's examination of expert judgment for risk
assessment in space programs.
6.2.1 Nuclear Regulatory Commission
6.2.1.1 NUREG/CR-1150
A well-documented use of expert judgment in the area of nuclear reactor safety is presented
in NUREG/CR-1150. The report summarizes an assessment of the risks from severe
accidents in five commercial nuclear power plants in the U.S. The risks were measured in a
number of ways, including: the estimated frequencies of core damage accidents from
internally initiated accidents and externally initiated accidents for two of the plants; the
performance of containment structures under severe accident loadings; the potential
magnitude of radionuclide releases and offsite consequences of such accidents; and the
overall risk (the product of accident frequencies and consequences) (NRC90).
The report notes that the risk analysis of severe reactor accidents inherently involves the
consideration of parameters for which little or no experiential data exist. Expert judgment
was used to supplement and interpret the available data. The principal steps used in the
formal elicitation of expert judgment for the NUREG/CR-1150 study are shown
schematically in Figure 6-1 and discussed briefly below:
• Selection of issues - The parameters considered were restricted to those with
the largest uncertainties, expected to be the most important to risk, and for
which widely accepted data were not available.
• Selection of experts - Seven panels of experts were assembled to consider the
various sets of principal issues. The experts were selected on the basis of their
recognized expertise in the issue areas. Representatives from the nuclear
industry, the NRC and its contractors, and academia were assigned to each
panel to ensure a balance of perspectives.
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Training in elicitation methods - Both the experts and analysis team members
received training from specialists in decision analysis. The team members
were trained in elicitation methods so that they would be proficient and
consistent in their elicitation. The experts' training included an introduction to
the elicitation and analysis methods, to the psychological aspects of probability
estimation (e.g., the tendency to be overly confident in the estimation of
probabilities), and to probability estimation.
Presentation and review of issues Presentations were made to each panel on
the set of issues to be considered, the definition of each issue, and relevant
data on the issues. Also, for the initial meeting, researchers, plant
representatives, and interested parties were invited to present their perspectives
on the issues to the experts. NUREG/CR-1150 notes that frequently these
presentations took several days.
Preparation of expert analyses - Following the initial meeting in which the
issues were presented, the experts were given time (from 1 to 4 months) to
prepare their analyses. During this period, several panels met to exchange
information and ideas. In some cases, panels were briefed by the project staff
on the results from other panels to provide the most current data.
Expert review and discussion After the experts had completed their analyses
a final meeting was held in which each expert discussed the methods he or she
used to analyze the issue. NUREG/CR-1150 states that while these
discussions frequently led to modifications of the preliminary judgments of
individual experts, the experts' actual judgments were not discussed in the
meeting because group dynamics can cause people to unconsciously alter their
judgments in the desire to conform.
Elicitation of experts - Following the panel discussions, each expert's
judgments were solicited. The elicitations were done privately with one expert
at a time so that the discussions could be performed hi depth and so that an
expert's judgments would not be adversely influenced by the others.
Composition and aggregation of judgments The analysis staff composed
probability distributions for each expert's judgments, and then aggregated the
individual judgments to provide a single composite judgment for each issue.
NUREG/CR-1150 notes that each expert's opinion was weighted equally in the
aggregation, based on findings in previous studies that this method performs
best.
Review by experts - Each expert's probability distribution and associated
documentation developed by the analysis staff were reviewed by that expert.
The purpose of this review was to ensure that potential misunderstandings
were identified and corrected and that the issue documentation properly
reflected the judgments of the expert.
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Selection
of Experts
Selection of
Issues
Elicitation
Training
Preparation
of Issues
Presentation
of Technical
Evidence
Presentation
of Issues
Expert Preparation
of Analyses
Discussion
of Analyses
Elicitation
of Experts
Composition
Documentation
Review
by Experts
From NRC90
Figure 6-1. Principal Steps in NUREG/CR-1150 Expert Elicitation Process
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Results of the elicitation are presented in NUREG/CR-1150, and documented in detail in two
separate reports (NRC89, NRC90a). While members of each panel are identified in
NUREG/CR-1150, specific judgments are presented anonymously (e.g., the members are
identified only as Expert A or Expert B).
6.2.1.2 Yucca Mountain Climate Study
More recently, the use of expert judgment elicitation was examined in a study to predict
future climate in the vicinity of Yucca Mountain, Nevada, the site currently undergoing
characterization for DOE's high-level waste repository. A report documenting that study was
presented at the Fifth Annual International Conference on High-Level Radioactive Waste
Management in Las Vegas, Nevada, in May 1994 (DeW94).
The expert elicitation procedure used consisted of the following 11 steps, adapted from
NUREG/CR-5411 (NRC90b):
1. Determine the objectives and goals of the study
2. Recruit the specialists (experts)
3. Identify the issues and information needs
4. Provide initial data to the specialists
5. Conduct the elicitation training session
6. Discuss and refine the issues
7. Provide a multi-week study period
8. Conduct the elicitation
9. Provide post-elicitation feedback
10. Aggregate the experts' judgments (if required)
11. Document the process
The elicitation team (those persons who would be responsible for conducting the elicitation)
and the expert panel were recruited concurrently with the development of the issue statement.
Nominations for expert panelists were formally requested from climatology/geography
associated societies and organizations. A formal peer-ranking based selection process
resulted in five final panel members, drawn from 42 nominees.
An initial meeting was held, with three goals: (1) to orient the experts; (2) to refine the
initial issue statement; and (3) to conduct elicitation training. The experts received
background information on the proposed repository system, the current and past climate in
the Yucca Mountain vicinity, and the NRC's performance assessment program. Extensive
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training was provided on probability elicitation, including the interpretation of subjective
probabilities, methods for generating subjective probabilities, and possible biases in the
judgment process. The experts refined the initial issue statement and generated a list of
factors and assumptions that would be considered by the group.
The experts had one month between the initial meeting and the individual elicitation to
review any relevant literature, run models, or otherwise prepare for the elicitation, including
providing a position paper. The experts had access to each other for consultation to
exchange data or clarify information, but they prepared their positions independently. The
actual elicitations were conducted individually to obtain the independent judgment of each
expert.
DeW94 concluded that while each of the steps used in the elicitation process influences the
outcome, four points are critical:
(1) the process of recruiting the experts should be formal and as unbiased as
possible;
(2) the credentials of the experts enhance the credibility of the elicitation, and
their ability to communicate their reasoning is a primary determinant of the
quality of the results;
(3) the conduct of the elicitation sessions themselves is extremely important and
should be well-planned and practiced ahead of time;
(4) concise and thorough documentation of the process including recording of the
elicitation sessions, as well as the results, differentiates between most informal
and formal expert judgment efforts and is essential in any formal expert
elicitation project.
6.2.2 Electric Power Research Institute
Expert judgment was used in a 1983 effort by the Electric Power Research Institute (EPRI)
and the 42 utilities in the Seismicity Owners Group (SOG) to develop a methodology for
assessing the seismic hazard at nuclear power plant sites. The results are documented in the
report Seismic Hazard Methodology for the Central and Eastern United States (EPR88).
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In the EPRI study, earth science expertise was provided by teams formed specifically to
promote interaction among different disciplines (geology, geophysics, seismology) and thus
avoid an overly narrow disciplinary focus. Work was conducted through a series of
workshops. Participating Earth Science Teams were required to identify and document
specific factors that could be used to evaluate the activity of tectonic features. For each
feature, each team was also required to assess the extent to which those factors were
exhibited. This established a distinction between scientific uncertainty (uncertainty in the
relationship between tectonic activity and physical characteristics) and information uncertainty
(the extent to which any particular feature'exhibits any given characteristics).
Six Earth Science Teams were formed to prepare and interpret input to the seismic hazard
analysis. EPRI used the team approach to achieve the interdisciplinary expertise needed to
evaluate various data sets and tectonic processes on a national scale. Team personnel were
chosen to strike a balance between, first, academic and applied experience, and second,
regional expertise. The stated overall aim was to minimize interpretation bias.
This study was accomplished through a series of seven workshops. The first workshop
defined data needs for the program. Workshops 2 through 7 were structured in pairs to
accomplish interpretations of tectonic stress regime, tectonic framework and seismic sources
and source seismicity parameters. Procedures were explored in depth during each workshop
to establish a common understanding among participants of the state of knowledge about
processes and the relative value of available data for making interpretations.
The Earth Science Teams proceeded with this information and their personal expertise to
develop their individual interpretations. Interpretations were shared among program
participants at the second workshop of each pan-. Each team shared the rationale and the
strength of theory and data supporting its interpretations. EPRI notes that although this
team-to-team interaction was desired, no effort was made to force a consensus interpretation
among teams on any element. Teams were asked to reach internal consensus on all
interpretations within a team.
EPRI stated that, with this approach, "..it is believed that uncertainty resulting from
incomplete understanding of tectonic processes has been captured. The estimated uncertainty
in hazard results ... reflects the state of the scientific community's uncertainty about
earthquakes causes and processes in the central and eastern United States." (EPR88)
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6.2.3 Other Countries
6.2.3.1 United Kingdom
A 1992 report commissioned in the United Kingdom by Her Majesty's Inspectorate of
Pollution, Department of the Environment (DoE) examined procedures for the elicitation of
expert judgments in probabilistic risk analysis of radioactive waste repositories (WAT92).
The report concluded that "Expert judgment is necessary for the measurement of uncertainty
about input parameters, since for many such parameters no frequency data are available."
WAT92 described several sources of bias that can influence judgments. The first noted was
that of availability. This is based on the observation that people are often influenced by the
ease with which they can remember the occurrence of similar events. An example was cited
in which a group of well educated people in the United States were told that about 50,000
people a year die in traffic accidents in the United States. They were then asked how many
people die from a long list of other causes, including common ailments such as heart disease
and rare ones such as smallpox vaccination. It was found that rare causes were
overestimated while common causes were underestimated. The explanation was that deaths
from a rare cause such as botulism are widely reported, but people commonly hear about
deaths from common causes such as stroke only when someone known to them dies in this
way. Because all cases of rare causes are available, but not all cases of common causes,
rare causes are overestimated.
A second possible source of bias is representativeness. This is based on the premise that, for
example, when people attempt to assess the probability that an individual belongs to a
particular class, on the basis of limited information, they judge the extent to which that
information suggests the individual is typical of the class, ignoring the underlying frequency.
In terms of a repository, an expert asked for the probability that the porosity of a rock
formation was greater than a certain figure might base his judgment on the extent to which
observable characteristics of the rock samples suggested that it was of a particular type of
known porosity, independent of the known distribution of different types of rock in the area.
This emphasizes the need for a published account of the reasons supporting a probability
judgment.
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The third potential source of bias presented was due to anchoring and adjustment. This
recognizes that a natural starting point for making a judgment may be chosen and the
judgment modified away from the initial position, but typically not far enough. Again using
the example of estimating the porosity of a rock, an expert might base judgment on a
different rock, for which the porosity was well known, but fail to adjust the estimate away
from this properly.
WAT92 also examined protocols for eliciting probabilities. It noted that the protocol
developed at the Stanford Research Institute, and referred to as the SRI protocol, has been
widely accepted. The SRI protocol has five stages.
In the first stage, the analyst motivates the person whose probabilities are to be
elicited. The first step in this process is to ensure that the expert understands the
nature and purpose of the analysis, and how the probabilities elicited will be used in
the analysis. In the second step, the analyst helps the expert to explore for possible
motivational biases, e.g., if the expert desires a low value of a variable, either
because he thinks that this would be consistent with what his superior is expecting
him to say, or because he wishes it were true, then such views should be discovered,
if possible, and the expert encouraged to account for them in the elicitation task.
The next stage in the protocol is the structuring phase. In this phase, the goal is to
make absolutely clear the definition of the variable for which the probability
distribution is being elicited. This should also include the exploration of assumptions
about the state of the world (e.g., in eliciting probability distributions for the porosity
of rock, assumptions made about physical variables such as temperature and pressure
which could affect porosity should be clearly defined).
The conditioning phase in the protocol is used to establish the data and arguments
which the expert is going to use to make judgments, and to cope with any identified
biases. Once the available data sets are listed, the expert is encouraged to consider
other possible ways of thinking about the variable, for example, focusing on scenarios
that might lead to extreme outcomes.
Numerical representation begins in the fourth, or encoding, phase. The techniques
for encoding can be categorized according to whether the probabilities are inferred
indirectly from the expert's judgments, or directly by asking the expert to respond
with a probability for a given event, or, in the case of continuous variables, for a
value such that the cumulative probability is equal to a given value. WAT92 notes
that there is no consensus that any one method is better than another, but that there is
a tendency for experts to be overconfident, i.e., they fail to spread their uncertainty
sufficiently.
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The final stage involves verifying that the numerical representations of uncertainty
properly support the expert's opinions. Assuming that no recognized protocols were
identified, the analyst must determine how best to conduct this process.
WAT92 does not reach a conclusion that any one protocol is superior, but it notes several
important considerations. First, the most important requirement for success is to devote
adequate time and effort to the process. Second, the experts should receive some general
introduction on the nature of elicitation, and what is known from psychology about
measuring perceptions of uncertainty. Third, the subject matter about which the judgments
are sought should be at the heart of the process. The analyst should have some knowledge of
the subject matter, and the experts should be encouraged to produce carefully reasoned
arguments to support his judgments. Fourth, throughout the process care must be taken to
ensure clarity and investigate unstated assumptions. Fifth, the encoding process should
follow a generally accepted technique.
6.2.3.2 European Space Agency
In February 1990 the European Space Agency (ESA) released a report entitled The Use of
Expert Judgment in Risk Assessment (COO90). The report provides the results of an
examination of expert judgment application. The examination included (1) a survey
identifying and studying the different and most important methods for the use of expert
judgment; (2) a survey of application of expert judgment data in industries, research institutes
and other organizations; and (3) development and evaluation of methodologies for expert
judgment application.
The major conclusion of the ESA research was that "The introduction of valid and effective
procedures for the use of expert judgment in risk assessment is a non-trivial, but worthwhile
task."
Nine phases in the procedures for using expert judgment were identified:
1. Problem identification phase
2. Expert identification phase
3. Expert choice phase
4. Question formulation phase
5. Seed variable choice
6. Elicitation phase
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7. Combination phase
8. Discrepancy analysis and feedback to experts
9. Documentation and communication phase
While the order and format may be somewhat different, these phases are approximately
equivalent in content to the steps previously described in section 6.1.3, with the exception of
step five, seed variable choice. The ESA study attempted to assess the quality of the
experts' judgments before their use, and establish a basis for calibrating the judgments. This
was done by eliciting the experts' judgment for quantities, known as seed or calibration
variables, that were known to the analyst but not to the experts.
One conclusion of the ESA study was that elicitation should be conducted individually, i.e.,
not in a group setting. The report notes "The advantage of group meetings is that experts
can discuss together with the analyst the interpretation of all the questions so encouraging a
shared understanding of their meaning. The disadvantage is that group processes naturally
suppress the spread of opinion and lead to underestimating uncertainty. The balance lies we
feel with the group not meeting." The report states that, if it is decided that the group
should meet, they meet to discuss the questions to be answered and separate before any
elicitation of the likelihoods takes place. This is consistent with the techniques used in the
NUREG/CR-1150 study, where groups met for presentation and discussion of the issues, but
the individual elicitations were conducted in private.
The ESA report also identified the need for the generation of an audit trail in the use of
expert judgment data to permit other analysts to repeat and check the information. The
report considered the issue of whether the experts should be protected with anonymity, and if
so, to what extent. Opinions were divided, but the report concluded the audit trail should be
such that "in circumstances of sufficient gravity" the ESA (as the client soliciting the expert
judgments) could identify all experts and their judgments.
6:3 REFERENCES
COO90 R. Cooke, S. French, and J. van Steen, "The Use of Expert Judgment hi Risk
Assessment," European Space Agency Reference CR(P) 3183, February 1990.
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DeW94 A. DeWispelare, M. P. Miklas, A. B. Gureghian, R. T. Clemson, L. T.
Herren, and J. R. Park, "The Use of Expert Judgement Elicitation to Predict
Future Climate for the Yucca Mountain Nevada Vicinity," Proceedings of the
Fifth Annual International Conference on High-Level Radioactive Waste
Management, Las Vegas, Nevada, pp. 1614-1621, May 22-26, 1994.
EPA93 U. S. Environmental Protection Agency, "Background Information Document
for Amendments to 40 CFR part 191, Environmental Standards for the
Management and Disposal of Spent Nuclear Fuel, High-Level and Transuranic
Radioactive Wastes," EPA 402-R-93-073, November 1993.
EPR88 Electric Power Research Institute, "Seismic Hazard Methodology for the
Central and Eastern United States," Volume 1, Part 2: Methodology (Revision
1), Final Report, November 1988.
NRC89 T.A. Wheeler et al., "Analysis of Core Damage Frequency from Internal
Events: Expert Judgment Elicitation," Sandia National Laboratories,
NUREG/CR-4550, Vol. 2, SAND86-2084, April 1989.
NRC90 NUREG/CR-1150, Severe Accident Risks: An Assessment for Five U.S.
Nuclear Power Plants," Final Summary Report, December 1990.
NRC90a F.T. Harper et al., "Evaluation of Severe Accident Risks: Quantification of
Major Input Parameters," Sandia National Laboratories, NUREG/CR-4551,
Vol.2, Revision 1, SAND86-1309, December 1990.
NRC90b NUREG/CR-5411, Elicitation and Use of Expert Judgement in Performance
Assessment for High-Level Radioactive Waste Repositories, May 1990.
NRC91 NUREG/CR-5424, Eliciting and Analyzing Expert Judgment, A Practical
Guide, March 1991.
WAT92 Watson, S.R., "Procedures of the Elicitation of Expert Judgements in the
Probabilistic Risk Analysis of Radioactive Waste Repositories: an Overview,"
Research commissioned by Her Majesty's Inspectorate of Pollution,
Department of the Environment, December 1992.
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7. Peer Review Procedures
7.1 INTRODUCTION
To operate the WIPP as a repository for transuranic radioactive waste, the DOE must
demonstrate that applicable health, safety, and environmental requirements have been
satisfied. Peer reviews may be employed as part of a comprehensive quality assurance
program. These peer reviews will give confidence that work completed, underway, or
planned was, is, or will be properly performed. The ASME "Quality Assurance Program
Requirements for the Collection of Scientific and Technical Information for Site
Characterization of High-Level Nuclear Waste Repositories," ASME-NQA-3-1989 Edition
(NQA-3), includes peer review among those activities affecting quality associated with the
collection of scientific and technical information, when other established methods cannot be
used to establish the adequacy of information.
Additional peer review is also necessary to establish the validity of procedures, methods, or
interpretations which may not be addressed by a quality assurance program. Because of the
nature of the assessments at the WIPP, in particular the potential uncertainties associated
with geotechnical data and their analyses, and the need to project performance over
thousands of years, peer reviews are essential to assure that all important factors are
considered in assessing the performance of the WIPP.
7.1.1 Background
The 40 CFR part 194 compliance criteria for the WIPP provide the following requirements
for peer review, at §194.27 of the final rule:
(a) Any compliance application shall include documentation of peer review that has
been conducted for, in a manner required by this section, for:
(1) Conceptual models selected and developed by the Department;
(2) Waste characterization analysis as required hi §194.24(b); and
(3) Engineered barrier evaluation as required in §194.44.
(b) Peer review processes required in paragraph (a) of this section, and conducted
subsequent to the promulgation of this part, shall be conducted in a manner that is
compatible with NUREG-1297 "Peer Review for High-Level Nuclear Waste
Repositories."
(c) Any compliance application shall:
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(1) Include information that demonstrates that peer review processes required
in paragraph (a), and conducted prior to the implementation of the
promulgation of this part, were conducted in accordance with an alternate
process substantially equivalent in effect to NUREG-1297 and approved by the
Administrator or the Administrator's authorized representative; and
(2) Document any peer review processes conducted in addition to those
required pursuant to paragraph (a) of this section. Such documentation shall
include formal requests, from the Department to outside review groups or
individuals, to review or comment on any information used to support
compliance applications, and the responses from such groups or individuals.
The EPA must be satisfied that peer review processes at the WIPP are sufficient to assess the
scientific premises properly on which the performance assessments are based.
7.2 OTHER PEER REVIEW PROGRAMS
7.2.1 Definition and Use of Peer Review
Peer review has a well-established role hi controlling various aspects of scientific research,
engineering research, scientific and engineering applications, and educational processes.
Editors and publishers of technical journals use peer review to ascertain the quality and
suitability of a manuscript submitted for publication. Funding agencies use peer review to
seek advice concerning the quality and promise of proposals for research support. Some
research institutions use peer review as another check on research in certain sensitive fields
such as human experimentation.l Some universities use peer review for promotions of
faculty (STE93).
Peer review serves a second objective of ensuring integrity in scientific research. Recent
spectacular cases of fraud in scientific research have led to federal regulations that require
' At the University of Michigan, potential researchers must complete a questionnaire that addresses certain
key areas of concern, prior to submitting their project funding request. If the applicant indicates that the
research will involve, for example, the use of human subjects, vertebrate animals, or radioactive materials, the
University subjects the project funding request to one or more peer review committees for approval. These
committees review the application for compliance with specific laws and regulations, why the research must be
conducted in the manner proposed, and how the research will be supervised (UMI90).
At the University of Michigan, proposals for research on human subjects can be reviewed by as many
as twelve peer review committees. The proposer must provide the rationale for and justify the use of each
subject. This peer review process is designed by the University to compel the would-be researcher to think
about their responsibilities, and to discuss these responsibilities with their colleagues (STE93).
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peer review as a measure to detect and prevent continuation of these frauds (NAT93). The
same pressures to enforce rules regarding misconduct in science have led to the inclusion of
mandatory peer review for ongoing scientific and engineering research and development by
government agencies.2 For example, one study reported that in 1987, five cases of fraud
and misconduct in science, primarily in biomedical research, were widely publicized. These
cases galvanized the U.S. Congress to renew its earlier interest in regulating certain aspects
of scientific research. In subsequent years, various federal agencies such as the Department
of Health and Human Services (DHHS), and within DHHS, the Public Health Service and
the National Institutes of Health (NIH), adopted rules to address these instances of fraud.
One rule was specifically related to improving the performance of peer review (GOL93).
Many Federal agencies use some form of peer review to evaluate the technical merit of
proposed research projects.3 Similarly, private institutions frequently use peer review to
evaluate research projects. Often the focus of these peer reviews is to determine which of
several proposed research projects will be funded. Those projects which, through the peer
review process, are deemed most promising are chosen for funding. Peer review is also used
to determine if a particular project merits continued funding. Some Federal agencies use
peer review to evaluate the technical adequacy of proposed or ongoing projects. This use of
peer review most closely parallels the peer review process required by 40 CFR part 194.
2 See e.g., 10 CFR Part 60
3 The National Science Foundation (NSF) uses a merit review system to make major awards in support of
important research facilities, centers, and other large-scale research-related activities. At the NSF, the entire
process of determining which research projects are to be publicly funded is called the merit review system. The
most important aspect of this merit review system is the technical peer review each potential research project
receives. The merit review system encompasses the administrative procedures for conducting this peer review
and the procedures for publicizing results (NSF94, 93, 92, 77; NSB67). Merit review also encompasses criteria
that the NSF considers necessary to augment technical quality and competence. These criteria include
immediate practical relevance, and the development of science and engineering capacity in all regions of the
country (NSF94). The purpose of the review process is to ensure that the most meritorious projects are chosen
for support, that the selection process is fair hi practice and perception, and that the results in each case are
clearly and publicly explained (NSF94; NAS93, 92). Officials at the NSF concluded that one of the reasons
that the United States has the most successful research system hi the world is because of the extensive use of
peer review to identify the best ideas for financial support. According to NSF94, "Peer review-based
procedures such as those in use at NSF, the National Institutes of Health, and other federal research agencies
remain the best procedures known for ensuring the technical excellence of research projects that deserve public
support. Motivating this process is clearly a true scientific interest in seeing that only technically feasible
projects get funded (NSF94). However, fiscal realities, and a growing occurrence of dishonesty in the research
process has focused even more attention on the adequacy of merit, or peer review at the National Science
Foundation (NSF94).
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7.2.2 Peer Review at the Department of Health and Human Services
7.2.2.1 Public Health Service
In the 1980s, the U.S. Congress began hearings on misconduct in scientific research because
of highly publicized cases of outright fraud. These problems seemed to have occurred to a
large degree in the biomedical research arena (GOL93). Congressional interest in this
problem led to several attempts to reform control of research at government agencies staffed
by the Public Health Service (PHS) (GOL93) and to require institutions receiving funding
from U.S. government agencies to put in place measures to detect and correct misconduct in
science (GOL93). In 1989, PHS promulgated a final rule which required any institution
applying for funds from the PHS to certify that it had adopted satisfactory misconduct
procedures. One element hi this set of procedures was a strong peer review procedure
(GOL93).
7.2.2.2 National Institutes of Health
Scientific research at the NIH is organized around Intramural and External projects, and the
peer review process differs somewhat according to project designation.
NIH Peer Review of Intramural Research. Scientists hi the Intramural Research Program
of the NIH are generally responsible for conducting original research consonant with the
goals of their individual Institutes, Centers, and Divisions. Senior NIH officials have
expressed concern regarding the rigor of scientific research conducted by NIH scientists
(NAT93). One of such research is peer review. In its 1990 "Guidelines for Conduct of
Scientific Research at the National Institutes of Health," NIH officials define peer review and
stress its importance as:
[an] expert critique of either a scientific treatise, such as an article prepared or
submitted for publication, a research grant proposal, a clinical research protocol, or
of an investigator's research program, as in a site visit. Peer review is an essential
component of the conduct of science.- Decisions on the funding of research proposals
and on the publication of experimental results must be based on thorough, fair, and
objective evaluations by recognized experts.4
4 See NAS93 quoting from "Guidelines for the Conduct of Research at the National Institutes of Health"
NIH, 1990.
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The guidelines go on to state the essential elements of the peer review process:
• the reviewer must be expert in the subject matter under review
• the reviewers should avoid any real or perceived conflict of interest that might
arise because of a direct competitive, collaborative, or other close relationship
with one or more of the authors of the material under review
• the review must be based solely on scientific evaluation of the material under
review within the context of published information and should not be
influenced by scientific information not publicly available
• the material being reviewed is privileged and should not be used to benefit the
reviewer unless the information has previously been made public
Peer review is used at the NIH at all stages of research carried on under the control,
direction, or funding of the organization. These stages are:
• idea/issue generation
• program formulation
• project proposal evaluation and selection
• final reports/other products review
• on-going program/project review
Peer review is performed to some extent in all these stages. There are, however, significant
variations both in the extent and nature of the peer review, depending on the stage of
research, and whether or not the activities are carried out by NIH employees, or if the
activities are NIH-funded.
At the NIH, peer review is mandated by law. Peer review policy and practice are generally
consistent throughout NIH with only limited variations among the institutes.
NIH Peer Review of External Research. Each institute of the NIH has a statutory peer
review panel, called a National Advisory Council (NAC, or variation) that conducts analyses
of projects. The NAC is usually composed of 16 to 18 senior level personnel, including a
mix of scientific disciplines as well as mandatory representation by ex officio federal officials
and public (i.e., nonscientific) members. The NACs have a broad charter to address any
matter affecting the performance of then- respective institute. All NACs prepare an annual
report assessing broad issues related to then: institute, including future directions and general
policy. These reports are then forwarded to the Director of NIH.
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NACs participate in peer reviews at many stages of a project. The first stage is generally
one involving the initial determination of project funding. Many different individuals and
activities compete for funding from the various institutes. NAC participation is generally
strong at this point in a potential project's life because the determination as to which projects
are funded bears heavily upon the direction the institute will take in the future. NIH
personnel need, and are statutorily required, to evaluate projects to ensure that only the most
promising ones are funded. The early stages of a project hi which the NACs are involved
include:
• Idea/Issue Generation Stage — This is likely to be an initial policy determination or a
determination to proceed along a certain avenue of research. Specific projects need
not be addressed.
• Program Formulation Stage — This stage receives peer review similar to that hi the
Idea Generation stage. One aspect of the peer review process at this stage is to
consider whether resource requirements related to long range goals are reflected in the
peer advice described under the Idea Generation stage.
• Project Design — For external research activities, the peer review is generally
limited. Following participation hi the formulation of research programs, the NACs
do not ordinarily play a part in the design of projects. According to one NIH official,
this is hi part because of time constraints on participation of council members. The
limited review is also due to inadequate scientific expertise (hi specialized areas of
research) of NAC members, which as mentioned, include senior officials and
members of the public, generally better qualified to provide broad ethical or legal
reviews.
• Project Proposal Evaluation, Selection, and Award Stage — At this point a dual
process of peer reviews takes place, looking at both technical merit and cost issues.
7.2.3 Peer Review at the National Aeronautics and Space Administration
The National Aeronautics and Space Administration (NASA) uses peer review primarily to
evaluate the merit of research and development proposals, and to allocate funding (NAS87,
94). NASA also uses peer review to decide if certain projects warrant renewal or continued
funding (NAS94). Organizationally, NASA includes three science offices and two
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engineering offices,5 each of which employs some form of peer review, particularly in
regards to evaluating proposals for research and development. However, NASA's science
offices use peer review to a much greater extent than do the engineering offices (NAS94).
The explanation for this difference hi use of peer review centers on the nature of work
performed by the two types of offices. The engineering offices are more likely to engage hi
straight procurement actions, such as purchasing rocket engines, whereas the science offices
have, as a goal, promoting specific types of research (NAS94). The discussion that follows
primarily describes the peer review process followed by the science offices. The NASA
engineering offices follow a similar, if somewhat truncated, version of the same process.
NASA uses two types of solicitations to obtain research proposals: an Announcement of
Opportunity (AO); and a NASA Research Announcement (NRA). NASA uses
Announcements of Opportunity for large research procurements such as designing an
instrument for installation on a satellite. NASA uses Research Announcements for more
narrowly focused scientific investigation, such as using data collected during the Magellan
spacecraft voyage (NAS94).
Either solicitation is written to address certain hardware, policy, or scientific needs as
envisioned by NASA. This narrow drafting of the solicitation eliminates the need to consider
policy during subsequent peer reviews of the proposals (NAS94). NASA has relied primarily
on panels convened under the auspices of the National Academy of Sciences (NAS) to advise
the administration on scientific goals and priorities. The panels report to various boards.
NAS creates these boards, excluding NASA scientists hi order to avoid any possible conflict
of interest. However, some panel members may be drawn from NASA offices (NAS94).
NASA takes the long-term science goals and priorities from the NAS and, through
committees, translates these goals and priorities into programmatic goals and strategies. The
committees are established by NASA under the auspices of an Advisory Council that is
composed of about 20 distinguished individuals, including corporate executives, university
5 The three science offices are:
Office of Life and Microgravity Sciences and Applications
Office of Mission to Planet Earth
Office of Space Science
The two engineering offices are:
Office of Aeronautics
Office of Advanced Concepts and Technology
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scientists, historians, and others. NASA selects the members of the Advisory Council.
Within the Advisory Council are a number of standing committees. These committees use
the NAS goals and priorities to develop recommendations for major programs such as the
Hubble space telescope (NAS94). These committees are, in effect, peer review panels.
When a NASA science office issues a research proposal (either an AO or a NRA), NASA
scientists compete for proposal acceptance on the same grounds as outside, or external,
scientists and external research entities such as universities. This intramural versus
extramural competition does not exist for Announcements of Opportunity issued by one of
the two NASA engineering offices, but may exist for NASA Research Announcements issued
by the engineering office.
Once the NASA standing committees have recommended programs, NASA officials in the
engineering and science offices translate them into specific program plans or projects,
including budget proposals. Each program or project is assigned to a program office within
the engineering or science office. At this point, the size of the program or project budget
partly determines whether an Announcement of Opportunity or a NASA Research
Announcement is made. The nature of the solicitation determines the following peer review
process (NAS94).
NASA uses Announcements of Opportunity to solicit proposals from scientists in the United
States and abroad. The AOs are typically used for larger budget items, about 100 million
dollars (NAS94). In response to an AO, NASA may receive up to 100 proposals. The
designated program scientist within a NASA science office establishes one or more peer
review panels to review each proposal, and selects the members for each panel.
Approximately 50 to 75 % of the peer review panel members are university scientists, with
the remaining members from NASA or other government agencies. The proposals are
mailed to the panel members for initial review. Some additional co-readers may also review
the proposals and can add comments, but do not participate further in the evaluation process.
After the initial review, the peer panel meets to discuss the proposals and reach consensus on
the evaluations. At the conclusion of the panel deliberations, the peer panel submits its
recommendations to the NASA program office. The recommendations are reviewed by
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NASA staff, and a NASA Associate Administrator makes the final determination as to which
proposals to fund (NAS94).6
NASA Research Announcements (NRAs) are usually for smaller-budget items and concern
more narrowly focused scientific research. The NASA Program Manager determines the
level of peer review to be conducted. Just as hi an AO, the proposals are usually mailed to
the members of the peer review panel. A follow-up panel meeting is often used to discuss
the proposals and to make recommendations to the NASA program office. An award under
an NRA will usually not call for a deliverable; rather, the expectation is that the results of
the research will be published. In this manner, NASA officials feel they are advancing the
boundaries of science (NAS94). Grants under the NRAs are typically for three years and the
awards are in the $100,000 range.
During the three-year period that an NRA is hi effect, NASA conducts periodic (e.g., annual
or mid-term) reviews of progress. This review is typically performed by the NASA Program
Manager, and could result in termination of the NRA award. If the scientific research must
extend beyond the original grant time period, the grantee submits a new proposal that is
reviewed using the same process as followed for initial selection.
Periodic review of projects funded under an AO follows a somewhat different course. The
initial peer panel that recommended particular grants will establish a Science Working Group
composed, essentially, of all the principal investigators of the winning organizations. This
Science Working Group is chaired by a NASA scientist. The Science Working Group meets
periodically to review progress and the final deliverables.
7.2.4 Peer Review at the U.S. Environmental Protection Agency
The EPA uses various panels such as Science Advisory Boards (SABs) (see Section 7.2.4.3)
and the National Advisory Council on Environmental Policy and Technology (NACEPT) to
advise the agency on scientific, technical, and policy matters. NACEPT activities are
authorized under Public Law 92-563, the Federal Advisory Committee Act (FACA), and are
6 Note that a single AO could result in numerous contracts, each concerned with a specific aspect of the
work described in the AO (NAS94). NASA awards under AOs typically call for production of specific
hardware such as scientific instruments to emplace on a research satellite. As such, the award is controlled
through a contract and the selection process is hi many respects similar to a regular government procurement
(NAS94).
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designed to provide recommendations and advice to the EPA Administrator. Review and
critique of documents and reports is a precursor to formulating sound advice.
The NACEPT is composed of several committees covering diverse technical areas. The EPA
established the WIPP Subcommittee under the aegis of the Environmental Measurements and
Chemical Accidents Committee to advise the Administrator in implementation of the WIPP
Land Withdrawal Act (PL 102-579) (EPA93). Members of the WIPP Subcommittee include
representatives from academia, an environmental activist organization, the New Mexico
Environmental Evaluation Group, the State of New Mexico Environment Department, and
various technical consulting companies. To date, the WIPP Subcommittee of the NACEPT
has provided advice to EPA in three general areas:
• criteria to be used in evaluating DOE Test Phase Plan and Waste Retrieval
Plan for WIPP
• criteria to be used in determining compliance with 40 CFR part 191
• selected issues related to 40 CFR part 194
The EPA uses SABs to provide, advice concerning on-going scientific studies within the
Agency. These SABs function in much the same manner as does the NACEPT. SAB
members include personnel from inside and outside the agency.
Finally, EPA makes use of external peer review groups. These groups can be constituted as
special panels formed by the agency, or can come from other government agencies.
Examples of these reviews are provided in sections 7.2.4.1 through 7.2.4.3.
7.2.4.1 Peer Review of Proposed Sewage Sludge Disposal Regulations
One particular EPA peer review effort concerned draft standards for the disposal of sewage
sludge, U.S. EPA Proposed Rule 40 CFR parts 257 and 503. The proposed rule was
reviewed by a peer review committee (PRC) created by an element of the U.S. Department
of Agriculture (USDA).
The EPA, under authority of the Clean Water Act,7 proposed regulations, to protect the
public health and the environment from any reasonably anticipated adverse effects of certain
7 33 U.S.C.A. 1251, et. seq.
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pollutants that might be present in sewage sludge. The proposed regulations were published
in 1989, and included standards for the final use or disposal of sewage sludge applied to both
agricultural and non-agricultural land, distributed or marketed, placed in disposal sites, or
incinerated. Part of the proposed regulation asked the USDA to review the scientific and
technical basis of the proposed rule. The review was conducted by a peer review committee
created by USDA's Cooperative States Research Service (CSRS), Regional Research
Technical Committee (W-170).8 Dr. A.L. Page of the University of California, Riverside
and Dr. T.J. Logan of Ohio State University were the co-chairs of the peer review
committee. The rest of the peer review committee consisted of 33 experts from academia,
government and private industry (USD89).9
The PRC met in Washington, D.C. for four days. The PRC broke into smaller workgroups
centered around specific aspects of the proposed regulation. For example, workgroups
analyzed those portions of the proposed regulations that dealt with monofills, with surface
disposal, with agricultural land application, etc. Each workgroup reviewed the proposed
regulations and prepared draft reports. During the four-day period, the entire 35-person
PRC would meet to discuss progress and to identify common areas. After the four-day
session, each workgroup reviewed and edited their section, and then the entire document was
reviewed and edited by each of the PRC members. The two PRC co-chairs, along with the
chairmen of each work group, met over a five-day period to revise and edit the complete
draft report (USD89).
The PRC draft report is organized as a series of workgroup reports, with an overall summary
and set of recommendations. The sections of the report prepared by the individual
workgroups list the workgroup members, but do not show which workgroup member
prepared any particular comment. The PRC draft report does not contain any information
regarding the background or qualifications of individual PRC members, nor does it include
8 The W-170 committee and its predecessors, W-124 and NC-118, are CSRS committees formulated for the
purpose of conducting regional research. These regional research projects are developed by researchers from
land grant universities, agricultural experiment stations, and USDA laboratories within four regions in the U.S.
(USDA89). As is obvious by the peer review of 40 CFR Parts 257 and 503, the W-170 Committee engages in
activities other than pure agricultural research.
9 Four members of the PRC were from the EPA, the agency whose work was being reviewed (USD89).
The PRC draft report does not explain how it avoided conflict of interest problems by having EPA personnel on
the PRC staff. However, the breadth and detailed nature of the comments prepared by the PRC tend to indicate
that the review was completely objective.
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any documentation regarding possible conflict of interest. In short, there is no way to know
from reading the PRC report if there was any conflict of interest.10
7.2.4.2 Ecological Risk Assessment Peer Review
In 1984, EPA organized the Risk Assessment Guidelines program to ensure scientific quality
and technical consistency hi the Agency's risk assessments. The first group of guidelines
was issued in 1986, and focused on evaluating risks to human health. In 1991, EPA issued
an agency-wide draft statement of general principles to guide ecological risk assessment.
This guide was titled "Framework for Ecological Risk Assessment."
To improve the technical basis for ecological risk assessment guidelines, EPA requested an
independent peer review of the draft "Framework for Ecological Risk Assessment." A panel
of twenty experts participated in the review (EPA92).11
The peer review of the draft framework consisted of three steps. First, the draft framework
document was mailed to each of the twenty members participating in the review. Each
reviewer prepared comments that were in turn distributed to all other reviewers. Next, a
peer review workshop was held to obtain an independent review of the logic, scientific
validity, and utility of the principles that were proposed in the draft framework document.
Workshop participants reached a consensus on the acceptability of some parts of the draft
framework, and made recommendations for changes to other parts. Finally, a written report
was prepared, summarizing the results of the workshop, and presenting the panel's
recommendations to EPA (EPA92).
To help frame the discussion and focus attention on certain critical issues, each workshop
participant was provided a set of "pre-meeting issue papers." These papers stated general
issues and then requested that the participants comment on specific aspects of the risk
10 One other point that should be noted is that there is no indication of follow-up on any of the PRC draft
report comments. While EPA may very well have incorporated all of the PRC comments into the proposed
regulation, that fact was not evaluated in this analysis.
11 Of the twenty participants, none were from the EPA, including its regional offices. Participants included
members of state and federal agencies, private industry and (primarily) public and private universities (EPA92).
Some of the entities represented had, however, done work for the EPA in the past. The EPA was represented
by about 30 "observers" (EPA92).
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assessment guidelines. For example, under the topic "Ecorisk Paradigm," each participant
was presented with the following statement and question to consider: "The proposed
paradigm for risk assessment is modeled after the National Research Council paradigm for
human health risk [reference omitted]. Is the modified paradigm presented in the framework
document appropriate for ecorisk assessments, or is another approach preferable?" (EPA92).
This general issue was followed by several sub-issues and questions that addressed specific
aspects of the proposed process for assessing risk.
The peer review described above did not indicate to what extent conflict of interest issues
were considered, although as previously noted, the report did state that the peer review was
"independent." The report also did not describe in detail the procedural steps followed
throughout the peer review process. For example, there is no discussion about how the peer
review members were selected, the process for incorporating comments, follow-up action if
the authors of the draft framework document disagreed with any comments of the peer
review team, etc.
7.2.4.3 The Science Advisory Board
The EPA uses a number of advisory councils, often known as Science Advisory Boards
(SABs), to provide guidance on a wide range of topics potentially affecting the environment
(EPA92a). These boards are part of EPA's advisory committee program, and operate under
the Federal Advisory Committee Act.12 The EPA constitutes and terminates SABs as the
need arises. For example, from 1992 to 1993, EPA formed eight new committees for topics
as diverse as wood furniture manufacturing and local government policy, and terminated five
committees (EPA93a).
The composition of some SABs indicate that their function extends beyond the technical
arena. For example, the Clean Air Act Advisory committee comprises "50 senior
representatives from state and local government, academic institutions, unions, environmental
and public interest groups, industries and service groups." (EPA92a). The four workgroups
formed from this committee addressed topics such as: effective communication/outreach
methods for implementing reductions in airborne emissions; regulatory reform options; and
12 Public Law 92-463, October 6, 1972. Further guidance on the functioning of federal advisory
committees can be found in General Services Administration Final Rule Subpart 101-6.10, "Federal Advisory
Committee Management," August 1989.
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alternative programs that would assist and encourage states to promote energy efficiency
(EPA92a), On the other hand, the Federal Insecticide, Fungicide, and Rodenticide Act
Scientific Advisory Panel is comprised entirely of experts in the field of pesticides and the
impact of their use on human health and the environment (EPA92a).
Each of the committees or SABs formed under EPA's advisory committee program develops
or is issued a charter stating the purpose, objective and scope of activity, functions, and
conduct of meetings (EPA92a). Committee members may or may not be compensated for
their services depending on the individual committee. The EPA may pay travel and per diem
expenses for all committee members. Some committee members may be government
employees. All members are subject to conflict-of-interest restrictions (EPA92a).13
The various committees meet periodically throughout the year, at times established in their
individual charters. The chair of each panel or committee submits a written report of the
meeting. This report includes the panel's recommendations and conclusions. Transcripts are
made and retained for the entire meeting (EPA92a).
7.2.5 Nuclear Regulatory Commission Peer Review Guidance
Compliance criteria in 40 CFR part 194 require peer review at the WIPP to be performed in
a manner compatible with NUREG-1297. NUREG-1297 contains a generic technical
position for peer review at high-level nuclear waste repositories (NRC88). NUREG-1297
provides guidance on the definition of peer reviews, the areas where a peer review is
appropriate, the acceptability of peers, and the conduct and documentation of a peer review.
The NUREG document defines the following peer review-related terms:
• Peer — a person having technical expertise in the subject matter, or a critical subset
of the subject matter, at least equivalent to that needed for the original work.
• Peer review group — an assembly of peers representing an appropriate spectrum of
knowledge and experience in the subject matter. The group should vary in size based
on the subject matter and the importance of that subject matter to safety or waste
isolation.
13 See 40 CFR Part 3, Subpart F standards of Conduct for Special Government Employees. This
regulation includes rules regarding conflict-of-interest. The rules require nominees to committees such as
EPA's, to submit a Confidential Statement of Employment and Financial Interests (EPA Form 3120-1), that
fully discloses any outside sources of financial support.
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• Peer review — a documented, critical review performed by peers who are
independent of the work being reviewed. The review is an in-depth critique of the
assumptions, calculations, extrapolations, alternate interpretations, methodology, and
acceptance criteria employed, and of conclusions drawn from the original work.
• Peer independence — a peer was not involved as a participant, supervisor, technical
reviewer, or advisor in the work being reviewed and, to the extent feasible, has
sufficient freedom from funding considerations to assure the work is impartially
reviewed.14
The NUREG document describes the circumstances under which a peer review is warranted.
These circumstances occur when the suitability of procedures and methods essential to
showing that the repository system meets or exceeds its performance requirements cannot
otherwise be established through: 1) testing; 2) alternative calculations; or 3) reference to
previously established standards and practices. (These circumstances are the same as those
listed hi ASME NQA-3, previously described in section 7.1.) NUREG-1297 provides
examples of these situations, including when:
• critical interpretations or decisions will be made in the face of significant
uncertainty,
• decisions or interpretations having significant impact on PA conclusions will
be made, or
• novel or beyond state-of-the-art testing, plans and procedures, or analyses are
or will be utilized.
The composition of the peer review group depends on: the complexity of the work to be
reviewed; its importance hi establishing compliance with safety or performance goals; the
degree of uncertainty in data or the technical approach; and the extent to which differing
viewpoints exist. The peer review group should include individuals representing major
schools of scientific thought. The actual number of peer reviewers is not as important as the
technical qualifications of the reviewers. The group should be structured to avoid a bias
toward particular theories, methods of analysis, or institutional practices (NRC88).
14 NUREG-1297 states that, because of the DOE's pervasive effort in the waste management area, most
persons who would be acceptable from a technical perspective are likely to have had some connection to DOE
in the past. As such, the NUREG document concludes that '[I\t may not be possible to exclude all DOE or
DOE contractor personnel from participating in a peer review." (NRC88). The NUREG document suggests that
in these cases, a documented rationale as to why someone of equivalent technical qualifications and greater
independence was not selected should be filed with the peer review report (NRC88).
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Each peer review group should be led by a chairman. In meetings and correspondence, the
peer review group should evaluate and report on:
• validity of assumptions
• alternate interpretations
• uncertainty of results and consequences if wrong
• appropriateness and limitations of methodology and procedures
• adequacy of application
• accuracy of calculations
• validity of conclusions
• adequacy of requirements and criteria
NUREG-1297 states that full and frank discussions are essential between the peer reviewers
and the persons who performed the work being reviewed (NRC88).
The peer review process should include written minutes of any proceedings, deliberations,
and activities of the peer review group. After the peer review group completes its analysis,
the agency responsible for quality assurance should produce a written report, under the
direction of the peer review group chairman, and signed by each member of the group. The
report should include statements by individual members stating any dissenting views or
additional comments as appropriate. The report should also include information concerning
the qualifications of individual peer review group members and their organizational
affiliations (NRC88).
7.2.6 Peer Review at the Department of Energy
The Office of Program Analysis (OPA) conducts peer review assessments of DOE research
and development. "Procedures for Peer Review Assessments," DOE/ER-0491P, dated April
1991, describes general processes for conducting these peer reviews (DOE91). _The peer
review procedures are intended to provide the basis for implementing the methodology
developed by OPA. The reviews are performed by examining individual projects which
comprise a program and by assessing the quality of the research, quality of the research
team, productivity, probability of success, and mission relevance for each project reviewed.
This OPA peer review is intended as a funding screening method, not a thorough scientific
analysis of the project, or project report. In fact, this peer review procedure is also
limited—the review is limited to 65 minutes, and of that period, 30 minutes are allocated to
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the "Principal Investigator" to present the project's hypothesis, scientific approach, and
results. After the peer review panel has completed all evaluations of all projects assigned to
it for review, the panel members make their recommendations considering two basic criteria:
the highest payoff research needs or opportunities, and their order of priority.
7.2.6.1 Methodology
The DOE assembles peer review panels, as required, in its primary functional areas of
research. Project reviews take place in panel sessions lasting from two to four days. Prior
to the panel session, each principal investigator submits a package of documents which is
distributed to the panel members to assist in evaluating the principal investigator's project at
the panel session.
7.2.6.2 Peer Review at the Yucca Mountain Site
Yucca Mountain is being considered as a site for long-term deposition of high-level
radioactive waste. As such, many of the detailed geologic, hydrologic, and other scientific
investigations being performed at the WIPP are or were also performed at Yucca Mountain.
This section examines peer review of one of these proposed studies and one of the
investigation reports.
Peer review panel members were provided advance copies of the draft documents to be
considered. In one case, this was a copy of two proposals for in-situ study of radionuclide
migration (NVO81). In the other case, the document was a report of an investigation of
hydrology and geology in the Yucca Mountain area (NVOSla). Panel members were given a
"charge," that is, they were asked by the Technical Project Officer to review the documents
with certain criteria hi mind. Following the panel session, each individual peer reviewer sent
comments to the Technical Project Office. The technical project officer then prepared
responses to these comments, and submitted the responses to NTS management.
Peer Review of Radionuclide Migration in Tuff and Granite, NVO 196-23. This review
concerned two proposals for future in situ investigations of radionuclide migration in
tuffaceous and granitic rock. The proposed radionuclide migration work was to be conducted
by scientists from Los Alamos National Laboratory, Sandia National Laboratories, and
Argonne National Laboratory for the tuffaceous rock and the Lawrence Livermore National
Laboratory for the granitic rock (NVO81).
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Comments in the radionuclide migration peer review report were compiled from a peer
review panel meeting conducted August 18-19, 1980 in Las Vegas, Nevada, as well as
individual comments submitted by each of the peer review panel members. Individual
comments were submitted after the panel meeting concluded, typically within one month
(NVO81).
The need for a peer review panel was determined the Office of Nuclear Waste Isolation
(ONWI) and the Nevada Nuclear Waste Storage Investigations (NNWSI).15 Reviewers
representing appropriate fields of expertise were invited to attend the review sessions.
Nationally known, as well as prominent state and local, scientists were selected to participate
in the peer review process. At the peer review meetings, the NNWSI Technical Project
Officers, Principal Investigators from the laboratories, and NTS technical staff members
involved in the radionuclide migration studies made detailed presentations and answered
questions about their investigative actions and findings, as well as the proposed study efforts.
The peer review panel consisted of eight scientists from universities, private industry, and a
government agency other than DOE.16
The peer review report contains summarized comments from the panel sessions as well as
individual comments from reviewers. Some summarized comments lack scientific precision.
For example, in discussing the use of tracers, the peer review panel noted that "Tracers
mentioned for cold experiment (except for U-235) will sorb like crazy and never be observed
at the collection point...." (NVO81).
A section of the peer review report is a reply to the peer review comments prepared by the
Technical Project Officer. The reply indicated that some of the panel's comments will be
incorporated; however, the panel provided no mandatory comments, and there was no
opportunity for the panel to concur with the reply prepared by the Technical Project Officer.
15 The NNWSI were a part of the National Waste Terminal (NWTS) Program of the DOE. The NNWSI
were formally organized in 1977 and managed by the Waste Management Project Office of DOE's Nevada
Operations Office. The NNWSI existed to develop or improve the technology for high-level nuclear waste
handling, containment, and isolation, and determining whether suitable rock units on or adjacent to the Nevada
Test Site (NTS) were technically acceptable for a licensed permanent nuclear waste repository (NVO81,
NVO81a).
16 The one government panel member was from the U.S. Geologic Survey.
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Peer Review of Geologic and Hydrologic Investigation of Yucca Mountain Peer Review
Documentation, NVO-196-22. This peer review was conducted in a somewhat different
manner than the previously described review. Peer review panel members were provided the
basic report ahead of time, and then met to discuss its technical merit. The panel session
included a site visit to a portion of the Nevada Test Site. Following the panel session,
individual members submitted written comments to the Technical Project Officer.
One reviewer felt that the panel meeting was not long enough to complete all required
discussions. The same reviewer was dissatisfied with breaking the panel review into several
different workshops. He felt that there was too much interrelationship between, for example,
the hydrology and geology sections for them to be discussed separately.
Several panel members commented that, although the peer review panel presentations were
generally useful, the panel members did not receive a handout of the material from the
individual speakers prior to their discussions. One of these reviewers went on to note that
"Most of us were not that familiar with the geologic formations, their positions in the
geologic column, or the details that characterize them." (NVOSla).
The lone representative (out of ten) from industry noted that a certain difference existed
between the academicians on the panel and himself.17 Because of what he perceived as the
"urgency of the problem" (i.e., finding a repository for high level radioactive waste), this
reviewer felt that "more forward, goal oriented (industrial) approach to the depository siting
should be considered." He urged inclusion of more industry representatives on future peer
review panels. (NVOSla).
Summary of NVO Peer Reviews. These two peer review documents might seem, on
cursory inspection, to be similar. However, NVO 196-23 appears to be related to a
research/investigation funding decision while NVO 196-22 is more of a technical review that
is intended to verify the adequacy of the investigation. However, the peer review comments
in NVO 196-23 are primarily technical in nature, and contain no specific recommendations as
to whether or not the two projects should be funded. The comments for NVO 196-22, on
the other hand, address several programmatic issues hi addition to technical comments.
17 According to this reviewer, "The university - industry differences in approach to applied research
investigations is well known." (NVOSla).
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The format for submitting comments, a general set of comments from the peer review panel
followed by written comments from individual panel members, appears to have posed a
dilemma for the NNWIS staff and the Technical Project Officer. Some individual comments
were diametrically opposed, while other comments reflected the views of only one or a few
members. One reason for this situation may have been the apparent lack of a panel
chairman. Instead of the panel evaluating and concurring in the work, hi effect several
different (eight to ten) peer reviewers commented. The project staff was thus put in the
position of having to respond to more comments than if a unified set of comments had been
forwarded by a panel chairman, and the project staff had to respond to comments that did not
always agree as to the direction the research should follow.
In addition, post-panel submission of written comments meant that any potential interactions
between panel members and the project staff was necessarily limited or non-existent. This
problem was apparent from several project staff replies. For example, several panel
members referred to the need to access "open-file reports" and to have project documentation
provided to the panel members in advance of the peer review panel meeting. The project
team responded that they were uncertain as to what reviewers mean by higher profile reports.
Misunderstandings like this might be resolved if the comments had been provided during the
panel session, or if some mechanism existed for the panel members and the project staff to
interact after submission of peer review comments.
7.3 SUMMARY OF PEER REVIEW
Peer review, as practiced at other government agencies and at private institutions such as
universities, varies from an informal process hi which reviewers are mailed a document and
after review, simply send back a set of comments, to a more formal process with specific
agendas, scheduled panel meetings, specific forms to use for recording comments, feedback
mechanisms between panel members and project staff, etc. Based on the different processes
used by various agencies for conducting peer reviews, the most effective peer reviews occur
when:
• sufficient advance notice is given;
• adequate numbers of reviewers are selected so that all aspects of the project
are represented;
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• for large projects, a panel session lasts several days and includes site visits, if
necessary;
• the panel has a chairman, and the chairman seeks to gain consensus on the
peer review comments and presents a unified list of conclusions to the
Principal Investigator;
• any strong disagreements among panel members are be highlighted;
• reviewers have a charter, or a check list of items or evaluation criteria, to
consider in the review;
• members are not discouraged from voicing opinions in any area related to the
subject being reviewed;
• a process is in place wherein responses to peer review comments are reviewed
by at least the panel chairman;
• any major differences between panel and project staff are resolved;
• the agency forming the peer review panel strives for a balance of expertise,
and of scientific views, on the panel.
Thus, use of peer review to establish the accuracy or adequacy of scientific procedures,
methods, scope of examination, or data is best accomplished when peer reviewers are
selected based on depth and area of expertise (and considering possible conflict of interest),
and when the process and results are thoroughly documented and responded to by principle
investigators.
After evaluating a variety of peer review programs for different purposes, EPA has identified
important criteria for conducting and documenting peer review, as described above. The
Agency determined that these criteria are clearly articulated in NUREG-1297, which provides
appropriate guidance for implementing such procedures. Thus, the final rule provides that
peer reviews required for the WIPP must be conducted in a manner that is compatible with
NUREG-1297.
7.4 REFERENCES
DOE91 "Procedures for Peer Review Assessments," U.S. DOE, Office of Energy
Research, DOE/ER-Q491P, April 1991.
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EPA92 "Peer Review Workshop Report on a Framework for Ecological Risk
Assessment," EPA/625/3-91/022, February 1992.
EPA92a "U.S. Environmental Protection Agency Advisory Committee, Charters,
Rosters, and Accomplishments, Addendum," EPA 202-B-92-005, June 1992.
EPA93 "Implementation Strategy for the Waste Isolation Pilot Plant Land Withdrawal
Act of 1992," EPA 402-R-93-002, Office of Radiation and Indoor Air, March
1993.
EPA93a "U.S. Environmental Protection Agency Advisory Committee, Charters,
Rosters, and Accomplishments, Addendum," June 1993.
GOL93 "Congressional Activities Regarding Misconduct and Integrity in Science,"
Barry D. Gold, published in Responsible Science, National Academy Press,
Washington, D.C., 1993.
NAS92 "Responsible Science, Volume I" report by the National Academy of Sciences,
National Academy Press, Washington, D.C. 1992.
NAS87 "Guidelines for Acquisition of Investigations, NASA Handbook 8030.6B,"
September 1987
NAS94 Personal Communications with Carrie Sorrels, May-June 1994.
NAT93 "Responsible Science, Volume II" report by the National Academy of
Sciences, National Academy Press, Washington, D.C. 1993.
NRC88 "Peer Review for High-Level Nuclear Waste Repositories," NUREG CR/1297,
1988.
NSB67 "Criteria for the Support of Research by the National Science Foundation,"
NSB-67-133, May 1967.
NSF77 "Processing Recommendations Requiring National Science Board Approval,"
NSF Circular No. 107, Revision 2, April 21, 1977.
NSF92 "Grants for Research and Education in Science and Engineering: An
Application Guide," NSF-92-89.
NSF93 "Proposal and Award Manual," NSF Manual No. 10, February 1993.
NSF94 "Major Award Decisionmaking at the National Science Foundation," report by
the National Academy of Sciences, National Academy Press, Washington,
D.C. 1994.
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NVO81 "Geologic and Hydrogeologic Investigation of Yucca Mountain Peer Review
Documentation," NVO-196-22, April 1981.
NVOSla "Radionuclide Migration in Tuff and Granite Peer Review Documentation,"
NVO-196-23, April 1981.
STE93 "Fostering Responsible Conduct in Science and Engineering Research:
Current University Policies and Actions," Nicholas H. Steneck, published in
Responsible Science. National Academy Press, Washington D.C., 1993.
UMI90 "Administration of Sponsored Projects," University of Michigan, 1990.
USD89 "Peer Review, Standards for the Disposal of Sewage Sludge, U.S. EPA
Proposed Rule 40 CFR Parts-257 and 503," T.J. Logan and A.L. Page, Co-
Chairman, Report for U.S.D.A., Cooperative State Research Service
Technical Committee W-170, July 24, 1989.on v. EPA, 627 F.2d 416 (D.C.
Cir. 1980).
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8. Uncertainty and "Reasonable Expectation"
8.1. INTRODUCTION
8.1.1 Background
The final rule places statistical requirements on the results of performance assessments in
order to consider the quantitative uncertainty inherent in long-term predictions. As stated in
EPA's 40 CFR part 191 disposal standards, "Because of the long tune period involved and
the nature of the processes and events of interest, there will inevitably be substantial
uncertainties hi projecting disposal system performance. Proof of the future performance of
a disposal system is not to be had in the ordinary sense of the word in situations that deal
with much shorter time frames."
Several physical processes take place over the design life of the WIPP. Because of
uncertainty propagation, a calculation based on worst case values for each input parameter
across all processes would yield results that would not be likely to portray disposal system
performance. This is only one dimension of the uncertainty issue. A paper presented by
Zuidema of Switzerland's National Cooperative for the Storage of Radioactive Waste
(NAGRA) at a 1991 NEA workshop on criteria for HLW disposal identified four sources of
uncertainty hi disposal system safety analysis:
• uncertainty hi scenarios;
• uncertainty hi conceptual models;
• parameter uncertainty; and
• parameter variability.
Recognizing these uncertainties, the disposal standards state that there should be "a
reasonable expectation ... that compliance ... will be achieved." This phrase represents a
general principal for due consideration of (1) the uncertainties involved hi projecting disposal
system performance for 10,000 years, and (2) the entire record submitted to the
Administrator.
The goal of the WIPP performance assessment is to develop predictions of the distributions
of the cumulative release, doses to individuals, and radionuclide concentrations in ground
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water over 10,000 years at the WIPP disposal site (SNL92, HEL93a). These distributions
are functions that indicate the probability of exceeding various levels of three parameters:
cumulative releases, doses to individuals and ground-water concentrations. Ideally, a single
function for each parameter would be formed by combining distributions resulting from all
possible scenarios, after considering all uncertainties in scenarios, in conceptual models, in
parameter uncertainty, and in parameter variability over the 10,000-year regulatory horizon
mandated in 40 CFR part 191.
Certain physical parameters may remain unchanged over 10,000 years, but changes, for
example, in geology and climate, must be forecast "to the extent practicable" (§194.25(b)).
Among the scenarios of interest is the number of times humans will inadvertently intrude into
the disposal system in search of resources. DOE has developed a human intrusion scenario
(SNL93, HEL93b) with sub-scenarios including two different intrusions: the first is a
penetration of the disposal system and a brine pocket below the disposal system which allows
brine to enter the disposal system; the second is the interception of the disposal system
without hitting a brine pocket. Both events result in the release of the radionuclides to the
accessible environment from the cuttings associated with drilling operations and can result hi
release to the accessible environment by lateral transport in ground water associated with
overlying geologic formations.
The predictions generated by the WIPP performance assessment model for a variety of
human intrusion scenarios are made conditionally: if scenario A occurs and the model
parameters are assigned certain values, then the model predicts the distribution of releases
under a specific set of assumptions. The predictions are made using very elaborate computer
codes requiring many input parameters to define the scenarios and their implications. These
input parameters may be based on actual data or on expert judgments.
8.1.2 General Approach to Evaluating Compliance
This section discusses the statistical concepts which are relevant to regulatory decision-
making. The random variable R denotes the cumulative release of radionuclides from a
disposal system, while a fixed numerical limit (L) is selected as the maximum allowable
release for the disposal system. Unless mere is a strict upper bound on its distribution, the
random release cannot be proved to satisfy a specific mathematical constraint of the form "R
is less than L." At best, if the probability distribution of the release is known exactly, then
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the probability that the release is less than the regulatory limit may be calculated from the
distribution. This probability is denoted by the notation Pr{ R < L }, which is read as the
"probability that R is less than L."
For most regulatory applications, it is sufficient to require that the probability of the release
»
being less than the regulatory limit is high. Probability values near 100 percent would be
necessary to ensure that compliance is almost always obtained. If one uses the symbol P
denote this high level of probability, the regulatory requirement would be written as
Pr{ R < L } > P
An equivalent statement of the regulatory requirement is that the P* percentile of the
distribution of R be less than the regulatory limit L. Let RP denote the P* percentile of the
distribution of R. Using percentiles, the regulatory requirement may be written as
RP < L
Under either of these equivalent interpretations of the requirements in §191.13(a), there
would be at least a probability P that the random release is less than the regulatory limit. In
this case of a known distribution for the release, all that remains is to select the appropriate
value of the required probability (P), and the appropriate value for the release limit (L).
If this were an enforcement problem for a hypothetical nuclear plant, the distribution of the
release could be determined by going to the site and measuring radionuclide releases from
the stack. From these observations, collected over time, the distribution of the random
release could be estimated, leading to a compliance determination based on data and standard
statistical procedures.
In this example, the need to estimate the distribution of the release from sample values
results in a sampling error for the estimated probability that the release is less than the
regulatory limit. If the percentile interpretation is used, there will be sampling error in
estimating the P"1 percentile of the distribution. These sampling errors should be considered
when comparing the estimated probability or percentile to the requirement. A 90 or 95
percent confidence interval is often used for sample-based estimates of the probability or
percentile. If the confidence interval lies entirely below the required value, the power plant
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is determined to be in compliance at the appropriate level of confidence. If the resulting
confidence interval was too broad to reach a clear determination of compliance, more
observations could be collected to reduce the confidence interval.
Confidence interval procedures are designed to be applied to samples of observations on the
random variable of interest. As more and more observations are collected, more and more
information is gained about the distribution of the random variable, and the resulting
confidence intervals on the estimated probabilities and percentiles become smaller and
smaller.
There is a fundamental difference between this well-known procedure of collecting actual
observations on a random variable and the process of making predictions of a random
variable. As more and more predictions of a random variable are generated, there is no
guarantee that more information is generated about the true distribution for a future
realization of the random variable.
Unlike the simple nuclear plant example above, the subject of analysis for the WIPP disposal
system is the distribution of cumulative releases of radionuclides to the accessible
environment over a 10,000-year tune frame. Because the cumulative release for this site is a
future realization of a random variable, its prediction involves considerable uncertainty.
Estimates of the probabilities and/or percentiles are required to verify compliance. Due to
the uncertainty intervals surrounding these estimates, there can be no absolute assurance that
the probability statements contained hi the regulations are satisfied. At best, compliance can
be determined only to within a certain level of confidence.
Estimates derived from the WIPP performance assessment modeling system will have errors
of prediction associated with each estimate produced by the model. However, unlike the
sampling problem referred to earlier, the "confidence intervals" for the estimates are not
necessarily reduced by running the model repeatedly, generating more and more predictions
based on the same assumptions. Rather, the WIPP performance assessment has attempted to
reduce the "confidence intervals" or, more generally, the uncertainty interval surrounding the
estimated probabilities and percentiles by running the model under a wide variety of
assumptions.
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The current WIPP performance assessment process addresses two sets of uncertainties:
1) the uncertainties surrounding probability distributions selected for values of
parameters used as WIPP performance assessment model inputs; and
2) the uncertainties surrounding the definition, screening, and quantification of
possible future scenarios and their probabilities.
DOE has expanded its efforts to quantify uncertainty distributions for the input parameters,
thus reducing uncertainty due to the first type of assumptions. Panels of experts have
addressed the probabilities of future scenarios in an attempt to reduce uncertainty regarding
the second type of assumptions.
In the nuclear plant example, there was a possibility of reducing the size of the confidence
intervals of estimates by collecting more observations at the site. In the prediction problem,
only one method is available for reducing the uncertainty of the resulting estimates. This
method involves quantifying the uncertainty associated with the assumptions on which the
forecast is based. The uncertainty surrounding model input parameters may be estimated by
specifying probability distributions for these parameters based on the best knowledge of the
disposal system. If the uncertainty distributions are developed based on observations of the
variability of physical parameters measured at the WIPP site, then uncertainty intervals
surrounding these physical parameters can be reduced further by collecting more information
about the site, waste characteristics, and their interactions. This leads to a greater reduction
in uncertainty concerning the estimated probabilities and percentiles for the total release.
Thus, uncertainty due to the first set of assumptions may be reduced by collecting more
information about the model input parameters.
The selection of possible future scenarios and the probabilities assigned to these scenarios hi
the second set of assumptions involve a different type of uncertainty.
8.1.3 Outline of Chapter 8
The following section presents a formalized concept of the probability of compliance.
Applications of the probability of compliance concept to a variety of compliance criteria are
discussed. Alternate criteria for compliance are compared and the advantages and
disadvantages of each noted in Section 8.3. Section 8.4 reviews other regulatory examples
of concepts related to reasonable expectation. Conclusions and recommendations are
presented in Section 8.5.
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8.2 PROBABILITY OF COMPLIANCE
8.2.1 Review of the Probabilistic Requirements of 40 CFR Part 191
§191.13(a) contains the following regulatory requirements:
Disposal systems for spent nuclear fuel or high-level or transuranic radioactive wastes
shall be designed to provide a reasonable expectation, based upon performance
assessments, that cumulative releases of radionuclides to the accessible environment
for 10,000 years after disposal from all significant processes and events that may
affect the disposal system shall:
(1) Have a likelihood of less than one chance in 10 of exceeding the
quantities calculated according to Table 1 (Appendix A); and
(2) Have a likelihood of less than one chance in 1,000 of exceeding ten times
the quantities calculated according to Table 1 (Appendix A). [Italics added for
emphasis.]
Table 1 of Appendix A of 40 CFR part 191 defines a set of permissible release limits for the
isotopes of concern. In instructions accompanying this table, guidelines are suggested for the
appropriate use of Table 1 quantities hi conducting a performance assessment. For each
isotope, the ratio of the predicted cumulative release to the accessible environment over
10,000 years to the permissible release limit listed in the table is to be calculated. The ratio
thus obtained is often referred to as the "normalized release" for each isotope. The
normalized releases for all isotopes in the table are then added together to form the sum of
the normalized releases. For example, the limit given in Table 1 for cumulative releases of
each listed plutonium isotope to the accessible environment for 10,000 years after disposal is
100 curies per unit of waste disposed of at the site. If the estimated cumulative release over
10,000 years for the isotope is 70 curies per unit of waste, the normalized release for this
isotope is calculated as 70/100 = 0.70. The normalized release is to be calculated hi this
fashion for each isotope in Table 1.
The sum of the normalized releases for all isotopes disposed of at the site is used for
evaluating the probabilistic requirements 1 and 2 in §191.13(a) given above. To satisfy
§191.13(a)(l), there must be a reasonable expectation that the probability of the sum
exceeding 1 is less than 10 percent. To satisfy §191.13(a)(2), there must be a reasonable
expectation that the probability of the sum exceeding 10 is less than 0.1 percent. In terms of
percentiles, the 90th percentile of the distribution of the summed normalized releases must be
less than 1, and the 99.9th percentile must be less than 10. §191.13(a) requires that estimates
be made for two upper percentiles of the distribution of releases and specifies upper limits on
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these percentiles.
To construct this distribution predicting the disposal system's performance requires
calculation of the summed normalized release (termed simply the "release" hi this discussion)
and its probability of occurrence for each intrusion sub-scenario. The cumulative probability
distribution for the release is obtained by sorting the estimated releases for each sub-scenario
in increasing order. The ordered releases and their associated probability values are then
used to construct the cumulative probability distribution, which is a step function defined
over the range of release values which are called the arguments of the function. The step
function starts at zero for a release value of zero. At each estimated value of the release, the
function steps up by an amount equal to the associated probability of the release. The
cumulative distribution function so defined is a non-decreasing function which cannot exceed
the value of unity (1). The value of the function is equal to the probability that the release is
less than or equal to the value of the argument of the function.
The Complementary Cumulative Distribution Function (CCDF) is defined as the difference
between the cumulative distribution function and the value 1. The probability and percentile
limits set forth hi §191.13(a)(l) and (2) and the reasonable expectation of compliance
concepts are described hi terms of the CCDF of the summed normalized release variable.
The value of the CCDF function gives the probability that the release is greater than the
value of the argument. The CCDF for the random variable R is denoted by the function
F( r ) = Pr{ R > r }.
The function F( r ) is always between 0 and 1, and can never increase as r increases. Using
the CCDF function, §191.13(a)(l) and (2) are commonly written hi statistical terminology as:
(1) F(l) < 0.1; and
(2) F(10) < 0.001.
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An equivalent statement of the regulatory requirements written in terms of the 90th percentile
(denoted by R9) and the 99.9th percentile (denoted by R^) as
(1) R9 < 1; and
(2) Rgg, < 10.
Because the summed normalized release is not only a random variable, but a random variable
with a distribution that can be predicted only with a considerable degree of uncertainty, only
estimates of the probabilities and percentiles (i.e., the quantities written on the left side of the
four inequalities above) can be developed. Each estimated probability and percentile will
have an error of estimation. One interpretation of the concept of reasonable expectation is
that there must be reasonable evidence that all or most of the uncertainty intervals for the
estimated percentiles and probabilities will fall below the designated regulatory requirements.
8.2.2 Statistical Interpretation of the Requirements of 40 CFR Part 191
As noted in the introduction, a random variable with no strict upper bound cannot be
"proved" to satisfy a specific mathematical constraint of the form R < L, where the limit L
is a specified number. This situation is encountered often in the application of statistical
methods to environmental problems. The random variable R denotes a random level of
emissions and the fixed numerical limit L is the maximum allowable emission for the
substance under consideration. If the probability distribution of the emission R is known
exactly, then an exact estimate of a percentile or a probability may be derived by calculus.
If this probability is large (i.e., near 100 percent), then it would be "almost always" true that
the emissions are less than the regulatory requirement.
The order of the inequality inside the probability statement may be reversed. In this case,
the probability that R exceeds L would be required to have a small value. The requirement
then would be stated as
Pr{R > L} < Q,
where Q is a small probability value. Using the CCDF function notation, the regulatory
requirement is stated as
F (L) < Q.
-------
The values Qj =0.1 (10 percent) or Q2 = 0.001 (0.1 percent) are used in 40 CFR section
191.13(a) at two different values of the radioactivity release, Lj = 1 and L, = 10,
respectively.
The required estimate of a probability or percentile can be calculated from the known
distribution and compared directly to the proscribed probability level. This procedure
provides a simple "Yes" or "No" answer for determining compliance with the probabilistic
requirements of 40 CFR part 191.
If the distribution of R is not known exactly, then the above approach is not sufficient to
define compliance with the regulatory requirements in a statistical framework. The estimated
probabilities or percentiles will have an associated uncertainty interval because the exact
distribution of R is unknown. Hence, there will be uncertainty in determining if the
estimated probability is less than the target probability. This uncertainty in determining
compliance could be addressed in two ways. Careful review of the modeling procedures and
the record before the Agency by panels of expert reviewers could be used to increase the
level of confidence hi the reported results. Or, a statistical determination of compliance
could be made from the reported results by conducting a hypothesis test at a specified level
of confidence. These two approaches are discussed in more detail in the following sections.
8.2.3 Use of Expert and Peer Review for Determining Level of Confidence
The use of statistical methods alone cannot assure that predicted releases of radionuclides to
the accessible environment from the WIPP disposal site will comply with the regulatory
requirements of 40 CFR part 191.
A compliance determination would be based on the entire record before the Agency, including both
qualitative and quantitative evaluations. For each of the quantitative requirements hi the regulations,
the determination of reasonable expectation may be based on quantitative analyses as specified in the
regulations, supported by qualitative judgment of the degree of confidence in the reported WIPP
performance assessment and compliance assessment results.
A qualitative assessment of the degree of confidence may be derived from the formal review
process. The review process would comprise, hi part, empirical testing of the model components, a
complete review of the documentation of the model, evaluation of all significant uncertainties, and
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peer review where required by 40 CFR part 194 and involving assessment of the assumptions and
findings reported in the WIPP compliance application.
The testing of model components would include evaluation of the physical basis for the model,
verification of the numerical accuracy of the calculations performed by the model, validation of
approximations made in the calculations, and review of the probability distributions assigned to
uncertain input variables.
The peer review process of the conceptual models provides an additional level of confidence for the
reported numerical results of the performance assessment process. However, the quantitative nature
of the requirements of 40 CFR section 191.13 (a) implies that qualitative evaluation alone is not
sufficient for determining compliance. In the following sections, quantitative measures of the degree
of confidence obtained by statistical procedures are reviewed.
8.2.4 Use of Statistical Methods for Determining Compliance
Many statistical methods can test hypotheses about the distribution of R when the underlying
distribution is not known exactly but must be estimated. The methods primarily differ according to
the assumptions made concerning the form of the unknown distribution. These methods generally
are classified as parametric or non-parametric.
Parametric methods contain specific assumptions concerning the form of the probability distribution
for the variable to be tested. A particular family of probability distributions, indexed by one or
more parameters, is selected as a general model for the observed data. Estimates of the parameters
are made from the data to select from the family a specific distribution to use for the hypothesis
test. A typical parametric example is the use of the family of normal or Gaussian distributions
parameterized by a population mean and a population variance which is the square of the standard
deviation. The normal distribution has the familiar "bell-curve" shape, being symmetric about the
mean of the distribution. Furthermore, the mean, median (or simply stated the middle value), and
mode (that is, the most likely value) are represented by the same point of the distribution.
Parametric methods begin by estimating the unknown mean and variance parameters of the
distribution using the available data, denoted by Xt, X2, ... XN. A typical estimator for the
population mean (M) of the normal distribution is the simple average:
M = £ X / N
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A typical estimator for the population variance (V) is the mean squared variation of each data point
Xj from the estimated mean:
V = £ (Xj M)2 / (N 1)
The population standard deviation is estimated as the square root of the variance of the data
samples.
Each of the estimated parameters has an associated measure of sampling error. In the case of the
estimated mean, its sampling variance (VM) is estimated by dividing the population variance by the
sample size N: VM = V / N. The standard error of the mean SE(M) is calculated as the square
root of VM. Thus, the standard error of the mean is smaller than the population standard deviation
by a factor equal to the square root of sample size N. As the number of observations increases, the
standard error of the estimated mean is reduced by a predictable amount. Estimates of the sample
size required to reduce the standard error of the mean to specified levels may be derived based on
this relationship.
Parametric models such as the normal distribution also are used for testing hypotheses. A typical
parametric hypothesis test for the mean of the normal distribution would begin by estimating the
mean and the standard error of the mean using the procedures above. If it is necessary to determine
if the mean is below a specified upper limit (L), then the hypothesis test is conducted by comparing
the estimated mean to the limit L. Because there is sampling error associated with the estimated
mean, the actual comparison is made using the upper bound of the confidence interval for the
estimated mean:
M + k,SE(M) < L,
where le, is selected to provide the appropriate a% confidence level for the test.
The lognormal distribution is often used as a parametric model for environmental quantities. This
distribution, which is defined for positive variables only, is not symmetric, but highly skewed to the
right or left (If there is a long tail on the right side of the distribution which extends up to large
values of the variable then it is referred to as positively skewed). The lognormal probability
distribution is obtained by an exponential transformation from a normal variable. If the variable X
has a normal distribution, then the transformed variable Y = ex is said to have the lognormal
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distribution, because the (natural) logarithms of Y values have a normal distribution. Applying the
exponential transformation to the mean of the normal distribution, which is also the median, with
probability of 0.5 below the median and 0.5 above the median, yields an estimate of the median of
the lognormal distribution.
Parameter estimates for the lognormal distribution are typically derived by taking logarithms of the
lognormal observations (Y,, Y2, ..., YN), then applying the estimators defined above for the mean
and variance of the normal distribution. For example, the simple average of the logarithms of the
Yj values is an estimate of the mean (and median) of the transformed normal distribution. The
exponential function is then applied to the average of the logarithms to calculate an estimate of the
median of the associated lognormal distribution. This procedure often is described as calculating the
geometric mean of the lognormal observations, which provides the same estimate for the median of
the lognormal distribution. Furthermore, if experts provide judgments on the upper and lower
percentiles of the distribution, an estimate can be made for the variance of the lognormal
distribution. Knowledge of these two parameters (the mean and variance) completely defines the
distribution.
Parametric hypothesis tests for the lognormal follow the same general procedures as for the normal
distribution. Tests for the median of the lognormal distribution are simpler than for the mean of the
lognormal and follow the same procedure as the test for the mean of a normal distribution. The
mean of the logarithms of the observations (plus a multiple of the standard error) is compared to the
logarithm of the regulatory limit L. Tests for the mean of the lognormal require more detailed
calculations to determine the standard error of the estimated mean.
The use of parametric methods is based on several assumptions. The most important of these is the
selection of a particular family of distributions as a probability model for the data. Generally, a
large number of observations is required to determine if this assumption is correct. However, even
with a small number of observations, it may be obvious that certain families of distributions may not
be applicable.
Non-parametric statistical tests, by comparison, make minimal assumptions concerning the specific
probability distribution for the observed random variable. The usual assumption that the variable
has a probability distribution with a defined mean and variance is very general. In terms of modern
statistics, non-parametric statistical tests of hypotheses are considered "robust" with respect to
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assumptions made concerning the form of the distribution of R. The results of a robust statistical
test will be affected less than a non-robust test if the actual distribution of the data departs from the
assumed distribution.
Specifically for the case when the distribution of the normalized release is not known exactly, 40
CFR sections 191.13(a)(l) and (2) may be written as two statistical hypotheses (Ht and H2) to be
accepted or rejected by applying a statistical test. Using the CCDF approach, these hypotheses can
be stated as:
(1) H,: F( 1 ) < 0.1; and
(2) H2: F( 10) < 0.001.
Or, in terms of percentiles, the regulatory requirements may be expressed as:
(1) H!: R9 < 1; and
(2) H2: R^ < 10.
More generally, a single hypothesis referred to as the "joint null hypothesis" may be written in
terms of the CCDFs as
H0: F( L;) < Q;, for i = 1, 2;
or, in terms of percentiles as
H0: RPi < L;, for i = 1, 2;
with Pj, Qi and L, defined as above. The joint null hypothesis is to be tested against all alternatives.
If there is no sufficient statistical evidence to reject the null hypothesis, then the null hypothesis is
said to be "accepted." Note that the statistical hypothesis testing procedure does not "prove" that
the null hypothesis is true; it states only that no sufficient statistical evidence could be found to
reject the null hypothesis.
Hypothesis tests concerning probability statements such as H0: F( Lj ) < Q; or H0: RPi < L; when
the distribution is not known with certainty are generally based on samples obtained from the
distribution of the random variable. The degree of "truth" obtained by applying these hypothesis
tests for probabilities or percentiles of predicted future realizations of the random variable R is more
difficult to quantify. In this case, the "truth" of the results can be assessed only probabilistically.
The probability of compliance with 40 CFR part 191 may be estimated using simulation results to
characterize the uncertainties involved in estimating the future distribution of the random release.
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8.2.5 Use of Sampling Methods
The performance assessment modeling process contains two stages of analysis. In the first stage, a
set of scenarios is selected for evaluation. Subjective probabilities are assigned to the scenarios. In
the 1992 performance assessment (SNL92), scenario probabilities are generated based on the results
of expert panels, which addressed the likelihood of human-intrusion processes and the deterrent
effect of markers. In the second stage of the analysis, the performance assessment model evaluates
the summed normalized release for each scenario. At this stage, specific values for the physical
parameters of the model must be selected. Rather than being assigned a single value, each
important input parameter has been assigned a probability distribution that reflects the uncertainty in
the value of the parameter. (These distributions have been assigned independently for each
parameter.) The model is run repeatedly, using a Monte Carlo simulation approach to generate a
distribution of possible values for the summed normalized release.
This approach is widely accepted, and is currently being followed by DOE for the WIPP
performance assessment and both DOE and NRC for the Yucca Mountain performance assessments.
However, other sampling strategies, generally known as "importance sampling" (WU93) have non-
equal probabilities in order to concentrate the samples hi the region of parameter space where the
models have the greatest sensitivities to parameter variations. The purpose of importance sampling
is to increase sampling efficiency in order to reduce the computational workload while minimizing
the need for model oversimplification or reduced coverage of the parameter space. Furthermore, 40
CFR part 191 requires only the determination of two points on the CCDF at probabilities of 0.1 and
0.001. The rest of the CCDF is not significant to determine these compliance points. (That is not
to say that only two points are considered in the compliance determination; other information is not
discarded. It is only after all information has been organized into a CCDF that it reduces to this
two point test. Because the CCDF algorithm orders the scenarios by calculated release, the
approach to the release limits is closest at these two points.) Techniques such as importance
sampling might be able to develop the compliance points with far fewer samples than the Latin
Hypercube Sampling (LHS) Monte Carlo technique selected by DOE for WIPP performance
assessment (SNL85).
In the general Monte Carlo approach, a single value for each input parameter would be sampled
independently from the appropriate distribution for the variable within each run of the model. The
LHS method is based on dividing the parameter distribution into strata which are intervals of equal
probability. An interval is selected, then a random sample is drawn from the selected interval of the
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distribution. The effect of the LHS procedure is to ensure a more uniform spread of sampled values
over the entire range of the parameter distribution than might be obtained by simple Monte Carlo
sampling. The LHS procedure would be unnecessary if large sample sizes could be used for the
Monte Carlo simulation. Due to the complexity and sometimes costs of the computer models, only
a relatively small number of samples are used currently. For small sample sizes, the LHS
procedure provides a more efficient sampling technique hi comparison to alternatives.
Many types of probability distributions are used to describe the uncertainty in the WIPP
performance assessment model input parameters. Physical parameters used in the model are
assigned distributions based either on available data or the subjective opinion of experts. Some
distributions, based on actual data, do not belong to a known family of distributions. These
distributions are constructed directly from the observed data by forming the cumulative distribution
function. Other families of distributions selected for the model input parameters typically include
the beta, gamma, exponential, normal, lognormal, uniform, loguniform, discrete uniform, binomial,
and Poisson distributions. The beta, uniform, and loguniform distributions are appropriate for
parameters that are assumed to lie in an interval between two known endpoints. The gamma,
exponential, and lognormal distributions are appropriate for positive parameters with distributions
which are skewed to the right, with a long tail extending to higher values of the parameter. The
first six families of probability distributions are defined for variables that can take a continuous
range of values. The discrete uniform, binomial, and Poisson distributions are appropriate for
parameters that have only integer values.
In summary, the performance assessment modeling process may be described as a two-step
procedure:
1. A complete set of possible future scenarios, denoted by the set {SJ5 j = 1, ..., J}, is
developed conceptually with the understanding that all possible outcomes have
been included. The scenarios are then assigned probability distributions, which
must sum to 1 over all scenarios.
2. At this stage, a specified number of independent samples are selected by the
LHS procedure, for example, from the subjective uncertainty distributions
assigned to the input parameters of the model. Specified pair-wise correlations
can be contained in the analysis. The model is then run for each LHS sample S,
generating N CCDFs for all scenario Sj. Using all scenarios j, and each list (or
vector) of subjective distributions on model input parameters, N distribution
functions are computationally generated (where N = number of sample vectors
taken using the LHS procedure.) For the scenarios analyzed, each LHS model
run generates a different CCDF Fn for the distribution of the summed normalized
release.
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8.2.6 Conditional Probabilities of Compliance
It may be possible to reduce the set of N LHS CCDFs for each scenario to a single
probability distribution. This reduction may be accomplished in two steps. First, the set of
LHS curves for each scenario can be reduced to a single distribution for the normalized
release. The resulting distribution would be conditional in the sense that the scenario would
be assumed to occur with a probability of 100 percent. Second, the conditional distributions
for each scenario would then be weighted by the scenario probabilities and combined into a
single (unconditional) curve for the WIPP site. This approach was not applied since single
probability estimates were not used. Instead, probability distributions were assumed by DOE
for the scenarios and a mean CCDF was calculated for the collection of scenarios using a
sampling based approach.
Several methods have been proposed to reduce the set of LHS CCDFs to a single curve
involving all scenarios (SNL92, ESL92). The methods for reducing to a single CCDF
include using these aspects of the LHS CCDF values:
1. the mean which is the simple arithmetic average (i.e. a vertical averaging of the
generated CCDF curves;
2. the median which is the 50* percentile;
3. selected upper percentiles (those higher than the median); or
4. selected order (or rank) statistics (e.g., the maximum which is the highest of the
ordered observations, the second highest, etc.)
As stated, 40 CFR sections 191.13(a)(1) and (2) require that estimated releases be evaluated
at the two values of the summed normalized release: R = Lj = 1 and R = l^ = 10.
However, it is informative to generate the entire CCDF for the scenario in graphic form.
An example of the reduction of 10 LHS CCDFs to a single curve using the maximum of the
set of LHS CCDFs is shown in Figure 8-1. The maximum curve (an example of alternative
4 above) is the solid line labeled A hi the figure. The maximum curve, defined as a function
of the normalized release R,is also defined to have the highest value of the set of LHS CCDF
probabilities at each value of the release.
Figure 8-1 also includes a graphic representation of the requirements of §191.13(a)(l) and (2)
indicated at the corners (1 and 2, respectively) of the step function (line B) in the upper right
corner of the figure. The regulations proscribe probabilities of release in the region above
and to the right of this step function. In this example, the maximum curve complies with
§191.13(a)(l) at R = 1 = 10°, while it slightly exceeds §191.13(a)(2) at R = 10 = 101
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Maximum curve
Requirements of 40 CFR 191
1 = 40 CFR 191.13(a)(1)
2 = 40 CFR 191.13(a)(2)
Intersection points
10-* io'3 itr* 10" 10"
Summed Normalized Releases, R
Figure 8-1. Ten Hypothetical LHS CCDFs, with Maximum
Although graphic representation of the entire curve helps in visualizing the curve reduction
procedures, the entire curve need not be calculated to determine compliance in the strict
statistical interpretation of the regulations. Returning to the discussion of Figure 8-1, the
dotted line C marks the value of R = 1 = 10°. To complete the requirement in
§191.13(a)(l) a similar line could be drawn to mark the value of R = 10 = 101. The
equivalent of reduction to a "single curve" is obtained by applying one of the four methods
of curve reduction noted above to the set of point estimates shown by the large dots along
line C. The set of 10 points along line C represent 10 independent estimates of the
probability that the normalized release exceeds 1. Although equivalent calculations may be
performed for the entire CCDF, only points on the curve noted in the figure and along a
similar line R = 10 = 101 need to be considered in determining compliance. Both sets of
point must be considered for the 40 CFR part 191.13(a) compliance test.
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Figure 8-2 shows the mean (A), 90th percentile (B), and median (C).of a set of LHS CCDFs.
The use of the mean was considered as the first alternative method for reducing the set of
LHS CCDFs to a single CCDF, while the median was the second alternative, and the 90th
percentile is an arbitrarily selected example of the third alternative. In Figure 8-2, all three
reduced curves indicate compliance with both §§191.13(a)(l) and (2). Note, however, that
the mean is below the median for small values of R, indicating an asymmetric (non-normal)
distribution. Also, it exceeds the 90th percentile at higher values of R. At very high values
of R, the mean lies close to the maximum curve shown in Figure 8-1. Calculation of the
unconditional probability of compliance for all scenarios is based on the conditional
distribution of intersection point values for each scenario. The range of the distribution of
these values reflects the range of uncertainty due to the model parameters for the scenario at
hand. The mean, median, and percentiles of this distribution, and the maximum, second
highest, etc., of the intersection point values, are all possible methods for reducing the set of
N intersection points to a single point estimate.
cc
A
0)
V)
CO
0)
0)
cc.
.a
03
O
CL
A = Mean
B = 90th percentile
C = Median
10'" 1(T* 10
10
Summed Normalized Releases, R
Figure 8-2. Mean, 90th Percentile, and Median Curves from Set of 10 LHS CCDFs
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Uncertainty in these point estimates should be considered in the assessment. Hence, the
spread of the distribution of intersection points must also be quantified. The combination of
the selected point estimate and an estimate of the spread of the distribution is required to
characterize the degree of confidence for the determination of compliance.
The following section addresses the derivation of the unconditional probability of compliance
from the collection of probabilities for all scenarios and subjective uncertainties on input
parameter values. The advantages and disadvantages of the alternative criteria for
determining compliance are related to the advantages and disadvantages of each form of
curve reduction. The advantages and disadvantages of each form of curve reduction are
discussed hi more detail hi Section 8.3.
8.2.7 Unconditional Probability of Compliance with the Containment Requirements
The probability of compliance may be estimated using one of the four curve reduction
methods noted in the preceding section. The uncertainty reflected in the multiple LHS
simulations performed for each scenario and uncertain model parameter results in a
distribution of point estimates for each release magnitude.
Derivation of the unconditional probability of compliance requires use of the scenario
probabilities and simulation of the uncertainty in the scenario probabilities. The use of
subjective methods to resolve scenario probabilities and their uncertainty has only just begun.
Recently, several expert panels were convened and assigned the task of estimating human
intrusion scenario probabilities. The unconditional probability of compliance cannot be
estimated using the computational approach until uncertainty ranges have been assigned to the
scenario probabilities.
The uncertainty distribution for the scenario probabilities, when available, may be used to
simulate a set of N scenario-weighted estimates of the probability of exceeding the various
compliance release limits. Again, this set of estimates will need to be reduced to a single
point estimate for the probability of compliance, and a measure of the uncertainty interval
will be needed for this estimate. Formally, this would require a second application of the
four curve reduction methods discussed above. However, it may be sufficient simply to
display the resulting distribution of estimates of the probability of compliance graphically to
determine if the uncertainty interval is sufficiently high (or low).
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8.3 COMPARISON OF ALTERNATIVE CRITERIA FOR COMPLIANCE
8.3.1 Advantages and Disadvantages of Alternative Compliance Criteria Using a Central
Point Measure
In section 8.2, a set of N point estimates for each scenario with parameter uncertainty was
developed for the probability or percentiles contained in the regulations by determining the
intersection of the N LHS CCDFs with the appropriate set of regulatory requirements. This
set of N independent estimates for the percentiles RPi and/or probabilities F( L, ) used in the
regulatory requirements determines the uncertainty distribution for each measure of
compliance. These estimates must then be compared to the appropriate limiting value (L;
and/or Qj, respectively) stated in §191.13(a). Because a distribution of estimates is produced
by this procedure, it is desirable to reduce this set of estimates to a single test statistic for
determining compliance.
After observing N independent samples from the LHS simulations, the probability of
compliance may be determined by a variety of methods. The simplest method is to estimate
a central point from the distribution of estimates. Alternative methods for determining a
central point are listed in Table 8-1, with a summary discussion of the merits and
disadvantages of each. The methods include the simple arithmetic mean (or "center of
gravity",) the weighted mean, and the median.
Alternative 1 in the list of curve reduction methods presented in the previous section suggests
taking the arithmetic average of the appropriate point estimates to use as the test statistic for
comparing to the limit in requirement i. The use of a weighted mean is not necessary for
conditional analysis due to the equally likely nature of LHS samples.
Alternative 2 is a variation of alternative 1 to account for unequally weighted scenarios.
Alternative 3 uses the median fu(.5) as the central point estimate for the test statistic. The
median is a more robust point estimate, but this point estimate will be smaller than the true
expected value (mean) if the distribution is skewed to the right (e.g., large values are
expected to be more common than small values).
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Table 8-1. Measures of the Central Point
Measure
Pros
Cons
Arithmetic Mean = M
• The true expected value if
probabilities are equal
• Easy to calculate
• Fair weight to low probability
but catastrophic failures
Inappropriate for distributions
highly skewed to right
Low robustness; one "outlier"
can dramatically change result
If used as criterion, level of
confidence is unknown
Probability Weighted
Mean = M^
• The true expected value if
probabilities are not equal
• Easy to calculate and explain
• Fair weight to low probability
but catastrophic failures
• Inappropriate for distributions
highly skewed to right
• Low robustness; one "outlier"
can dramatically change result
• If used as criterion, level of
confidence is unknown
Median = MED = f,
A true center: there are 50%
above and 50% below the
median
Non-parametric: no distribution
shape is assumed
Known to be very robust; not
affected by "outliers"
Easy to calculate for reasonably
large sample sizes
• Lower than expected value for
distributions skewed to right
• If used as criterion, level of
confidence is unknown
• Discounts low probability
catastrophic failures,
information is lost
Regardless of the decision to use the mean or median as a central point individually, neither
test statistic for determining compliance would reflect the uncertainty indicated by the spread
of the distribution of estimates. If the concept of reasonable expectation is interpreted to
indicate due consideration of uncertainty, then single point test statistics that are measures of
the central point of the distribution of estimates are not adequate for determining compliance.
If the median is used as the central point, and the median is barely below the required limit,
then there is almost a 50% chance that the regulatory limit would be exceeded. In this case,
evidence from the spread of the distribution dictates that the probability of compliance can be
no larger than 50%. Furthermore, the level of assurance or confidence at which this rather
weak statement of compliance can be made is unknown because the uncertainty in the
estimate of the median has not been considered.
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To reflect the uncertainty in the distribution of estimates, appropriate measures of the spread
of the distribution are required. Alternative estimators for the spread of the distribution of
estimates are listed in Table 8-2, with a summary of the advantages and disadvantages of
each estimator. Estimates of the spread of the distribution include the population standard
deviation, the mean absolute deviation from the median (MAD), and the interquartile range
(IQR) which is the difference between the 75th and 25th percentiles. The population standard
deviation, based on sums of squared deviations from the mean, is less robust than the mean.
Small changes in values far from the median can have a large influence on this estimator.
The mean absolute deviation from the mean is more robust but is only appropriate for
symmetric distributions. The interquartile range is more robust and appropriate for
asymmetric distributions.
Possible numerical measures of compliance which can be used to reflect uncertainty in
reduced data sets are summarized in Table 8-3. One test statistic for determining compliance
is based on the mean and its standard error of estimation. The test statistic is defined as the
upper end of the 95 % confidence interval for the estimated mean. Use of the upper
confidence bound for the sample mean is one of the statistical compliance criteria suggested
by EPA for evaluating the attainment of soil clean-up standards (EPA89). The Nuclear
Regulatory Commission (NRC) has adopted EPA clean-up criterion in the guidance for
determining compliance with the requirements for license termination (NRC92). In these
regulatory applications, actual measurements of residual contamination after site clean-up are
used for computing the sample mean and standard error of the sample mean.
Note that the "confidence interval" found by this procedure provides an uncertainty range for
the estimate of the mean, not for the spread of the population values. The mean plus a
multiple of the standard error of the mean can be used to form this uncertainty interval. The
advantages and disadvantages of this test statistic are shown hi the first row of Table 8-3.
Determination of the appropriate value of the multiplier "k" to yield the desired confidence
level is based on the assumption of a Student-t distribution. This assumption may be
inappropriate for asymmetric distributions.
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Table 8-2. Measures of Spread or Dispersion
Measure
Pros
Population Standard
Deviation = Sigma
Maximum likelihood estimate
for variance of normal
distribution
Easy to calculate, even for
weighted samples
Low robustness; one
"outlier" can
dramatically change
result
Inappropriate for
skew distributions
Mean Absolute Deviation
from Median = MAD
Non-parametric: no
distribution is assumed
Provides robust estimate of
population Sigma
Provides robust estimate of
Standard Error (SE) of
estimates of means and
percentiles obtained by
simulation method
Known to be very robust;
insensitive to "outliers"
Easy to calculate for
reasonably large sample sizes
Link to sample
variance depends on
distribution
Inappropriate for
skew distributions
Interquartile Range
IQR = f. 7S - f.^
Non-parametric: no
distribution is assumed
Provides robust estimate of
population Sigma
Provides robust estimate of
Standard Error (SE) of
estimates of means and
percentiles obtained by
simulation method
Known to be very robust;
insensitive to "outliers"
Easy to calculate for
reasonably large sample sizes
Link to sample
variance depends on
distribution
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Table 8-3. Numerical Measures of Compliance
Measure
Pros
Cons
Classical Confidence
Limit on Weighted Mean
UCL = MM + k'SECMJ
Classical upper bound for
the expected value
Standard Error (SE) reflects
uncertainty in estimate of
mean
k can be adjusted for desired
level of confidence
Easy to calculate, even for
weighted samples
May be inappropriate for
skewed distributions
Low robustness; one
''outlier" can change result
dramatically
Uncertainty in the estimate of
SE is not addressed
Level of confidence is based
on the t-distribution
assumption
* Percentile = f
(Includes Median = f J0)
Percentile p reflects
dispersion due to uncertainty
in parameters and
probabilities
Non-parametric: no
distribution is assumed
p can be adjusted to desired
probability point
Easy to calculate, even for
weighted samples
Extreme upper percentiles
require large sample sizes
Uncertainty in the estimate of
fp is not addressed
Classical Confidence
Limit on Upper Percentile
UCL, = fp +
k»Sigma(fp)
Classical upper bound for an
upper percentile of the
population distribution
Percentile p reflects
dispersion due to uncertainty
in parameters and
probabilities
p can be adjusted to desired
probability point
Standard Error (SE) reflects
uncertainty in estimate of
percentile
k can be adjusted for desired
level of confidence
Extreme upper percentiles
require large sample sizes
Uncertainty in SE( fp)
estimate is not addressed
Level of confidence is based
on the t-distribution
assumption
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Table 8-3. Numerical Measures of Compliance (Continued)
Measure
Pros
Cons
Classical Tolerance Limit
on Upper Percentile
t =
kp»Sigma
Tolerance limit addresses
uncertainty in estimation of
UCL
To account for uncertainty in
SE, k is adjusted higher
kp can be found for desired
level of assurance for
percentile p
Relatively easy to calculate,
even for weighted samples
Inappropriate for skewed
distributions
Low robustness; one
"outlier" can dramatically
change result
Level of confidence is based
on the t-distribution
assumption
Non-parametric Tolerance
Limit on Upper Percentile
Maximum of N samples
Non-parametric: no
distribution is assumed
Use of order statistics
reflects uncertainty in
parameters and probabilities
Level of assurance requires
no distribution assumption
Maximum out of N provides
a non-parametric upper
tolerance limit for upper
percentiles
Easy to calculate
Low robustness; one
"outlier" can dramatically
change result
Moderately large sample size
for acceptable tolerance
levels
Difficult to define tolerance
limit if samples have unequal
weights
Non-parametric Tolerance
Limit on Upper Percentile
f[j:N] ^j* highest
of N samples
Non-parametric: no
distribution is assumed
Use of order statistics
reflects uncertainty in
parameters and probabilities
Level of assurance requires
no distribution assumption
j* highest out of N provides
a robust non-parametric
upper tolerance limit for
upper percentiles
Higher robustness against
"outliers" as j increases to
median
Easy to calculate
Larger sample size for
acceptable tolerance values as
j increases
Difficult to define tolerance
limit if samples have unequal
weights
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A similar central point test statistic can be constructed for the median, i.e., the estimated
median plus a multiple of the standard error of the median. Given that k is selected for 95%
confidence, if the upper bound of the confidence interval on the median is less than the limit
in the requirement, there will be 95 % confidence that the conditional probability of
compliance is at least 50%. This is a weak statement concerning compliance, but the
probability of compliance can be improved by increasing the percentile from the median at
0.50 to a higher level. Thus, the median is a special case of the family of percentile
estimators and their associated test statistics, which are the subject of rows 2 through 6 of
Table 8-3. This family of test statistics is discussed in the following section.
8.3.2 Advantages and Disadvantages of Alternative Compliance Criteria Using Percentiles
A simple way to determine if the higher values of the distribution of intersection point values
exceeds the regulatory requirements is to count the number of conditional LHS values which
are higher than the limiting value. The ratio of this number to N is an estimate of the
proportion of LHS runs that are not in compliance. Hence, the probability of compliance is
1 minus this ratio. For example, if there are 100 LHS samples and 10 values exceed the
limit, then an approximate estimate of the proportion of samples exceeding the limit is 10
percent. This indicates that the probability of compliance is near 90 percent (if 100 points
are adequate to describe the uncertainty distribution in the mean CCDFs.)
A more formal way to do this comparison is to specify that the p* percentile of the
distribution of intersection point values is below the regulatory limit. This value, denoted by
fp hi row 2 of Table 8-3 was discussed under alternative 3 of Section 8.2. The following
statement may be made concerning this percentile: If fp is less than the limit in the
requirement, then the probability of compliance will be at least as large as p. The level of
confidence for making this statement is unknown, since the sampling error for the estimated
percentile has not been addressed. Additional advantages and disadvantages of the percentile
estimator are presented in row 2 of Table 8-3.
The test statistics in rows 3 through 6 of Table 8-3 are designed to provide a known level of
confidence that the estimated percentile is less than the limit in the requirement. In row 3 of
Table 8-3, the classic 95% confidence interval for the p^ percentile is formed by adding a
multiple of the standard error to the percentile estimate. As for the confidence interval on
the mean in row 1, the multiplier k can be adjusted to provide a 95% confidence interval.
The level of confidence thus obtained is based on the assumption of a t-distribution.
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The classic tolerance interval test statistic shown hi row 4 of Table 8-3 is related
conceptually to the confidence interval on the upper percentile discussed hi row 3. Although
this test statistic utilizes the estimated mean and standard deviation of the population of
intersection points, it is applied as an upper tolerance limit for percentiles of the distribution.
Due to the asymmetry of the distributions encountered here, the tabulated values for kp
provide only a rough approximation of the true level of tolerance for the test.
The order statistics of the sample of intersection points may be used to provide non-
parametric upper tolerance bounds for the percentiles of a distribution (GLI78). The use of
the maximum as a test statistic is discussed hi row 5 of Table 8-3. For example, it may be
demonstrated that the maximum value hi a sample of 90 independent values from the same
distribution has at least a 99% chance of being larger than the 95th percentile of the
distribution. Thus, the maximum in a random sample of at least 90 observations is said .to
provide a 99% upper tolerance limit for the 95th percentile. A simple proof proceeds as
follows.
By definition, the probability of exceeding the 95th percentile is 5 %. If observations are
made independently, each has a 5% chance of exceeding the 95th percentile. As more and
more observations are collected, the chance that at least one of them will exceed the 95th
percentile increases with the number of observations. For a sufficiently large sample, there
will be a high probability that the maximum hi the sample exceeds the 95th percentile. For
example, this probability is calculated for N=90 observations by the formula
1 .95N = 1 - .9590 = 0.9901.
Thus there is over a 99% chance that the maximum of 90 independent observations will
exceed the 95th percentile, regardless of the distribution. If the maximum of the sample is
lower than the requirement of §191.13(a), then there is at least a 99% chance that the 95th
percentile of the distribution lies below the regulatory requirement. In statistical terms, the
maximum value in a random sample of size 90 provides a robust 99% upper tolerance bound
for the 95th percentile of the distribution.
By a similar argument, the second highest value hi a random sample of at least 76
observations has at least a 90% chance of exceeding the 95th percentile. And, the third
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highest in a sample of at least 75 has a 75 % probability of exceeding the 95th percentile of
the distribution. These test statistics based on the higher order statistics are discussed in row
6 of Table 8-3.
The advantage of using the order statistics of the estimates obtained from the intersection
points of the LHS CCDFs with the regulatory requirements is that no assumptions are
required concerning the specific form for the distribution. The distribution need not be
symmetric. Appropriate tolerance bounds for any percentile may be found using the order
statistics of the intersection points. Non-parametric tolerance limits for the upper percentiles
of the distribution do not require specification of the adjustable multiplier k; hence, no
reliance on tables based on the t-distribution assumption are necessary. For this test, the
sample size itself is the adjustable parameter.
8.4 OTHER REGULATORY CONSIDERATIONS
This section examines both national and international regulatory requirements for parallels to
the concept of "reasonable expectation."
8.4.1 Environmental Protection Agency
8.4.1.1 40 CFR Part 268, Land Disposal Restrictions.
The Land Disposal Restrictions identify hazardous waste that is restricted from land disposal
and define those limited circumstances under which an otherwise prohibited waste may
continue to be land disposed. §268.6 sets out requirements for exemption petitions which if
granted allow land disposal of a prohibited waste. These petitions are generally referred to
as "no migration petitions."
The regulations require that the demonstration include the folio whig components:
(1) An identification of the specific waste and the specific unit for which the
demonstration will be made;
(2) A waste analysis to describe chemical and physical characteristics of the subject
waste;
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(3) A comprehensive characterization of the disposal unit site including an analysis of
background air, soil, and water quality;
(4) A monitoring plan to detect migration at the earliest practicable time; and
(5) Sufficient information to assure the (EPA) Administrator that the owner or operator of
a land disposal unit receiving restricted waste(s) will comply with other applicable
Federal, State, and local laws.
Treatment of uncertainty is addressed in Section 268.6(b)(5), which states:
"An analysis must be performed to identify and quantify any aspects of the
demonstration that contribute significantly to uncertainty. This analysis must include
an evaluation of the consequences of predictable future events, including, but not
limited to, earthquakes, floods, severe storm events, droughts, or other natural
phenomena."
The EPA guidance manual offers further instruction on dealing with uncertainty (EPA92).
The manual states that a petitioner must identify and evaluate the impacts of predictable
future events that could contribute to or result in inadequate waste isolation, such as
earthquakes and resulting ground motion, floods and droughts, severe storm events, climatic
fluctuations, geologic activity, and likely human-induced processes and events which may
affect the isolation capability of the unit, such as disturbance of the hydrologic regime and
future land uses.
The manual notes that the level of detail required in individual petitions will depend on site-
specific factors. Neither the manual nor the regulations provide limits or assumed values for
any specific parameters such as future populations, land use, or climatic changes.
8.4.1.2 40 CFR Part 148, Hazardous Waste Injection Restrictions (Underground
Injection Control)
40 CFR part 148 codifies EPA's regulatory framework for implementing the 40 CFR part
268 land disposal restrictions for hazardous waste that is disposed in Class I injection wells.
Subpart C, Petition Standards and Procedures, sets out the requirements for seeking a "no
migration" petition under these regulations. §148.20 requires that the petitioner demonstrate,
with a reasonable degree of certainty, that hazardous constituents will not migrate as long as
the waste remains hazardous, by demonstrating either (1) that the injected fluids will not
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migrate within 10,000 years, or (2) that before the injected fluids migrate, they will no
longer be hazardous because of attenuation, transformation, or immobilization within the
injection zone.
§148.21 lists the information that must be submitted in support of a "no migration" petition.
It states:
An analysis shall be performed to identify and assess aspects of the demonstration that
contribute significantly to uncertainty. The petitioner shall conduct a sensitivity
analysis to determine the effect that significant uncertainty may contribute to the
demonstration. The demonstration shall then be based on conservative assumptions
identified in the analysis.
8.4.2 Nuclear Regulatory Commission
U.S. Nuclear Regulatory Commission regulations for the disposal of high-level waste in
geologic repositories (10 CFR part 60) also address long-term uncertainty issues. These
regulations require a finding that issuance of a license for such a geologic disposal system
will not constitute an unreasonable risk to the health and safety of the public. Subpart E of
the regulation provides performance objectives, site criteria, and design criteria.
The general standard in 10 CFR part 60 for judging whether the performance objectives and
criteria are met is "reasonable assurance." §60.101 (a)(2) of Subpart E characterizes this
general standard as follows:
While these performance objectives and criteria are generally stated in unqualified
terms, it is not expected that complete assurance that they will be met can be
presented. A reasonable assurance, on the basis of the record before the
Commission, that the objectives and criteria will be met is the general standard that is
required. For 60.112, and other portions of this subpart that impose objectives and
criteria for disposal system performance over long times into the future, there will
inevitably be greater uncertainties. Proof of the future performance of engineered
barrier systems and the geologic setting over time periods of many hundreds or many
thousands of years is not to be had in the ordinary sense of the word. For such long-
term objectives and criteria, what is required is reasonable assurance, making
allowance for the time period, hazards, and uncertainties involved, that the outcome
will be in conformance with those objectives and criteria. Demonstration of
compliance with such objectives and criteria will involve the use of data from
accelerated tests and predictive models that are supported by such measures as field
and laboratory tests, monitoring data, and natural analog studies.
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§60.112 requires that the disposal system be designed to assure that releases of radioactivity
to the accessible environment conform to generally applicable EPA standards.
In 10 CFR part 60, the general standard for judging whether the performance objectives and
criteria are met is "reasonable assurance," while 40 CFR part 191 uses the term "reasonable
expectation." In its comments to EPA on the Advance Notice of Proposed Rulemaking
(ANPR) on compliance criteria for 40 CFR part 191, NRC urged the Agency to reexamine a
position taken in 1985 hi 50 FR 38071 that "reasonable expectation" was different from
"reasonable assurance" (NRC93a).1 It was NRC's view that the terms were similar, and the
Commission suggested that EPA explain any differences in the rulemaking process.
In its comments on the ANPR, the NRC also noted the difficulty in attempting to apply
numerical standards (such as a specified confidence level) where data are insufficient to draw
rigorous statistical conclusions. The NRC went on to say, "Because a specific statistical test
cannot be applied, a more general qualitative 'level of confidence' should be the required
measure of compliance. DOE should be required to demonstrate (by a preponderance of the
evidence) the required level of level of confidence — e.g., 'reasonable assurance' — in future
performance of the disposal facility."
In the development of a manual for determining compliance with license termination
requirements, NRC has suggested methods for comparing soil and surface measurements of
residual contamination to mandated clean-up standards (NRC92). One suggested method is a
comparison of the upper bound on the 95% confidence interval to the regulatory standard.
NRC notes that the comparison may have three possible outcomes:
(1) If the sample mean exceeds the standard for clean-up, then the site is not in
compliance and further cleaning is required.
(2) If the sample mean is lower than the clean-up standard, but the upper limit of
the 95 % confidence interval for the mean is not below the standard, NRC
offers two choices: the site operator may make more measurements to reduce
the width of the confidence interval; or the operator may decide to re-clean the
site.
1 A similar standard - "reasonable degree of certainty" - exists in the RCRA regulations. DOE recently
requested that EPA consider documenting in the 40 CFR part 194 rulemaking that the terms were equivalent for
demonstrating regulatory compliance for geologic repositories. This standard requires the Agency to consider
only future events that could reasonably be predicted; proof beyond a reasonable doubt is not required.
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(3) If the upper limit of the 95% confidence interval for the mean is below the
standard, then the site is considered to be in compliance at the 95% confidence
level.
8.4.3 Department of Energy
DOE also commented on the 40 CFR part 191 Compliance Criteria ANPR (DOE93). While
the Department advocated retaining the concept of "reasonable expectation," it took the
position "that additional attempts to specify a numerical or statistical test of compliance
would not be productive." Rather, DOE argued that the degree of confidence should not be
predetermined; instead, it should be based on all considerations reflected in the record. DOE
position was similar to that taken by the NRC. Because, in DOE's view, there are no
statistical tests appropriate to the kinds of information contained in the performance
assessment, EPA should choose a "more general level of confidence" based on a substantial
understanding of the disposal system and the surrounding environment and on peer review of
the WIPP performance assessments.
8.4.4 Non-U.S. Disposal Systems
Substantial work on performance assessment and treatment of uncertainty has been
undertaken in several countries. These are primarily "total systems" studies, addressing
more than just numerical techniques. For example, site selection issues are considered; i.e.,
one way of reducing uncertainty is to consider sites that are easy to characterize as opposed
to sites that look good but are difficult to characterize.
However, finding useful parallels addressing the issue of "reasonable expectation" in
programs involving non-U.S. geologic repositories has not been possible because these
programs are not nearly as close to actual disposal as those in this country. In 1990, Pacific
Northwest Laboratory (PNL) investigators compared programs on repositories for high-level
waste2 and spent fuel in eight countries including Belgium, Canada, France, Germany,
Japan, Sweden, Switzerland, and the United Kingdom (PNL90). Conclusions reached by the
PNL team include:
• The United States has one of the most developed disposal system concepts and
one of the earliest scheduled disposal system startup dates.
2 It should be emphasized that this report focused on high-level wastes, not TRU wastes.
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• The United States has the most prescriptive regulations and performance
requirements for the disposal system and its components.
• Most countries have established only general performance requirements for
repositories.
• Some countries do not believe that detailed performance requirements and
regulations are required and do not plan to develop them.
• No disposal system is scheduled for operation for at least 20 years.
• Only three countries have selected the host rock for their disposal system.
Regulatory status and approach to safety in each country surveyed by PNL are summarized
in Table 8-4.
Table 8-4. Regulatory Status and Approach to Safety in Foreign
Geologic Repositories
Country :
Belgium
Canada
France
Germany
Japan
Sweden
Switzerland
United Kingdom
; : Disposal; System
.,; ,:; ^e^rmance; Requirements
General only
General only
General only
General for total system
Not yet established
General for total system
Total system objectives
General only
Status of Regulations ;
Details to be developed
Under development
Under development
Regulations complete
Not yet established
Regulations complete
Regulations complete
Deferred for disposal
system
; Approach to Proving .Safety :?
Deterministic & stochastic
Deterministic & stochastic
Deterministic
Deterministic, conservative
Stochastic
Conservative, deterministic,
some stochastic
Conservative, deterministic
Conservative, deterministic
& stochastic, time-
dependent simulation
modeling
Outside the United States, Germany has the most advanced program. A candidate site for a
high-level waste disposal system has been chosen at Gorleben hi Lower Saxony. The
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disposal system will be situated in a salt dome at a depth of about 800 m. Safety must be
demonstrated for 10,000 years, and the maximum allowable dose to the most exposed
member of the general public is limited to 30 mrem/y for unavoidable occurrences before
and after closure. The assessments to demonstrate compliance are deterministic and are
characterized by PNL as conservative and bounding. Conservative is taken to mean that
errors would be on the side of protectiveness of people and the environment.
8.4.4.1 Organization for Economic Co-operation and Development (OECD) Nuclear
Energy Agency
Several studies have been conducted in the last few years at the international level on
assessing the performance of high-level radioactive waste repositories and treating uncertainty
in performance assessment. For example, the status of performance assessment methodology
development was reviewed at a Symposium on Safety Assessment of Radioactive Waste
Repositories, convened by the OECD Nuclear Energy Agency (NEA) in Paris in October
1989. NEA also held a 1987 symposium on "Uncertainty Analysis for Performance
Assessment of Radioactive Waste Disposal Systems" (Seattle, February 1987), and published
a methodology for scenario development ("Systematic Approaches to Scenario
Development") in 1992. The various NEA working groups continue to focus much attention
on uncertainty of long-term modeling of disposal system behavior.
8.4.4.2 Canada
The Atomic Energy Control Board (AECB) issued a regulatory policy statement concerning
long-term aspects of radioactive waste disposal hi 1987, which provides, inter alia, that
individual risk from a waste disposal facility must not exceed 10"6 fatal cancers and serious
genetic effects per year (AEC87). The policy statement includes guidance on how to account
for the probabilities of various exposure scenarios when applying the basic regulatory
requirements, stating that such probabilities "should be assigned numerical values either on
the basis of relative frequency of occurrence or through best estimates and engineering
judgements." Specifically, it indicates that low-probability exposure scenarios should be
assigned values through best estimates and engineering judgments, and that:
the assignment should be made using quantitative analytical techniques to assess as
broad a base of expert opinion as reasonably possible. The use of subjective
probability is appropriate as long as the quantitative values assigned through best
estimates and engineering judgements are consistent with the quantitative values of the
actual relative frequencies in situations where more information is available. The
uncertainty of the probability assigned should also be estimated.
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Furthermore, the AECB policy statement indicates that calculations of individual risks should
be made by either of the folio whig methods:
• using deterministic pathway analysis to calculate annual individual dose, and
applying a risk conversion factor of 2 x 10~2 per sievert; or
• using probabilistic analysis to determine a distribution of annual individual
doses, calculating the arithmetic mean value of these doses, and applying a
risk conversion factor of 2 x 10~2 per sievert.
In the case of deterministic analysis, the AECB urges that analysts not be excessively
conservative, but instead make a balanced choice of assumptions to ensure that the
assessment describes reasonable situations covering the spectrum of exposure pathways and
assesses their impacts rationally. In either the deterministic or probabilistic approach, the
AECB indicates that sensitivity analyses should be conducted to investigate the effect of
changes in input assumptions and model parameters on the magnitude of the single dose
estimate (deterministic) or mean value of dose (probabilistic).
Since the arithmetic mean of a dose distribution could potentially hide the significance of
values at the high end of the distribution, the AECB states that "it is judged acceptable to
allow 5% of the estimated doses to exceed a dose of 1 mSv (100 mrem) per year to take
account of normal statistical variations which are inherent in the probabilistic assessment
process," and that the general risk requirement of 10"6 fatal cancers per year (which
corresponds to a dose of 0.05 mSv/year (5 mrem/year)) takes account of this since a 5%
chance of a dose of 1 mSv (100 mrem) corresponds to an average dose of 0.05 mSv (5
mrem).
8.4.4.3 France
In June 1991, France's Directorate for the Safety of Nuclear Installations (DSIN) issued
Fundamental Safety Rule No. III.2.f on high-level and alpha waste disposal (FRE91).
According to a summary presented by Raimbault et al. (RAI92), the rule provides that
demonstration of safety be based on "deterministic evaluations of the radiological impact for
two types of envisaged situations:
• a reference situation which corresponds to the occurrence of very probable or
certain events.
• hypothetical situations corresponding to occurrence of low probability events
that may lead to preferential transfers."
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Under the reference situation, individual doses should be <_ 0.25 mSv/year (25 mrem/year)
for long exposures associated with very probable or certain events.
The DSIN rule details procedures for conducting performance assessment. The rule requires
validation of numerical models; sensitivity analysis with respect to scenarios, models,
phenomena, and parameters; and results expressed in terms of radiation exposures, with
associated uncertainty bounds.
Based on the DSIN guidance, the Agence National pour la Gestion des Dechets Radioatifs
(ANDRA) has developed an iterative procedure consisting of the following principal steps in
the safety evaluation:
• identify all applicable radiological criteria
• select scenarios to be considered, with detailed radiological impact analysis
and associated sensitivity and uncertainty analysis
• develop a first generation of global safety models corresponding to the selected
scenarios
• develop specific detailed models to consider, e.g., glass corrosion
• test and utilize sensitivity and uncertainty analysis techniques
• develop information on consequence uncertainties and identify the most
important scenarios, phenomena, geosphere data, or concept parameters
• draw conclusions on concept constraints, and define the most important site
data acquisition work.
Input from architectural engineers and site investigators will be provided to the safety
assessment team, and vice versa. According to Raimbault et al., at the end of ANDRA's
planned underground laboratory phase, this procedure should facilitate "a complete and
validated compliance assessment of the disposal system with maximum confidence" and "an
optimized and justified design and program of operation for the underground facility"
(RAI92).
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8.4.4.4 Sweden.
The Swedish Nuclear Power Inspectorate (SKI) has indicated that it has not established a
specific policy on how uncertainty should be treated but instead has analyzed the question on
a case-by-case basis (AND93). However, SKI indicated that it will need to reach more
definite conclusions about how to treat uncertainty at the time of licensing. SKI considers
over-reliance on probabilistic assessments to be inappropriate. The Swedish agency wants all
countries to agree on performance assessment and uncertainty analysis methodologies. SKI's
principal focus in reducing uncertainty is to gain as great an understanding as possible of the
physics and chemistry involved in the disposal system and to conduct research in key areas
of remaining uncertainty.
The Swedish Nuclear Fuel and Waste Management Co. (SKB) addressed some of these
issues, including ambitious probabilistic assessments of ground-water flow in a 1991 report.
The report (SKB91) used stochastic modeling for hydrology, bringing in uncertainties in
conductivities and other parameters and treating them statistically. However, other types of
uncertainties are not treated statistically, but rather by variation analysis and special runs of
the statistical model, e.g. with respect to fracture zones. SKB does not want to choose a
performance assessment method now but recognizes that when the tune comes for evaluating
an actual site, the company will have to choose an approach and defend its selection
(PAP93).
8.4.4.5 Switzerland
The Swiss, along with the Swedes and Finns, are skeptical of fully probabilistic treatment of
uncertainty. Although they do use probabilistic codes, which accept distributions of values
for various parameters, they do not assign probabilities to different values and instead
randomly select numbers assuming a flat distribution. They emphasize that they do not
necessarily try to predict disposal system performance as close to the truth as possible, but
rather to predict that repositories are safe; tools that are known to be wrong but that are
known to over-predict risks are considered acceptable (McC93).
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A paper presented by Zuidema of Switzerland's National Cooperative for the Storage of
Radioactive Waste (NAGRA) at a 1991 NBA workshop on criteria for HLW disposal
identified four sources of uncertainty in disposal system safety analysis:
• uncertainty in scenarios;
• uncertainty in conceptual models;
• parameter uncertainty; and
• parameter variability.
Zuidema indicated that performance assessment tools can be used to quantify different types
of uncertainty: for scenario uncertainty, several alternative future evolutions must be
analyzed; for conceptual uncertainties, several alternative conceptualizations must be
considered; and for parameter uncertainty, both probabilistic and deterministic models can be
used.
Zuidema also identifies disposal system siting and design considerations that are important in
reducing uncertainty. Factors to consider are the ease of exploring the geological
environment and predicting the evolution of the geological environment over time; and the
robustness of the disposal system, including simplicity of physical and chemical properties,
availability of large safety margins, and maximum redundancy of barriers.
8.5 CONCLUSIONS
8.5.1 Essential Role of Peer Review
The criteria for evaluating compliance of the WIPP disposal site with the requirements of 40
CFR part 191 include both a qualitative evaluation of the performance assessment and a
statistical approach based on evaluation of the multiple CCDFs generated by the performance
assessment model using the random-sampling procedure. Each portion of the determination
of compliance must evaluate the degree of uncertainty in the results of the performance
assessment process. Uncertainties that must be addressed include the selection of a
conceptual model for evaluating the likelihood of releases to the accessible environment over
10,000 years, the selection of specific scenarios for evaluation, and uncertainties in the
assignment of numerical values for the probabilities of each scenario considered and the
physical parameters required by the models.
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Although the uncertainty surrounding the selection of values for the physical parameters in
the models is significant, it is equally likely that the subjective uncertainties surrounding the
selection of conceptual models and scenarios and their probabilities are also a source of
uncertainty in the final estimates of cumulative releases generated by the performance
assessment model. These subjective sources of uncertainty are not addressed as completely
in the random-sampling procedures as are the physical parameters of the model. Hence, a
determination based solely on statistical analysis of the performance assessment model LHS
results would not address all uncertainties. Peer review of the conceptual models is essential
in assessing compliance. However, statistical methods also may be required to evaluate the
uncertainties addressed by the LHS procedures.
8.5.2 Selection of a Statistical Criterion for Compliance
Statistical procedures for evaluating compliance with 40 CFR part 191 are expected to add
few costs to the WIPP performance assessment program. Compared to the existing software,
the additional resources to develop software to perform the calculations required for
evaluating compliance using any of the alternate statistical compliance criteria will be minor.
The use of the upper 90% or 95% confidence limit for the sample mean as a test criterion
was introduced in Section 8.3.1. Use of the upper bound of the 95% confidence interval for
the sample mean has been suggested (EPA89, NRC92) as an appropriate method for
determining compliance with soil clean-up standards at decommissioned sites. The
hypothesis test for this alternative is defined by the inequality
UCL = M + k SE(M) < L
where the symbol UCL denotes the upper limit for the confidence interval for the sample
mean (M), and the symbol SE(M) denotes the standard error of the sample mean, and L is
the appropriate regulatory limit specified in §191.13(a). The multiplier k may be selected
appropriately to provide a 90% or 95% confidence level by reference to standard statistical
tables on the t-distribution. If the upper confidence limit for the mean is less than the value
required limit, a determination of compliance would be made.
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The advantages and disadvantages of the upper confidence limit for the sample mean were
summarized in row 1 of Table 8-3. The advantages of this criterion are repeated here:
• The mean CCDF yields the true expected value.
• Use of the upper bound for the confidence interval is a standard statistical
method.
• Use of the standard error of the mean reflects uncertainty in the estimate of the
mean.
• The multiplier k can be adjusted to obtain the desired level of confidence.
• The mean is easy to calculate, even for large samples.
For distributions on positive variables which are skewed toward higher values, the mean will
often lie above the median of the distribution. Use of the mean in this case is comparable to
using a higher percentile than the median for determrning compliance. For highly skewed
distributions, the mean may exceed the 90th percentile of the distribution. This results in a
test which is protective of the environment by increasing the likelihood of a non-compliance
test result. Furthermore, the upper confidence limit for the mean will exceed the mean
itself, thus adding to the conservativeness of the test procedure. For small sample sizes, the
standard error of the mean will be larger, thus making the test even more protective.
The low robustness of the test also results hi a more protective test for compliance. If very
high "outliers" occur in the sample values, the estimated value for the mean will be very
sensitive to these high sample values. The estimated value for the standard error of the mean
may also be inflated by outliers. The increases in the estimate of the mean and its standard
error due to outliers tend to make the test more protective. Larger sample sizes would
reduce the impact of uncertainty due to the t-distribution assumption for selecting the
multiplier k. For smaller sample sizes, the value of k indicated by the t-distribution may
provide only an approximation for the proper value of the multiplier to obtain a true 90% or
95 % confidence interval for the mean.
It appears that the relatively lower robustness of the probability-weighted mean relative to the
median is a fundamental characteristic of the uncertainties of performance assessment rather
than a flaw of that statistical method. Outliers which lower the robustness of the statistical
test represent valid possibilities for large releases from the disposal system. These outliers
are the mathematical representation of the recurring theme that "Proof of the future
performance of a disposal system is not to be had in the ordinary sense of the word..."
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In 40 CFR part 194, EPA decided that the statistical portion of the determination of
compliance with 40 CFR part 191 will be based on the sample mean. The LHS sample sizes
should be demonstrated operationally (approximately 300 when 50 variables are considered)
to improve (reduce the size of) the confidence interval for the estimated mean. The
underlying principle is to show convergence of the mean.
8.6 REFERENCES
AEC87 Atomic Energy Control Board, Regulatory Policy Statement, "Regulatory
Objectives, Requirements and Guidelines for the Disposal of Radioactive
Wastes - Long-Term Aspects," Regulatory Document R-104, June 5, 1987.
AND93 Andersson, J., Swedish Nuclear Power Inspectorate, personal correspondence,
August 12, 1993.
DOE93 DOE Comments on Advance Notice of Proposed Rulemaking for the
Certification of Compliance with 40 CFR Part 191 Disposal Regulations, in
Letter to Michael H. Shapiro (EPA) from Paul D. Grimm (DOE), ATTN:
Docket A-92-56, March 31, 1993.
EPA89 U.S. Environmental Protection Agency, "Methods for Evaluating the
Attainment of Clean-up Standards", Vol. 1, Soils and Soil Media, EPA
230/02-89-042, February 1989.
EPA92 U.S. Environmental Protection Agency, "No Migration" Variances to the
Hazardous Waste Land Disposal Prohibitions: A Guidance Manual for
Petitioners. EPA530-R-92-023, Office of Solid Waste, Washington, DC, July
1992.
ESL92 Letter to John Davidson (EPA) from P.W. Eslinger (Battelle-PNL),
February 11, 1992.
FRE91 Direction de la Surete des Installations Nucleaires, Ministere de 1'Industrie et
du Commerce Exterieur, Fundamental Safety Rule No. III.2.f, "Definition of
objectives to be met during the study phase, and of the work for final storage
of radioactive waste in deep geologic formations, in order to assure safety
after the period of using the facility," June 10, 1991.
GLI78 N. Glick, "Breaking records and breaking boards," American Mathematical
Monthly, Vol. 85, 1978, pp. 2-26.
HEL93a J.C. Helton, and R.J. Breeding, "Calculation of reactor accident safety goals,"
Reliability Engineering and System Safety, Vol. 39, 1993, pp. 129-158.
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HEL93b J.C. Helton, "Drilling intrusion probabilities for use in performance
assessment for radioactive waste disposal," Reliability Engineering and System
Safety, Vol. 40, 1993, pp. 259-275.
McC93 McCombie, C., "National Cooperative for the Storage of Radioactive Waste
(NAGRA)", personal correspondence, August 13, 1993.
NRC92 Manual for Conducting Radiological Surveys in Support of License
Termination, NUREG/CR-5849, draft report for comment, June 1992.
NRC93a Letter to J. William Gunter (EPA) from B.J. Youngblood (NRC), March 22,
1993.
PAP93 Papp, T., Swedish Nuclear Fuel and Waste Management Co. (SKB), personal
correspondence, August 16, 1993.
PNL90 Pacific Northwest Laboratory, "Comparison of Selected Foreign Plans and
Practices for Spent Fuel and High-Level Waste Management," PNL-7293,
April 1990.
RAI92 Raimbault, P. et al., Agence Nationale pour la Gestion des Dechets
Radioactifs (ANDRA), "Methodology Developed by the French National
Nuclear Waste Management Agency (ANDRA) for the Performance
Assessment of a Deep Geological Repository," proceedings of the Third
International High Level Radioactive Waste Management Conference, Las
Vegas, Nevada, April 1992, pp. 510-516.
SKB91 SKB91: Final Disposal of Spent Nuclear Fuel, Importance of the Bedrock for
Safety. SKB Technical Report 92-20, May 1992.
SNL85 Sandia National Laboratories, "A Comparison of Uncertainty and Sensitivity
Analysis Techniques for Computer Models", SAND84-1461, R.L. Iman and
J.C. Helton, March 1985
SNL92 Preliminary Performance Assessment for the Waste Isolation Pilot Plant,
SAND92-0700, Sandia National Laboratories, December 1992.
SNL93 Sandia National Laboratories, "Conceptual Structure of Performance
Assessments for the Waste Isolation Pilot Plant", SAND92-2285, J.C. Helton,
et al., April 1993.
WU93 Wu, Y.T., A.B. Gureghian, B. Sagar, and R. Codell, 1993, "Sensitivity and
Uncertainty Analyses Applied to One-Dimensional Radionuclide Transport in a
Layered Fractured Rock. Part II: Probabilistic Methods based on the Limit
State Approach", Nuclear Technology, Vol 104, pp 297-308 (Nov. 1993)
8-42
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9. Consideration of Human Intrusion
9.1 INTRODUCTION
The containment requirements (§191.13) of 40 CFR part 191 specify that waste disposal
systems must be capable of constraining movement of waste to the accessible environment
for 10,000 years. To demonstrate this capability, DOE must show that there is a "reasonable
expectation" of not exceeding specified release limits "from all significant events and
processes that may affect the disposal system." Significant events and processes include
those that are both natural and human-initiated. The final rule, 40 CFR part 194, includes
specific requirements on human intrusion. These criteria are based on the assumption that
inadvertent and intermittent drilling for resources is the most severe scenario to be
considered when addressing human intrusion for performance assessment calculations because
it provides a direct intersection with the waste and a pathway to the surface. Mining of
resources is a very important, though less direct form of a human-initiated process or event.
Under the provisions of the WIPP LWA, no surface or sub-surface mining or oil or gas
production, including slant drilling from outside the boundaries of the sixteen square mile
withdrawn area, is permitted with one exception. Directional (slant) drilling is permitted
from outside the land withdrawal boundary into oil and gas leases in the extreme
southwestern corner of the WIPP site, but only at depths below 6,000 feet. This depth is
well below the repository horizon at 2,150 feet. EPA can make a determination after
consulting with DOE and the Secretary of the Interior that DOE should acquire these leases
to assure compliance with the disposal regulations.
The 40 CFR part 194 rule defines two types of human intrusion which must be considered in
addition to mining:
• deep drilling events that reach or penetrate the level of waste in the disposal
system
• shallow drilling events that do not reach the level of waste in the disposal
system.
A variety of drilling events can be envisioned as occurring in the vicinity of the WIPP site.
These include exploration and development drilling for oil and gas, exploration drilling for
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potash, drilling of water wells, and exploration drilling for other minerals. For example,
water well drilling around WIPP is currently limited to geologic strata which lie substantially
above the WIPP repository horizon—this would be defined as shallow drilling. Even though
the wells are located in strata above the repository, if those strata became contaminated with
radioactivity from the repository, water extraction could accelerate movement of
radionuclides laterally to the boundary of the accessible environment or could directly
transport contaminated water to the land surface which is also defined as part of the
accessible environment.
This chapter provides background information related to possible human intrusion by drilling
and by underground mining. The regional geology relevant to understanding human
intrusion issues is described in more detail hi the following section. Subsequent sections
discuss specific drilling and mining issues.
9.2 GEOLOGIC SETTING
Understanding the regional geology is an essential step in developing a defensible description
of human intrusion which might impact the WIPP repository. The WIPP is located in the
northern part of the Delaware Basin, which is a large sedimentary basin in Southeastern New
Mexico and West Texas. This deep oval-shaped structural depression, which is about 135
miles long and 75 miles wide (COO71), was an embayment covered by a deep sea (i.e., 305
to 245 million years before present [Ma]). Sedimentation within the basin resulted in
formation of thick marine strata. Organic activity at the margins of the basin produced
carbonate reefs — the present day Capitan Reef — that separated the deep-water sediments
from the shallow-water shelf deposits which developed landward from the reefs (SAN92).
The depositional process, as described by Cooper and Glanzman (COO71), is summarized as
follows:
The irregular floor of the sea was characterized by structural basins, platforms
and broad shelves. Fine sand and limestone accumulated in the basins; reefs
formed on the margins of the shelves and platforms; limestone and sand
accumulated immediately behind the landward side of the reefs; and gypsum,
anhydrite, and other evaporite rocks, and silt and clay accumulated in the
shallow waters of the shelves. Eventually, the reef growth was halted by
increasing the salinity of the sea water and evaporite sediments (Castile,
Salado, and Rustler Formations) were deposited in the Delaware Basin.
Evaporite deposition was interrupted during two intervals of time, during
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which the water was less saline and limestone was deposited. Toward the end
of Permian time, deposition of the evaporite rocks ceased and deposition of
terrestrial red beds (Dewey Lake Redbeds) began. Terrestrial deposition
continued during parts of Triassic time. Additional thin deposits of sediments
accumulated in Quaternary time. A total of 18,500 feet of sedimentary rocks
were deposited in places in the Delaware Basin.
The region surrounding the WIPP may be characterized in terms of the geologic composition
of the sub-surface features, the hydrologic properties and the history of hydrocarbon (i.e. oil
and natural gas) accumulation. The horse-shoe shaped Capitan Reef was formed by the
deposition of organic material which differs from the non-organic material which developed
into the evaporite salt formations which lie along the interior. King noted this distinction
when concluding that the Capitan Reef was separate from the Delaware Basin.
The rocks of the Guadalupe Mountain region were deposited near the edge of
a feature of Permian tune known as the Delaware Basin, along whose margin
they show complex changes in facies. The rocks laid down outside the basin,
in what is here termed the shelf area, are thus very different from
contemporaneous basin deposits. (KIN42)
The "complex changes in facies" referred to by King result from the geologic dissimilarities
of the carbonate Capitan Reef from its contemporary, the evaporite Bell Canyon formation
which adjoins the Capitan Reef on the interior side. (The WIPP is located within the Salado
Formation, which is a sequence of evaporite rocks deposited in the Late Permian Epoch
(258-245 Ma). A portion of the stratigraphic column that represents this depositional
sequence is shown in Figure 9-1 (WEI77).) The Guadalupe Mountains or shelf area, noted
by King, in fact contain a portion of the Capitan Reef. Other authors, including Cooper and
Glanzman, quoted earlier, have remarked as well upon this distinction: "This structural
[Delaware] basin is generally considered to be the area surrounded by the Capitan Limestone
[i.e. the Capitan Reef]."
Subsequent to its formation, the hydrologic properties of the Capitan Reef were enhanced by
fracturing and dissolution such that the effective porosity of the Capitan Reef increased.
Partially as a result of this, the Capitan Reef is today a significant aquifer in the region and is
a major source of water for the City of Carlsbad. In contrast, the salt formation which
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Elevation Depth (ft)
3UOO"V y
4000— / \v
-750 ~m '' ' •••'•-[
5000
^uuu — T
6000 — -1. ...'..
•
o i rn ^3-
7000 -^^^ " " " "
^J
8000 -S -•-.•.:.:.-..•.-•
-4750 [ V : ' T '
^
9000 —\ — i 1
•
m nnn
• _
• — ' ' —
•
1 1 ,000 -J .......
Tr.rn. — - -M
t — ' — ' —
f --
12,000 — t::::::::::::
-9100 -^— i 1 —
i T nnn J 1 — ' 1
-9700- «— -i — • 1
-10,200 t.-I.-i.-J...
14 noo-J - -
-10,950 ^.•..•I:'li.'.;.:l
11 100 i • '
15,000 —
l,bUO [
-11 8^^ J ' '
• / /
16,000-3 /
/
/ /
-13,050 -- /
/ /
I7L000 —
-lo,DUU 1
...:::
-13,900 — i. — , — i —
18.000
Ochoan,
Triassic and
Younger
Ramsey Sand
Bell %
Canyon J
TO O
Cherry | §
Canyon | §
0) CNJ
Brushy Q
Canyon Permian
Bone
Spring
(2900-3500)
Wolfcamp
(1000-1500)
Strawn
A+ i Pennsv
At0ka (1800-'
Morrow
— ' Bamett i^/iiooio
— ; : : IV1ISS1S
Mississippian /,-£-n -
Limestone (550- /
— \Woodford
Silurian- Siluria
Devonian (1250
Carbonate
Montoya ^
Orrin
Simoson (10Q(
j Ellenburqer
Measurements
are in feet
I Potential Pay Zones
t;nnn
rlvanian
>100)
sippian
r50)
n/Devonian
-1700)
vician
)-14QO)
Precambrian
Annn
3000
- 2000
1000
500
- 0
Figure 9-1. Stratigraphic Column and Potential Hydrocarbon Pay Zones (Source: POW78)
9-4
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contains the WIPP, known as the Salado, has a small primary porosity combined with a lack
of transmissive fractures, and thus groundwater flow through the Salado is not significant.
The ulterior regions as a whole are noteworthy for the relative scarcity of potable
groundwater.
Over the geologic history of the Delaware Basin, oil and natural gas have accumulated
underneath the Capitan Reef. Although formed in the ulterior portions from organic
material, the hydrocarbons preferentially migrated outward until being trapped by the cap-
like profile of the underside of the Capitan Reef. Organic material which generates
hydrocarbons did, in fact, exist in the ulterior portions and was deposited during the middle
and late Permian era, but as Hills describes (HIL84), the ultimate fate of the hydrocarbons
lay elsewhere: "The hydrocarbons contained in the source beds [in the regions within the
Capitan Reef] migrated primarily to interbedded sandstone reservoirs within the [Delaware]
basin and later to the porous carbonate reservoirs on the margins." This "trapping
mechanism" possessed by the Capitan Reef does not have a parallel in the interior portions,
such as the Salado and Bell Canyon formations. Hills described this difference: "no large
Upper Permian structural traps were formed in the [Delaware] basin, and most hydrocarbons
migrated to the surrounding shelves [i.e. the Capitan Reef]."
Various estimates of the area of the Delaware Basin have been cited by different authors. In
most cases, neither the basis for the estimate nor whether the cited area includes or excludes
the Capitan Reef is mentioned. For example, Hills (HIL84) said the area is about 13,000
square miles (33,500 km2) while Richey et al. mention that the area (which presumably
includes the Capitan Reef) is about 12,000 square miles (31,000 km2). This same area is
cited without attribution by Powers et al. (POW78, v. 1, pp. 3-59). An earlier site selection
report had noted that "the area of the Delaware Basin was assumed to be about 30,000 km2
(CLA74). These differences are not surprising since the boundaries of the Basin mostly lie
below the surface and the estimates were intended for descriptive purposes rather than
quantitative analysis.
To obtain better quantitative values for the Basin area, the area on several maps was
measured using standard software designed for operation with geographic information
systems. The areas of the Delaware Basin as depicted in Figure 1 in HIL84, in Figure 3.4-1
of POW78, and Figure 6.3-8 of POW78 were calculated with results as follows:
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HIL84, Figure 1 - 28,000 km2
• POW78, Figure 3.4-1 - 30,200 km2
POW78, Figure 6.3-8 - 25,200 km2
POW78, Figure 6.3-8 indicates more detailed mapping of the basin boundary and the
surrounding Capitan Reef. Various versions of this figure are widely used in the technical
publications prepared by Sandia National Laboratories for the WIPP project. This
representation of the Delaware Basin embracing 25,200 km2 is reproduced in Figure 2-2 and
is appropriate for estimating intrusion rates.
9.3 INTRUSION BY DRILLING
9.3.1 Oil and Gas Drilling
Drilling for oil and gas has been conducted in the Delaware Basin since the turn of the
century. Over the past decade, drilling for oil and gas in the vicinity of the WIPP site has
increased significantly (SIL94). Typical oil drilling targets within the Delaware Basin around
the WIPP site include Permian age rocks such as the Cherry Canyon and Brushy Canyon
Members of the Delaware Mountain Group and the Bone Springs Formation. The tops of
these geologic formations lie about 5,700 feet and 8,300 feet below the land surface
(GUZ91a). These formations were not generally recognized as exploration and development
targets until the late 1980s because their reservoir production characteristics were not well
understood (NBM95). However, recent improvements in borehole logging procedures have
allowed petroleum geologists to determine which Delaware Mountain Group sediments have
a high potential for fluid hydrocarbons.
Gas drilling targets reside in the Pennsylvanian age Strawn, Atoka, and Morrow formations
at depths of about 12,700 feet, 13,200 feet, and 13,700 feet below the surface, respectively.
All current oil and gas targets are below the WIPP repository horizon and would be defined
as deep drilling under 40 CFR part 194.
Data on oil and gas drilling are available from a variety of sources. In New Mexico, the
U.S. Bureau of Land Management and State of New Mexico Oil Conservation Division keep
records on wells that have been issued permits. In Texas, this information resides with the
Texas Railroad Commission. Borehole information is also obtainable from commercial data
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EXPLANATION
DISSOLUTION WEDGE
EXTENT OF CENOZOIC FILL
DEEP CENOZOIC FILL
O IO 20 »O 4OKM
AREA SUBJECT TO EROSION
DURING JURASSIC TIME
(BACHMAN 1976)
Reference: Based on Figure I
from Anderson, 1978
DISSOLUTION FEATURES AND EXTENT
OF EROSION IN DELAWARE BASIN
Figure 9-2. Outline of the Delaware Basin (Source: POW78, Figure 6.3-8)
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bases such as that managed by the Petroleum Information (PI) Corporation in Denver,
Colorado. The earliest Delaware Basin information in the PI data base indicates that five
holes were drilled between 1909 and 1914. Four of these holes were in Texas and one was
in New Mexico.
9.3.1.1 Permitting Practices
In order to conduct oil and gas drilling, a permit must be obtained either from the U.S.
Department of the Interior, Bureau of Land Management (BLM)—if Federal lands are
involved—or the Oil Conservation Division (OCD) of the State of New Mexico Energy,
Minerals, and Natural Resources Department—if state or private lands are involved. Drilling
activity on Federal lands is regulated under 40 CFR part 3160 (Onshore Oil and Gas
Operations; Federal and Indian Oil and Gas Lease: Drilling Operations: Final Rule,
November 18, 1988). The regulation prescribes minimum levels of performance and
enforcement action when rules are violated. Procedural requirements are not included.
OCD Rules and Regulations are more specific than those of 40 CFR part 3160 and BLM
often follows specific OCD practices. Many detailed procedures—including well casing
practices, well completion, and methods of borehole sealing—requiring BLM or OCD
approval are area specific, not included in regulations, and not always in writing.
OCD Rule 104 specifies the maximum density for oil and gas wells. Wildcat (i.e.,
exploration) gas wells in Eddy and Lea Counties of New Mexico are granted a minimum
spacing of 160 acres for drilling depths of less than 11,000 feet, or 320 acres if drilled to
greater than 11,000 feet. Development (i.e., production) gas wells are also limited to one
per 160 or 320 acre tract.
The minimum spacing for wildcat and development oil wells in these two counties is 40
acres. There is also a limit of four development wells per tract when special pool rules
apply. However, more wells are allowed if the tract is permitted for active secondary
recovery. In each case, oil and gas wells must also conform to boundary offset distances,
placing each well within a specified area within a tract. In the absence of secondary
recovery operations or special pool rules, the spacing requirements thus allow up to 6.2 oil
wells/km2 or up to 1.5 gas wells/km2 for each potentially productive formation. BLM
follows OCD well spacing rules although they are not legally required to do so. BLM also
requires an environmental assessment before a permit is granted.
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9.3.1.2 Well Drilling and Casing
After the permit is granted, the drilling contractor will set up equipment at the drill site. For
deep holes drilled around WIPP, about 7-10 days are typically required to prepare the site,
set up the drill rig, construct the mud pits, and set the surface conductor casing to secure
surface sediments before drilling commences.
The drilling of a gas or oil well usually requires a program involving two or more bit sizes
to complete a borehole and one to four bit changes per borehole (BER94) to change worn or
damaged drill bits. A typical gas or oil well starts with a large diameter hole at the surface
into which a conductor pipe is placed. A smaller size drill bit which can pass through the
conductor is used to drill a hole to accommodate the surface casing, through which a still
smaller bit passes to drill for the production casing. Should bottom hole conditions warrant,
a liner may be inserted in the lower portion of the production casing. The bottom hole is
usually a minimum of 2 to 3 inches in diameter and telescopes outward to the larger
diameters required to accommodate uphole conditions.
The OCD has specified for the past decade or more that all gas and oil wells on New Mexico
state and private lands (with minor exceptions) be drilled in the 171/2, 121/4, and 77/8-inch
diameter size sequence. As each step in the drilling sequence is completed, the drill string is
removed from the borehole and the hole is lined with tubular steel casing which is set in
place with cement. The larger diameter surface casing is set from the surface to the top of
the Rustler Formation at a depth of about 500-600 feet. The next drilling sequence is
initiated which involves penetration of the salt section (the Salado and Castile formations).
When the hole reaches the bottom of the Castile at a depth of about 4,000 feet, the drill
string is again removed from the borehole and the intermediate casing is set. After the
intermediate casing is set, drilling is reinitiated and continues until the target horizon, for
example, the Cherry Canyon Formation of Delaware Mountain Group, is reached and the
production casing is then set. This size sequence appears also to be the current common
practice for drilling on Federal lands administered by the BLM.
OCD records indicate that a two casing program was used during the 1970s and earlier, in
which smaller bits and casings were common. A frequent practice in a two-casing run was
to combine the surface and intermediate casing intervals into one extending from the surface
to a depth of 2,000 to 4,000 feet above the production zone before a smaller bit was used.
9-9
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A drilling sequence of 103/4 or 121/4-inch bit followed by a 77/8-inch bit was most common for
this practice, though no count has been made as to the number of such wells. Gas wells
have been completed at depths ranging from 10,000 to 16,000 feet deep and oil wells have
been completed at 5,200 to 8,200 feet deep. Drilling and completion of the well typically
take about 100 days. Virtually all gas and oil wells in the area are drilled to depths which
would penetrate the WIPP horizon.
OCD Rule 107 imposes a number of requirements concerning casing, tubing, cementing,
etc., for wells being drilled and completed. Section III8 of 43 CFR part 3160 establishes
similar requirements for Federal lands. However, in all cases the detailed procedures for
specific locations are not in the regulations and are specified by the appropriate District
Offices. OCD states that they inspect 100% of new wells being drilled, cased, and
completed. BLM inspects wells under its jurisdiction on a random, less than 100%, basis.
If the WIPP site were to be penetrated by inadvertent human intrusion, such an event would
occur during drilling through the salt section before the intermediate casing is set. Once the
intermediate casing is set in place and the annulus between the casing and the borehole wall
is sealed with cement, then the possibility of radionuclide contamination reaching the surface
will be prevented as long as the casing remains intact. Typically, casing integrity is
demonstrated by pressure testing and ultrasonic logging of the cemented section for bonding
between the casing and the cement and between the cement and the formation. It is
estimated that this critical section of a borehole would remain uncased for no more than three
days during drilling. While there is no solid data base which describes the frequency of
occurrence of improperly cased holes or holes with casing failures, such failures have been
reported (KIR94).
9.3.1.3 Detection of the Repository During Drilling
A key question when developing the possible range of human intrusion rates to which the
WIPP repository might be subjected, is whether the drilling contractor is likely to detect the
presence of the WIPP repository during drilling for oil or gas. The drilling operator could
penetrate the WIPP and not realize he had done so. Assuming his drill hole was successful
and the anticipated oil pool or gas reservoir was reached, he might drill additional
development wells to exploit the resource. These additional holes might also penetrate the
waste and be undetected.
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Because of the nature of drilling practices employed in the Delaware Basin, the existence of
the WIPP may not be revealed by a borehole which intersects the repository. Drilling is
typically done by independent drilling contractors whose main goal is to "make hole" and
efficiently meet the contract requirements. Drilling through the salt section is typically at a
rate of 50 to 100 feet per hour, so only about 8 to 15 minutes would be required to penetrate
the repository (based on room height prior to creep collapse of the repository). J. W.
Berglund of the New Mexico Engineering Research Institute states, "While in the salt
section, drilling mud (brine) is supplied from a large, plastic lined reserve pit dug hi the
ground with a surface area of about 4,000 ft2. Drilling mud is pumped from the reserve pit
down through the drill pipe and drill bit and up the annulus formed by the drill string and
drilled hole. The drilling mud and the drill cuttings are returned directly to the reserve pit
where the cuttings settle out. While drilling hi the salt section, no formal attempt is made to
monitor the character of the cuttings or the fluid volume of the reserve pit. A gas analyzer
is not attached to the returns until the hole is much deeper than the depth of the WIPP
repository" (BER95).
Even if gas flows were generated when the drill bit intersected the WIPP, this event would
more likely be interpreted as a release from naturally occurring gas pockets. If the drill bit
encountered brine from the WIPP, this too might be interpreted as a naturally occurring
phenomenon. Naturally occurring gas and brine pockets are commonly found hi the
sedimentary rocks above and below the WIPP horizon hi the Delaware Basin. Brine pockets
are encountered during drilling of deep boreholes into the Castile Formation below the
repository, and to a lesser extent into the Salado Formation above the WIPP horizon. For
example, records available at the OCD office hi Artesia, NM, reveal that a brine flow
blowout (back pressure hi the drill stream sufficient to cause actuation of over-pressure relief
valves to protect piping from damage) occurred for a recently drilled well (API no. 30-015-
27406) (S2, T18S, R30E) at a depth of 898 feet below the surface which is a few feet into
the McNutt potash zone of the Salado Formation. Drilling was temporarily halted for four
hours during which the flow rate was estimated at 40 to 50 barrels per minute (1,680 to
2,100gpm). The hole was shut hi and allowed to pressurize to 350 psi. Drilling of the
121/4-inch diameter hole resumed to a depth of 1,555 feet. Casing was set and cemented. As
the brine inflow was similar hi characteristics to the brine water drilling fluid being used, no
apparent effect was seen except for the break hi drilling.
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Gas kicks (blowouts) also occur in the Salado Formation in which the WIPP is located.
When these are encountered, drillers routinely let the drilling mud blow from the hole—this
appears as a cut brine solution when drilling through the salt section—with no effort to retard
the blowouts by closing blow-out preventers on the drill rigs. These kicks are of short
duration and when completed, drillers restart mud circulation and adjust the drilling mud to
the desired weight. Presence of these small pockets of water or nitrogen has been detected
both by mining and drilling (COO71). The largest cavity found through 1971 was about 176
m3 and was not pressurized—at least when discovered by mining (CLA74). There are at
least seven reported incidents of pressurized gas blowouts in potash mines. Two of these
resulted in fatalities (CHA84).
Brine inflows from Castile Formation brine reservoirs are well documented for two wells—
ERDA-6 (S35, T21S, R32E) and WIPP-12 (S17, T22S, R32E)—in which the initial inflow
was 20 gallons per minute (gpm) and 350 gpm, respectively. The brine reservoirs were
estimated at 26.5 million gallons total for ERDA-6, with 69,000 gallons flowing to the
surface during testing, and 714 million gallons total for WIPP-12, with 3.3 million gallons to
the surface (POP83). Although historic information is vague and incomplete, similar-sized
brine reservoirs were observed in: 1) Mascho-1 (S20, T22S, R33E) drilled in 1937 which
had a reported initial flow of 230 gpm, 2) Belco (S25, T23S, R30E) drilled in 1974 with an
initial inflow of 350 gpm while flowing for 26 hours, and 3) Shell (S36, T22S, R32E) drilled
in 1964 with an initial inflow of 580 gpm. Brief commentary on these wells (POP83)
describes stopping of drilling operations until artesian flow is completed, followed by a
resumption hi drilling. Recent interviews with drillers substantiated this practice. Brine
pocket blowouts are like gas kicks, in that they cause no problems beyond drilling breaks.
In one well, Pogo (S26, T21S, R31E), a moderate weight drilling mud (15 pounds per gallon
ppg) was applied after four days of flow. Whether the flow was really stopped by this
weight of mud or whether the reservoir pressure was exhausted is unknown. A similar
weighted mud of 12 ppg did not stop the inflow at another well drilled hi 1962.
9.3.1.4 Borehole Plugging and Abandonment
When a borehole is no longer useful it must be plugged and abandoned according to specified
procedures. One mechanism identified for releasing wastes from the WIPP repository is the
escape of waste-generated gas or contaminated brine through an unsealed or improperly
sealed borehole. In some scenarios used to assess the performance of the WIPP, boreholes,
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plugs, or seals are assumed to remain intact for the full regulatory period, and in some cases
the seals are assumed to degrade. Effectiveness of borehole seals is important to maintaining
the integrity of the WIPP repository. Borehole permeability was identified in the 1992 WIPP
PA as one of two critically important parameters (SAN92).
The New Mexico OCD Rules and Regulations (OCD93) pertaining to sealing off geologic
strata and notification are covered in Rule 106, which requires protection of oil- and
gas-producing strata from each other and from overlying water strata. Also, "All fresh
waters and waters of present or probable value for domestic, commercial or stock purposes
shall be confined to then: respective strata and shall be adequately protected by methods
approved by the Division." OCD Rule 202 (Plugging and Permanent Abandonment) requires
prevention of the contamination of fresh waters. The Rules and Regulations define fresh
water as less than 10,000 mg/1 total dissolved solids (TDS). An underground source of
drinking water is defined as an aquifer having less than 10,000 mg/1 TDS and containing a
sufficient quantity of water to supply a public water system, unless it has been exempted.
The Federal regulations also require protection of usable water which means generally those
waters containing up to 10,000 ppm of total dissolved solids.
OCD Rule 101 requires a surety bond, payable to the State of New Mexico, before new
wells can be drilled. The purpose of this bond is to pay for the proper plugging, sealing,
and abandonment of the well if the owner is financially unable to do so hi the future. In
Eddy and Lea County, the amount of this bond varies from $5,000 to $10,000 per well,
depending on well depth. Alternatively, a blanket plugging bond of $50,000 can be obtained
to cover all wells drilled by an operator. The BLM also has bonding requirements for new
wells, although the adequacy of this program has been questioned by the Department's
Inspector General (DOI92). Large numbers of older wells on both Federal and non-Federal
lands do not have adequate plugging bonds, and the State and Federal Government's financial
obligation to seal and abandon these wells properly may be significant.
OCD Rule 201 requires a well be either properly plugged and abandoned or temporarily
abandoned within 90 days after: (1) a 60-day period following suspension of drilling
operations, (2) a determination that a well is no longer usable for beneficial purpose, or (3) a
period of one year in which a well has been continuously inactive. Despite these
requirements, the current status of a large number of wells under OCD regulations is
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unknown (OCD94). Audits by DOI Office of Inspector General reveal a significant unknown
well status problem also exists on BLM lands (DOI89, DOI92).
OCD Rule 202 requires written notice on Form C-103 at least 24 hours in advance of
commencing any plugging operations. Verbal approval of the method of plugging and time
to begin is permissible for a newly drilled dry hole. The well operator is required to notify
the OCD after plugging and site clean-up operations for an inspection of the well and
location. However, the operator has up to one year after completion of plugging operations
to complete the site clean-up. The Federal Abandonment Requirements, contained in 43
CFR part 3160 III G, specify details of cementing, plugging, and capping boreholes, but do
not identify any procedural requirements. Field interviews suggest that 100% of all holes
under OCD aegis are inspected, and less than 100% of holes drilled under BLM aegis are
inspected for compliance with plugging regulations.
OCD Rule 203 describes conditions under which a well may be temporarily (rather than
permanently) abandoned. Temporary abandonment can be for a period of up to 5 years and
the operator can apply for renewal at the end of this period. In seeking renewal, the
operator is required to test the integrity of the casing with a specified procedure and provide
evidence that there will not be migration of water or hydrocarbons between strata.
BLM proposed procedures for reviewing the status of non-producing wells following findings
hi a 1989 Inspector General Audit Report (DOI89) that large numbers of wells had been
inactive for years without meeting BLM's procedures or requirements for temporary
abandonment. The 1992 follow up audit report did not consider implementation of these
procedures to be adequate. After the 1989 audit, BLM had proposed procedures for
reviewing the status of non-producing wells. These proposed procedures would require that
BLM field offices review the status of non-producing wells listed monthly and determine
whether each well was usable for further oil and gas production. The procedures would also
require that the field offices request the operators to either submit a justification for shut-in
status, obtain temporary abandonment approval, plug and abandon the well, or resume
production. If implemented, the Inspector General felt that its 1989 recommendation would
be satisfied (DOI92). However, the Inspector General advised the Secretary of the Interior
on March 20, 1992 that BLM did not agree with the Inspector General's recommendation to
devote resources and management oversight to improve the Inspection and Enforcement
Program.
9-14
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Gas and oil wells are usually plugged in two ways. Plugs are either placed inside the
production casing or inside the intermediate casing when the upper portion of the production
casing is removed. In either situation, plugging is performed within a cased hole. As noted
previously, in gas and oil wells in the Delaware Basin, surface casing extends from the
surface to the bottom of the Rustler Formation and is cemented in place, thereby rendering it
permanently fixed. The intermediate casing which is placed inside the surface casing extends
from the surface through the salt section and terminates at the bottom of the Castile
Formation (a depth of approximately 3,600 feet at the WIPP). This intermediate casing is
cemented from bottom to surface and is likewise permanently fixed in place. The production
casing which extends from the surface to the Delaware Mountain Group strata for oil or
deeper to the Morrow or Strawn Formations for gas may either be cemented inside the
intermediate casing from bottom at 6,000 to 8,000 feet deep to the surface, or it may only be
cemented for the lower 3,900 feet, thereby allowing the removal of several thousand feet of
the inner casing string when the well is abandoned. At least one, and sometimes two,
cement-shrouded casing strings separate the WIPP horizon rock from the open hole. A four-
part casing is unlikely in the vicinity of the WIPP since four casing strings extending from
the surface are required by the OCD only over the Capitan Reef. For deep gas wells
incorporating a liner, the fourth inner string is hung from the lower portion of the production
casing and does not extend into the intermediate casing.
All downhole tools and fluids are removed from a typical gas and oil well prior to
abandonment. A class "C" type cement plug is placed at intervals throughout the hole
starting at the guide shoe of the inner casing string (usually the production casing) which has
not been drilled out but was previously cemented hi place when the inner casing string was
set. The inner casing is usually set to the bottom or below the producing zone and
perforations are made in the casing for a length of up to 100 feet above the guide shoe.
Plugs are placed at the top of each producing formation followed by intermediate plugs,
plugs at the bottom and top of the salt section, and a surface plug. Each plug is about
35 feet thick and is placed at intervals no greater than 2,000 feet as specified by OCD.
Drilling fluid is placed between each plug.
The position of each plug is carefully monitored because the plug can slip before it sets.
Cement plugs are more dense than the fluid upon which they rest and can possibly move or
disintegrate into the fluid before hardening. An OCD field representative certifies the
placement of each plug for holes on state and private lands before the next interval of fluid is
placed.
9-15
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Field investigations failed to uncover any long-term monitoring of borehole seal integrity.
Neither the BLM nor OCD conducts follow-up studies after borehole seals have been
installed. The American Petroleum Institute, recognizing that "an unknown but large number
of abandoned wells exist that are unplugged or inadequately plugged by today's standards,"
conducted an analysis of whether abandoned wells could act as conduits to move saline water
from deep underground to more shallow-lying underground drinking water sources
(WAR90). The studies focused on brine flow from the Lower Tuscaloosa Sand, a mature
oil-producing trend in Mississippi and Louisiana at a depth of about 10,500 feet, to the
Sparta sands and shales, the drinking water source which bottoms at about 3,100 feet. A
nearby injection well in the Lower Tuscaloosa was assumed to provide the driving force for
flow in the abandoned well. Two scenarios were examined, viz.:
• Uncased well scenario — The upper 1,500 feet of the well are cased and
cemented, but the balance of the borehole remains open. In time, overlying
shale sloughs into the hole to form a 154.5 foot column of shale with a
porosity of 3% and a permeability of 0.0001 Darcy. Above this is a 4,620
foot column of settled solids from the drilling mud having a porosity of 84 %
and a permeability of 0.001 Darcy.
• Cased well scenario — The well is cased from top to bottom and the lower
2,000 foot production casing is cemented. The annulus between the balance of
the casing and the borehole is left filled with drilling mud. It is assumed that
a corroded interval develops hi the casing at a depth of 6,000 feet.
The two scenarios modeled using the SWIFT III computer code indicated that—over the
range of injection rates considered (20 to 600 barrels per day)—there was no flow into the
underground drinking water source. Thus, for the conditions examined, unplugged or poorly
plugged boreholes were not a problem. One should also note that the permeabilities used in
the API study are about four orders of magnitude lower than used by SNL in the 1992 WIPP
PA (SAN92, Volume 3).
Currently, all dry holes from gas and oil exploration are plugged per federal and state
standards. Producing wells are not monitored, nor are abandoned formerly-producing wells
certified as plugged. Whereas OCD conducts an active program of institutional control for
all new wells on state and private lands, BLM performs only random and infrequent checks
on new wells located on Federal land. The number of unplugged boreholes drilled prior to
the more stringent institutional controls now employed is unknown, but has been
characterized as "many" by OCD field personnel.
9-16
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The 40 CFR part 194 rule assumes that natural processes will degrade or otherwise affect the
permeability of boreholes over the regulatory time frame. The issue of unsealed or
improperly sealed boreholes must also be factored into analysis of the repository integrity.
9.3.1.5 Human Intrusion Scenarios
In addition to the radioactivity release scenarios involving direct transfer of waste to the
surface by a borehole which penetrates the repository, several other scenarios involving
human intrusion can be theorized. For example, SNL developed a family of scenarios
involving boreholes which penetrate the waste and are then plugged above the overlying
transmissive Culebra Member of the Rustler Formation. Solubilized waste then moves up
the borehole by brines found in underlying formations and then, due to the borehole seal,
laterally through the relatively transmissive Culebra toward the WIPP site boundary
(SAN92). The assumption of borehole sealing depth is consistent with the sealing practices
in the area and the regulatory requirements. The assumption is also reasonably analogous to
the other geologic systems that were analyzed so that DOE should be able to defend the
above scenario.
Conceivably some boreholes would miss the waste, but be drilled sufficiently near the
disposal region to encounter contaminated brine and rock in, for example, Marker Bed 139,
a brittle, fractured anhydrite layer, which immediately underlies the repository.
9.3.2 Exploratory Drilling for Potash
Potash is the generic name for various potassium salts often formed by the evaporation of
natural brines whose potassium content is normally expressed in terms of equivalent K2O.
Additional background information on potash niining is included in Sections 9.4.1.
The potash reserves and resources1 at WIPP lie within the McNutt potash zone of the Salado
Formation. The depth of the 11 identified ore zones in the McNutt, based on the ERDA-9
borehole, ranges from about 1,372 feet to 1,741 feet near the WIPP site (POW78) and the
McNutt dips generally to the east (CHE78). As noted above, the WIPP repository is located
1 According to GUZ91b, reserves are those resources that are currently economically recoverable with
currently available technology, and resources are mineral deposits that are not currently economical or have not
been discovered.
9-17
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in the Salado at a depth of 2,150 feet. The deepest potash resources are thus about 400 feet
above the waste repository. These ore zones vary widely in thickness and mineralization.
The zones are not continuous across the Delaware Basin, and certain ore zones are not
present in some of the boreholes evaluated. Even when mineralization is present in an ore
zone, it may not be sufficient to be of commercial interest. In some cases, mineralization is
absent altogether.
The potash resources of the Designated Potash Area (so designated by the Secretary of the
Interior, see 9.4.1) lie roughly in an alignment extending from northwest to southeast. Early
potash mining started along the northern and western fringe of the district and moved in
southerly and easterly directions into the Delaware Basin. Exploratory boreholes have
preceded the underground workings, thereby delineating the reserves for further exploitation.
Potash boreholes tend to cluster around these mines with occasional boreholes located far-
afield.
Exploration drilling is conducted in the area to delineate additional ore reserves. Since this
drilling is generally to depths of less than 2,150 feet (except to the east where the ore zones
dip downward), this event would be characterized as shallow drilling by 40 CFR part 194.
Drilling for potash is significantly different from drilling for oil and gas. In addition to
being more shallow, the holes are also smaller in diameter. Approximately 1,892 potash
coreholes have been drilled in the Delaware Basin (per BLM estimate), mostly within the
designated boundary of the Known Potash Leasing Area. Potash boreholes typically have
been drilled either into an undesignated competent stratum of the upper Salado Formation or
into the Vaca Triste Sandstone member, which forms the upper contact with the McNutt.
The size distribution for all holes examined can be grouped into Rustler Formation drill bit
sizes of 51/2 inches to 83/4 inches in diameter and Salado Formation core bit sizes of 31/2
inches to 53/16 inches in diameter. After a surface casing is set through the Rustler
Formation—sometimes extending into the upper Salado Formation—core bits are used to drill
through eleven ore zones in the McNutt potash zone. The cored section of the hole is not
cased. The bit size distribution is as follows:
9-18
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Bit Diameter Distribution
Rock Bit Dia.: 83/4" 8" 77/8" 611'16" 61/r 61M" 55/r 51/2"
Percentage: 6% 2% 8% 2% 6% 57% 17% 2%
Core Bit Dia.: 53/16" 43/4" 315/16" 37/sr 31/2"
Percentage: 1% 3% 12% 54% 30%
Most of the potash holes are terminated at or near the lower contact of the McNutt. Three
test holes drilled by the U.S. Geological Survey (USGS) were about 2,000 feet deep (JON78)
near the WIPP site. Holes as deep as 2,800 feet have been logged in contiguous townships.
Based on discussions with BLM personnel, potash drilling has occurred over a period of
about 70 years. (This is consistent with the fact that the U.S. Bureau of Mines (SEA94) has
reported saleable potash production from New Mexico since 1933 and records ore grades as
early as 1930.)
BLM has a permitting procedure similar to oil and gas for exploratory potash coreholes on
Federal lands. The State Engineer requires approval for drilling of potash coreholes on non-
Federal lands if the drilling is through artesian aquifers or other water zones that require
protection. This requirement may not affect all locations around the WIPP site since the
Rustler Formation is not considered artesian.
Under a scenario that includes drilling for potash, all potash boreholes would pass through
the Culebra Member of the Rustler Formation. If a borehole contacted a contaminated plume
of Culebra water resulting from a prior human intrusion into the repository, it would bring at
least as much contaminated fluid to the surface as was in the volume of the Culebra rock
intercepted by the borehole. Furthermore, any radionuclides adsorbed on the solid material
would also be brought to the surface. For a 7.0 m thick Culebra aquifer with 0.16 porosity,
a 6 inch (15 cm) borehole would bring a bulk volume of 0.128 m3 to the surface. This
volume would contain 0.021 m3 of fluid. Assuming solubilities of 10"6 M for plutonium and
americium and 10"1 M for uranium results in a concentration of about 0.9 Ci/m3.
Considerably greater quantities of radionuclides could be present in the solid material brought
to the surface if surface adsorption in the Culebra is considered. The concentration (Cs) of
an element (e.g., Pu, Am, U) in the solids can be shown to be:
9-19
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Cs(g/cm3) = rs(g/cm3)*Kd(cm3/g)*Cbrine(g/cm3)
where rs is bulk density of the solids (about 2.0 g/cm3) and Kd is the solute distribution
coefficient.
In the 1992 PA (SAN92), SNL sampled on a matrix Kd value for plutonium from zero to
100,000 cm3/g with a median value of 261 cm3/g. Thus, for the median Kd value the
concentration of plutonium in a cubic centimeter of rock is 522 times that in a cubic
centimeter of fluid. Since there is only 0.16 as much fluid volume as rock volume in the
Culebra (i.e., the porosity is = 0.16), there would be 3,262 times the plutonium in the rock
phase as in the liquid phase, resulting hi a 59 Ci release to the surface.
If a potash corehole was not properly plugged and were to become a conduit for surface
water inflow, it could become a significant source of recharge to the Culebra. If located
upgradient from the repository, this recharge could increase the gradient which would
shorten water flow time to the accessible environment. Also, the larger water flow rates
could increase the quantity of radionuclides being transported if Culebra solubility is limiting
or it could decrease the amount being adsorbed in the solid matrix if the larger flow
decreases the radionuclide concentration. If the borehole is located down gradient from the
repository, recharge might be beneficial because it would decrease the gradient between the
repository and the recharge point. However, this inflow of fresher water would increase the
gradient between that point and the accessible environment and could also desorb previously
adsorbed radionuclides.
9.3.3 Water Well Drilling
Only limited water well drilling occurs around the WIPP site, since most of the water in the
area is too high in solids content to be suitable for drinking. Water wells may be used to
support oil and gas drilling, mining operations, or stock watering. An application to drill a
water well within the boundaries of a declared underground water basin, such as the
Carlsbad Underground Water Basin, must be made to the State Engineer. The State
Engineer requires a prospective water well driller to publish his application weekly for three
weeks in a local newspaper before a permit will be granted.
9-20
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Based on information obtained from the New Mexico State Engineer's Roswell District
Office and various groundwater reports (HEN51, NIC61), the following observations can be
made about water well drilling activities around the WIPP:
(1) Water wells are drilled for a variety of purposes. About 20% of the deeper wells
(i.e., wells in the Santa Rosa sandstone and lower-lying formations) drilled since 1952
were for oil and gas applications such as drilling muds or mining purposes. About
20% were drilled for stock watering. Several percent were listed as domestic and
observation use. Over 40% of these new wells are presently listed as unused and
there is no indication of why they were drilled.
(2) There are essentially no data on well pumping rates. Two Rustler Formation wells in
Nash Draw (in T22S, R30E) had reported pumping rates of 260 and 700 gpm. A
Rustler Formation well in T23S, R30E measured 3 gpm. Two wells in the Triassic
strata (in T23S, R31E and T23S, R32E) had yields of 10 gpm.
(3) No assessment has been made of water quality in these wells.
(4) There are very few data available on how extensively these wells are used. In the
wells listed as not being used it is not known whether these were dry holes, whether
they were ever used, or if they are likely to be used in the future.
(5) Within a given township, new wells are periodically being drilled at the same time
existing wells are classified as unused.
No water well drilling in the Carlsbad Underground Water Basin reached repository depths.
Therefore, water well drilling would be considered shallow drilling. A pumping water well
could, depending on its location, either increase or decrease the gradient in the Culebra
between the repository and the accessible environment. A borehole drilled into a
contaminated plume of Culebra water would bring some contaminated fluid and solid
material containing adsorbed radionuclides to the surface during drilling. For an average
borehole diameter of 12.5 inches (32 cm), based on 14 wells near the WIPP site, the area of
the hole is 0.079 m2. For a 7.0 m thick Culebra aquifer with 0.16 porosity this would bring
0.089m3 fluid in the 0.553 m3 of solids (bulk volume) brought to the surface. For assumed
solubilities of 10"6 M for plutonium and americium and 10"4 M for uranium, only 0.080 Ci
would be transferred to the surface hi the fluid. Radionuclide quantities adsorbed on the
solids brought to the surface would be somewhat greater if some adsorption of radionuclides
on the dolomitic Culebra rocks occurs.
9-21
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The largest potential consequence would come from an on-site well pumping from a
contaminated aquifer. The testing of a well could have significant consequences. For
example, some wells on the WIPP site have been pumped at 3 gallons per minute (0.19 1/s)
or greater for extended periods. Pumping of a well at this rate for 72 hours would bring 49
m3 of water to the surface. This volume would carry about 43 Ci to the surface (accessible
environment) at 1,000 years (for 10"6 M plutonium and americium and 10"* M uranium
solubilities). Greater quantities of radionuclides could be brought to the surface if a well
were placed in regular production. However, the number of curies brought to the surface
could be much less than suggested by this calculation if the actinide concentrations in the
plume were lower (because of lower solubility limits or because most radionuclides had been
removed by chemical adsorption), or if much of the water being pumped was not from the
contaminated plume.
9.3.4 Other Exploratory Drilling
Limited exploratory drilling for other resources has occurred around the WIPP site and could
occur in the future. For example, drilling for uranium in shallow lying sediments has been
conducted in the past. No evidence of uranium was found in the gamma logs from 36
boreholes near the WIPP site which penetrated the near surface Dewey Lake Redbeds, the
Santa Rosa sandstone, or the Gatuna Formation. Although uranium could occur in these
types of sediments, no significant occurrence has been found in the Delaware Basin
(POW78).
Sulfur is found in the Castile Formation in the Central Delaware Basin mainly in Culberson
County, Texas about 50 miles south of the WIPP site (SIE78, POW78). The sulfur appears
to be associated with portions of the Castile which lack halite either due to removal by
dissolution or to absence during deposition. These controls predominate hi the southern and
western portion of the Delaware Basin. Since the WIPP site lies east of the edge of the
Castile halite, occurrence of economic sulfur deposits is unlikely there.
Quantities of lithium are found dissolved hi the brine reservoirs in the Castile Formation
which underlies the Salado Formation containing the WIPP repository. Average lithium
concentrations of 240, 280 and 360 mg/1, respectively, were reported for the ERDA-6,
WIPP-12, and Union wells (DOE83). The reservoir intersected by the WIPP-12 well lies
within the LWA boundary and has a "representative" estimated volume of 2.7 x 106 m3 or
about 17 x 106 bbl (DOE83). Based on the estimated reservoir volume and measured brine
9-22
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chemistry, the reservoir would contain about 0.75 x 106 kg of Li. This is equivalent to only
about two or three months of domestic production at current rates (BOM94c). In an area of
about 775 km2 around the WIPP site, 12 of 92 boreholes penetrating the Castile Formation
intercepted brine (SAN92, Vol. 3), but only a few of these holes were assayed for lithium.
A variety of other minerals are present around the WIPP site, including salt, caliche, and
gypsum. However, these minerals are generally sub-economic, and no significant drilling is
currently involved in their discovery and exploitation (SIE78). Depending on the depth of
the drilling target, exploration for these other minerals could be classified as either shallow
drilling or deep drilling as defined hi 40 CFR part 194.
9.4 INTRUSION BY MINING
9.4.1 Introduction
EPA requires that consideration of mining-related scenarios should be included in assessing
the performance of the WIPP repository (§194.32). This requirement applies to mining of
all minerals, although the major commodity currently extracted hi the Delaware Basin by
underground mining is potash. Economic deposits of this mineral are confined to the
northern portion of the Basin in Eddy and Lea Counties, New Mexico near the WIPP site.
No other significant underground mining occurs in the Delaware Basin, although some sulfur
is extracted via Frasch process wells (hi the Castile Formation) hi Culberson County, Texas.
As previously noted, potash is a general term for a variety of potassium bearing minerals for
which the chemical compound K2O is often used as a surrogate to characterize the potassium
content. About 95% of U.S. potash sales are to the fertilizer industry with the balance
primarily to the chemical industry. Historical sources of potash include kelp, wood ashes,
lake brines, alunite, cement dust, sugar beet waste, blast furnace dust, and various
potassium-rich minerals. Today, U.S. potash production is principally from the rock
sylvinite - a mixture of the minerals sylvite (KC1) and halite (NaCl) - and from langbeinite
a potassium magnesium sulfate (K2SO4»2MgSO4). Potash is typically recovered either by
underground excavation mining or by solution mining where water is injected into a
mineralized zone and saturated brine is extracted and recrystallized hi evaporation ponds. In
the Delaware Basin, potash is recovered only by excavation mining.
9-23
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Extensive underground potash mining is currently being conducted in the vicinity of the
WIPP site. During 1992, southeastern New Mexico supplied 81% of U.S. production
(DUP94). Mining operations occur in the McNutt potash zone of the Salado Formation. A
generalized stratigraphic column showing these Upper Permian potash-bearing rocks and
younger strata is included as Figure 9-3 (CHE78). Eleven ore zones have been identified
within the McNutt. Primary current mining targets are the 10th ore zone for sylvite and the
4th ore zone for langbeinite. Some mineralization has been identified in ore zones 2, 3, 5,
8, 9, and 11 in the WIPP vicinity2(NMB95). These mineralized zones extend within the
WIPP Land Withdrawal Boundary as shown in Figure 9-4 which plots the boundaries of the
current Bureau of Land Management (BLM) Lease Grade criteria3 as estimated by Griswold
(GRI95). Reserve and resource estimates inside the WIPP boundary are summarized in
Table 9-1 (NMB95). When the WIPP site was selected in 1976, most of the site lay outside
the boundary of the Known Potash Leasing Area (KPLA) (i.e., the area which contains lease
grade reserves). However, subsequent site evaluation by DOE (then ERDA) included
drilling and coring 21 exploratory holes for potash (POW78). This drilling program
indicated that potash mineralization was more extensive than expected. As a consequence,
the U.S. Geological Survey used these drill hole data to extend the KPLA. The KPLA now
embraces all of the WIPP site although most of the southwestern quadrant of the site is
barren of mineralization, as is the repository location.
The WIPP site also lies within what is called the Designated Potash Area. This area, which
is defined by Order of Secretary of Interior (51 FR 39425) under the authority of two
mineral leasing acts, is slightly larger than the KPLA. It should be noted that the northern
most townships within the Designated Potash Area lie outside the northern boundary of the
Delaware Basin4 According to the Secretarial Order, potash enclaves are delineated within
the Designated Potash Area as regions containing currently economically minable ore
2 The New Mexico Bureau of Mines and Mineral Resources reserve and resource estimates are based on 40
drill holes in and around the WIPP site. Other drill holes exist in the area, but the data are proprietary. These
40 drill holes cover the WIPP Land Withdrawal area and an area extending about 1 mile outside the boundary
except for the southwest quadrant of this perimeter area (GRI95).
3 The current BLM leasing criteria for potash reserves specify ore seams containing at least 4 feet of 4%
K20 (a grade-thickness product of 16) for langbeinite and 4 feet of 10% K2O for sylvite (a grade-thickness
product of 40). These criteria have been in effect since 1969. According to BLM, sylvite is being mined
below the 10% K2O minimum cutoff grade and langbeinite is being mined below the 4% minimum (CON95).
• About 50% of the KPLA lies outside the Delaware Basin.
9-24
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Alluvium- O-3O' thick, volley alluvium, caliche M*l«, eoHon (and.
Go tuna Formation- O-2OO' thick, eandelono. locally derived conatomereto, elltitone.
Santo Rosa Sandstone- O-2OO' thick, coarM eandetono
-+*f*K-V-l
Dewey Lake R«d Beds- 20O-23O' thkk, eondetono, elltrtone.
Rustler Formation- S5O-4SO'thick, anhydrite, aypeum, holht, dolomite, mudetone.
Maoenta Member
Culebro Dolomite Member
Salado Formation- I9OO-2IOO' thick, halite, anhydrite, polyhalllo, potaeh, rare cloy.
Mofk«r bed no. IO3 • palyhome/anhydrrte.
Vaca TrUN S« Mwntor- top of Me Nut t paloiti ZOIM.
ion ZOMC ovfwrolly orglllocooiM hallt*)
Union AMiydrlt* M«nb«r
Marker Bod 126, DOM of Me Null fxrtath zora.
It Of Adorns (1944)
& Of Dean and Tung (I9T4)
Castile Formation-oporailnMtoly 2000'thick,
an*-ydrllo, gypuim, halite.
Fletcher Anhydrite Member
Voca Trttte Si Memoer
EleveMh are zone
Marker bed 117
Marker bed 118
Marker bed 119
Tenth ore zone • important >yMte end
longbemtte bearing on IOM
Marker bed 120
Ninth ore zone
Marker bed 121
Marker ted 122
Eighth are zone
Union Anhydrite Member
Seventh ore zone
Sixth are zone
Fifth ore tone
Marker bed 123-anhydrite and pdyMIIK
Marker bed 124 onnydrIM ono polynallte
Fourth are zone-Important longbelnile bearing or* ton.
Third ore lone
Second ore tone
Marker bed 129 - (not atwayi pr«Mep in oottern part.
Fclyr-eliT* and onrtydrit* marker bedf numbered leauenllally from No IOO at approx SO' below top ot Salado Fm. down 10 Nc. M3 01 appro*. HX> above
Figure 9-3. Generalized Stratigraphic Column of Permian and Younger Strata, Eddy
County, New Mexico (CHE78)
9-25
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• in mi mi «n sin
-------
Table 9-1. Potash Reserves and Resources Within WIPP Site Boundary (GRI95)
: AREA
4th Ore Zone
(Langbeinite)
10th Ore Zone
(Sylvite)
Other Ore Zones
TYPE OF RESOURCE
In-place resource (>4% K2O and actual
thickness)
BLM Lease Grade reserve (>4% K20 and 4-
foot mining height)
Minable reserve (>6.25% K2O and 6-foot
mining height)
In-place resource (> 10% K2O and actual
thickness)
BLM Lease Grade reserve (> 10% K20 and 4-
foot mining height)
Minable reserve (> 12.25% K2O and 4.5 foot
mining height)
In-place resources3
! MILLION
SHORT^TONS
47.0
40.5
18.0
53.7
52.3
30.6
18.4
K£)(%y
7.12
6.99
7.59
14.26
13.99
15.00
5.74-15.71
a Generally do not meet lease grade standards. According to GRI95, these resources could only be
minable if advanced thin-seam mining techniques are developed in the future.
reserves. Inside these enclaves, it is Department of Interior policy to deny approval of most
oil and gas drilling permit applications from surface locations with two exceptions (51 FR
39425): •
"a. Drilling of vertical or directional holes shall be allowed from barren areas within the
potash enclaves when the authorized officer determines that such operations will not
adversely affect active mining operations in the vicinity of the proposed drillsite.
b. Drilling of vertical or directional holes shall be permitted from a drilling island
located within a potash enclave when: (1) There are no barren areas within the
enclave or drilling is not permitted within on the established barren area(s) within the
enclave because of interference with mining operations; (2) the objective oil and gas
formation cannot be reached by a well which is vertically or directionally drilled from
a permitted location within the barren area(s); or (3) in the opinion of the authorized
officer, the target formation beneath a remote interior lease cannot be reached by a
well directionally drilled from a surface location outside the potash enclave."
For perspective, the Designated Potash Area, as of October 1986, occupied 497,002 acres as
compared to the area of the WIPP site which is 10,240 acres.
9-27
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Drilling on state and private lands is controlled by the New Mexico Oil Conservation
Division (OCD). Because of problems in implementing then existing OCD regulations, a
revised order (No. R-lll-P) was approved by the State Oil Conservation Commission on
April 21, 1988 (OCC88). Under the terms of R-lll-P, the New Mexico "Potash Area" is
coterminous with the KPLA. Within the Potash Area, drilling for oil and gas cannot be
conducted at any location containing life-of-mine potash reserves (LMR) except by mutual
agreement of the lessor and lessee of both the potash and oil and gas interests. Outside the
LMR, drilling of shallow wells can be no closer than 0.25 miles of the LMR boundary or
110% of the ore depth, whichever is greater. (Shallow wells are defined as those in all
formations above the base of the Delaware Mountain Group or less than 5,000 feet deep,
whichever is less.) Deep wells must be at least 0.5 miles from the LMR boundary. One of
the objectives of R-lll-P was to eliminate the need for drilling islands and three-year mining
plans required by the Secretarial Order on Federal lands.
Potash ore reserves in the Carlsbad KPLA were estimated to be about 100 million short tons
(90.7 million metric tons) of recoverable K2O based on 1973 prices (WEI79)5. At current
production rates of about 1.4 million metric tons per year (DUP92), this reserve would be
exhausted in about 65 years (about 15 years after projected completion of the WIPP disposal
phase, but during the period of active institutional controls)6. In the 1993 WIPP Resource
Disincentive Report, DOE commented on the finite nature of the langbeinite supply noting
that langbeinite operations would continue for another 28 years if only current reserves are
considered and the production period would be extended to 46 years if resources were also
included (DOE93). In 1993, the New Mexico Bureau of Mines and Mineral Resources
provided a breakdown of the expected operational life of each mine in the area. As shown in
Table 9-2, life of the Mississippi Chemical operations is projected to be 125 years while the
other five mines should wind down hi 33 years or less (BAR93). It should be noted that the
mine life estimates are based on published information. Data on actual mining reserves are
regarded as proprietary information by the potash mining companies and actual mine life may
be longer than projected here.
5 In a 1978 study, AIM Inc. estimated potash reserves for the Carlsbad District including those within the
WIPP site to contain 109 million tons of recoverable products - a total very similar to the 1973 Bureau of Mines
estimate (SEE78).
6 In 1973, the U.S. Geological Survey stated that, based on then current production levels, crystalline
deposits and brines in the U.S. would last for at least 100 years (SMI73). Nearby Canadian resources are
adequate for thousands of years.
9-28
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Table 9-2. Active Potash Mines in New Mexico Showing Estimated Capacity, Average
Ore Grade, and Mine Life at the Average 1992 Price of $81.14/st product
Operator
Eddy Potash Inc.2
Horizon Potash Co.
IMC Fertilizer, Inc.
Mississippi Chemical
New Mexico Potash2
Western Ag-Minerals4
County
Eddy
Eddy
Eddy
Eddy
Eddy
Eddy
Product Capacity
(st/yr1)
550,000
450,000
1,000,0003
300,000
450,000
400,000
Ore Grade
(% Kp)
18
12
II3
15
14
85
Mine::Lrfe
(yxs)
4
6
33
125
25
30
Data from J.P. Searls, U.S. Bureau of Mines, oral communication, 1993.
1 May not be operating at full capacity.
2 Owned by Trans-Resource, Inc.
3 Muriate, langbeinite, and sulfate combined.
4 Owned by Rayrock Resources of Canada.
5 Langbeinite only.
Current mining operations can be economically extended to the WIPP site boundary and it is
likely that this will occur (GRI95). Although economic mineralization also lies within the
WIPP site, the WIPP Land Withdrawal Act (LWA) (Public 102-579) precludes mining within
the withdrawn area. However, at some future time, when active institutional controls no
longer exist and if passive institutional controls are ineffective, mining of the potash inside
the boundary is a conceptual possibility. The economics would, of course, be different and
exploitation would probably require creation of a new infrastructure to transport ore to the
surface and beneficiate it since existing facilities would have been abandoned. GRI95
estimates of minable reserves within the site boundary assume that new mine and plant
facilities would not be needed if the reserves were exploited now. As noted in Table 9-1,
minable reserve estimates are based on higher grades and greater ore seam thicknesses than
for Lease Grade reserves.
Potash was first produced from the Delaware Basin hi 1931 (BAR93). Measured potash
reserves cover an area of approximately 200 mi2 in the Delaware Basin with the remainder of
the reserves located over the Capitan Reef or outside of the Delaware Basin. Since 1931,
mining of the different potash ore zones has covered an area (hi the Delaware Basin south of
T20S) of over 40 mi2 as estimated from a 1993 map of the potash resources (BLM93).
Using 9700 mi2 as the approximate area of the Delaware Basin, it can be estimated that about
0.4% of the Delaware Basin has been mined over the past 62 years (1993-1931). This
produces a conservative estimate of the rate of mining of 1 % of the Delaware Basin area
9-29
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over the past 100 years. Any mining of potash or other minerals of current interest
elsewhere in the Delaware Basin would raise this percentage.
The following sections discuss potential impacts of mining on the anticipated long-term
performance of the WIPP repository and elaborate on the position taken by EPA in the 40
CFR part 194 rule (§194.32(a)) that performance assessment shall consider the effects of
mining on the disposal system and these effects can be limited to changes in the hydraulic
conductivity of the disposal system induced by mining.
9.4.2 Mining Scenarios
Consideration of mining effects on PA involves scenarios where mining occurs up to the land
withdrawal boundary and where mining occurs within the withdrawn area up to the limits of
economic mineralization. Mining outside the site boundary could occur at anytime until
available resources are exploited. Mining activities inside the boundary should not occur
until sometime after active institutional controls are no longer practicable. The types of
scenarios will generally be the same regardless of the assumed location of the mining
operations and will, in the main, involve events which alter the rate and volume of
radionuclide movement through groundwater to the boundary of the accessible environment.
It does not appear that mining can seriously impact repository performance unless boreholes,
which intrude the waste panels, are also present. Without the presence of an intruding
borehole, there is no obvious way to connect the waste with the overlying water-bearing
formations which can then provide a lateral transport path.
The most common mining scenario assumes that subsidence of overburden into the excavated
region can alter the hydraulic conductivity of the overlying water-bearing strata (e.g., the
Culebra Dolomite Member of the Rustler Formation), possibly increasing transport velocities
and/or radionuclide mass-fluxes to the accessible environment. SNL summarized the
situation as follows (AXN94):
"Although the land surface in subsiding areas is lowered and there may be local changes in
drainage patterns, the overall topographic features that have the primary effect on the
water table will remain similar to those of the present. However, subsidence may have
impacts other than lowering of the land surface, including possible fracturing of units that
overlie the potash zone. This fracturing could lead to an increase in conductivity for those
units. The degree of increase and the relative change hi conductivity from unit to unit
could have an effect on the long-term groundwater flow behavior for Rustler units.
9-30
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Because the Tamarisk and Forty-niner members presently have very low conductivities,
fracturing may cause larger [percentage] increases in conductivity in those units than in the
Culebra and Magenta. The effect on flow would be similar to that described for boreholes
that do not intrude the repository [ref. omitted] for the same fundamental reasons. That
effect would be a change in the direction of the hydraulic gradients in the land withdrawal
area. Currently they direct flow hi the Culebra from north to south. If the scenario were
to occur, they would direct flow in the Culebra towards the southwest."
Detrimental mining-related scenarios might include:
• Increased hydraulic conductivity of water-bearing formations above the mining horizons
due to subsidence (Section 9.4.4)
• Change in flow directions within water-bearing members if a vertical hydraulic
connection is created by subsidence (Section 9.4.5.2)
• Formation of subsidence-related surface depressions where water could accumulate and
alter local recharge characteristics (Section 9.4.5.3)
• Increased hydraulic gradient if significant flow from water-bearing strata into the mine
workings occurs (Section 9.4.5.4)
• Damage to borehole or shaft seals by subsidence effects (Section 9.4.5.1)
• Problems created by solution mining (Section 9.4.5.1)
• Increased hydraulic conductivity of the Salado due to excavation induced stresses
(Section 9.4.5.6)
Depending on the location of the mining operations, some of these same scenarios may
actually be beneficial. Depending on the location, for example, flow of water into
underground mine workings might also reduce the hydraulic gradient hi the currently
envisioned flow path. Of the potentially detrimental scenarios, the only one expected to be
of concern is hydraulic conductivity increases hi certain strata above the mining location.
The detrimental aspects of these scenarios will be discussed in more detail subsequently, but
a review of relevant technical literature will be presented first to establish a framework for
that discussion.
9-31
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9.4.3 Literature Review
9.4.3.1 WIPP Related Studies
Final Environmental Impact Statement
In the WIPP Final Environmental Impact Statement (FEIS) published in 1980, DOE
summarized, without comment, prior studies on potash mine subsidence in the area as
reported by the BLM in 1975 (DOE80). At that time, it was estimated that subsidence was
likely to have occurred over an area of 14 square miles and was expected over an additional
40 square-mile area. The nearest subsidence to the WIPP site occurred at a distance of 3.5
miles. Observed maximum surface subsidence varied from 2.7 to 5.3 feet. This is about
two-thirds the height of the mined ore zone.
D 'Appolonia Studies
The impact on the WIPP of neighboring potash mines was examined in greater detail by
D'Appolonia in 1982 (DAP82). They observed that, even when subsidence occurs, the
integrity of the overlying salt section is not jeopardized as demonstrated by the absence of
water flow into the potash mines from units higher in the stratigraphic section.
However, D'Appolonia noted that "the opening of entries for underground potash mining
causes a redistribution of stresses within the surrounding rock that can lead to opening of
fissures and/or increase the hydraulic conductivity of the surrounding rock. Mining can also
lead to the more gross effects of surface subsidence and subsidence-induced fracturing above
the mined level." Both empirical and simple analytical techniques were used to characterize
the extent of such disturbances.
Using a secondary creep law for the salt, they calculated the zone of influence in a horizontal
plane around a hypothetical potash mine (at depth of 2000 ft) and a repository room to be
1,900 and 200 feet, respectively. Thus, if the horizontal separation is 2,100 feet, there
would be no stress-induced interaction between the two mined regions. D'Appolonia
believes this calculation to be conservative because the WIPP also has a vertical separation •
from the McNutt of about 400 feet.
9-32
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Estimates were also made of the impact on the hydraulic conductivity of the salt from
reducing the confining stress in the salt. This occurs due to stress relief around an
excavation. Based on an empirical relationship between salt permeability, octahedral shear
stress, and mean confining stress, D'Appolonia calculated the increase hi hydraulic
conductivity to be less than one order of magnitude. At a distance into the salt of six times
the width of mine opening the calculated hydraulic conductivity was only about twice the
conductivity of the undisturbed salt.
D'Appolonia suggested that a generalized subsidence equation developed for coal mines in
the Appalachian region could be used for making preliminary estimates of the magnitude of
surface subsidence as follows:
S = sHbe (1)
where
S = maximum subsidence (ft)
s = subsidence factor (dimensionless)
H = cavity height (ft)
e = extraction ratio (dimensionless)
b = fraction of cavity remaining after backfill (dimensionless)
The subsidence factor is the ratio of the actual vertical displacement to cavity height which in
the Carlsbad area is about 0.67. From the equation, assuming no backfill (b=l), a mining
height of 6 feet, and an extraction ratio of 90%7 the maximum subsidence would be about
3.6 feet (1.1 m).
As noted previously, potash is sometimes recovered by solution mining although this
technique is not being used in the vicinity of the WIPP. According to D'Appolonia, solution
mining of langbeinite is not technically feasible because the ore is less soluble than the
surrounding evaporite minerals. Solution mining of sylvite was unsuccessfully attempted in
the past. Failure of solution mining was attributed to low ore grade, thinness of the ore
beds, and problems with heating and pumping injection water. Unavailability of water hi the
area would also impede implementation of this technique. For these reasons, solution mining
is not currently used in the KPLA.
7 According to BAR93, 60 to 75% of the ore is extracted during initial mining, but subsequent removal of
the remaining pillars results in extraction ratios exceeding 90%.
9-33
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IT Corporation Backfill Engineering Analysis
In 1994, IT Corporation reported the results of analytical and empirical subsidence studies of
the WIPP repository (ITC94). The thrust of these studies was to evaluate the effects of
various backfill options on repository subsidence. The effects of potash mines in the vicinity
on repository integrity were not addressed, per se. Never-the-less, some generally applicable
subsidence information was developed. IT used four techniques to analyze subsidence caused
by excavation of the repository:
• Mass conservation method
• Influence function method
• National Coal Board method
• Two-dimensional numerical modeling (with the Fast Lagrangian Analysis of Continua
[FLAG] computer code)
As shown in Table 9-3, reasonable agreement was obtained among the four techniques with
maximum subsidence at the surface calculated to vary from 0.55 to 0.95 meters for the
empty waste area.
Using the FLAG two dimensional, finite element code, the maximum vertical tensile strain in
the Culebra Dolomite due to projected WIPP subsidence was calculated to be 0.0034%.
Using the influence function method,8 ITC developed contour plots showing the areal extent
of surface subsidence caused by repository excavation. The limit of subsidence area was
about 850 feet beyond the southern edge of the repository footprint. From this analysis, ITC
concluded that, since the maximum subsidence was about 0.4 m and since local surface
topography varied by more than 3 meters, a subsidence basin would not be created and
repository subsidence should not be visible.
8 The influence function method assumes that each point in an excavation has an identical circular area of
influence on surface subsidence. These influence areas are superimposed to obtain the cumulative effect of all
extraction elements.
9-34
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Table 9-3. Summary of IT Corp. Subsidence Prediction Results for WIPP Repository (ITC94)
Underground Area
Waste Emplacement
Area8
Shaft Pillar Area
Northern Experimental
Area
Contents of
Excavation
Empty
Waste Only
Waste plus loose
backfill
Waste plus
compacted backfill
Empty
Loose backfill
Compacted backfill
Empty
Loose backfill
Compacted backfill
Subsidence
Mass Conservation
(m)
0.86
0.62
0.55
0.52
0.28
0.12
0.06
0.24
0.11
0.05
Influence Function
Method
(m)
0.56
0.40
0.36
0.34
0.10
0.04
0.02
0.08
0.04
0.02
NCB Method
(m)
0.73
0.53
0.47
0.44
0.04
0.02
0.01
0.02
0.01
0.01
FLAC Single-
Room Model
(m)
0.95
NA
0.33
0.30
NA
NA
NA
NA
NA
NA
FLAC Full-
Panel Model
(m)
0.55
NA
NA
NA
0.13"
NA
NA
NA
NA
NA
a Waste emplacement area includes Panels 1 through 8; 2 through 8 are not yet excavated.
b At the Waste Shaft
NCB National Coal Board
FLAC Fast Lagrangian Analysis of Continua
NA Not Available.
m Meters.
-------
Sandia Studies of Subsidence
Sandia has explored the possible impact of WIPP subsidence on performance assessment
(PA). In a 1989 study to select events and processes which should be considered in forming
possible scenarios, SNL considered three possible processes related to repository-induced
subsidence (HUN89):
• Increased hydraulic conductivity of the Salado Formation
• Fracturing
• Disruption of surface drainage
Based on the fact that repository excavation would produce a maximum of a 0.2% increase
in the volume of the overlying Salado salt, they concluded that increased Salado hydraulic
conductivity would be insignificant. They further concluded that fracturing of the Salado
could also be neglected. This conclusion was based on the expectation that the repository
would adjust to excavation by creep rather than fracturing. This position was supported by
observations in local potash mines where mining was conducted with two levels of
extraction. The observed response of the rock in the upper horizons was flexure rather than
fracture. However, SNL stated that effects on the Culebra were unknown. With regard to
surface drainage, SNL concluded that this would not be a factor because, with a maximum
expected surface subsidence of 2 feet, there was no integrated drainage which would be
disrupted.
As noted in Section 9.4.2 above, SNL revisited the subsidence issue in 1994 concluding that
subsidence could cause fracturing in the more brittle overlying units which could result in
increased hydraulic conductivity and possible redirection of flow in the Culebra from a
generally north to south direction to a more southwesterly direction (AXN94). Surface
subsidence effects were not expected to be of sufficient magnitude to significantly alter the
position of the water table.
9.4.3.2 Other Relevant Studies
IT Corporation summarized subsidence observations made at potash mines in southeastern
New Mexico (ITC94). Observed angles of draw, measured from vertical edge of the mine
workings to the point where surface subsidence ceased, varied from 25 to 58 degrees. ITC
noted that the maximum observed subsidence over four potash mines in the area varied from
9-36
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0.4 to 1.5 m which was between 16 to 66% of an assumed excavation height of 2.6 m
(8.5 ft)9. ITC felt that the maximum observed subsidence was less than that which will
ultimately occur over the excavated area.
A large body of subsidence literature has been developed based on coal mining in the United
States and the United Kingdom. In a number of studies, subsidence-induced increases in
transmissivity are described. Some examples are provided here.
The U.S. Geological Survey described the effects of subsidence associated with longwall
mining of coal in Marshall County, West Virginia (USG88). Three tests were recounted
where the transmissivity of a perched aquifer was measured before and after mining a coal
seam. In each case, the overburden was about 800 feet thick and the tested aquifer was
between 25 and 150 feet below the surface. In two tests, the transmissivity was found to
increase significantly, from 3.7 to 160 ft2/day in one case and from less than 0.001 to
36 ft2/day in the other. In the third test, only a slight increase between pre- and post-mining
transmissivity was observed (from 0.20 to 0.31 ft2/day). This small change was attributed to
the fact that significant subsidence fracturing had not occurred.
Booth discussed to similar studies related to longwall coal mining hi the Illinois Basin
(B0092). One series of tests was conducted at a site hi Jefferson County, Illinois where
coal seams 9 to 10 feet thick were mined at a depth of about 725 feet. The overburden
consisted primarily of low permeability shales, siltstones and limestones. An aquifer in
sandstone exists about 75 feet below the surface which is confined by an overlying shale unit.
Subsidence produced visible surface tension cracks. Subsurface strain measurements and
borehole examination indicated fractures and bedding plane separation. In three
presubsidence tests, the measured values of hydraulic conductivity hi the Mt. Carmel
sandstone were 2 x 10"6, 2 x 10"5, and 3 x 10"6 cm/s. After subsidence, measured values
were 5 x 10'5, 3 x 10'5, and 4 x 10'5 cm/s. In another paper discussing the same site, it was
reported that post-subsidence values of the hydraulic conductivity in the shale were increased
by two to three orders of magnitude (KEL91).
At a second site in Saline County, IL, investigations involved subsidence related to mining a
five- to six-foot coal seam at a depth of about 400 feet (BOO92). The Trivoli sandstone
9 The maximum observation period varied from one week to more than one year.
9-37
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aquifer lies above the seam and about 180 feet below the surface. Initial conductivities in the
Trivoli were less than 10'8 cm/s and these increased to about 5 x 10"6 cm/s after mining.
Booth attributed this increased conductivity to the supposition "that subsidence had probably
improved the interconnectedness of permeable fractures."
The U.S. Bureau of Mines described hydrologic changes associated with longwall mining of
coal in Cambria County, Pennsylvania (MAT92). The coal seams studied were at a depth of
740 to 845 feet and were overlain by fine-grained sedimentary rocks and thin coal beds.
Only small changes in hydraulic conductivity of the overburden due to mining were
measured. Increases were a factor of 2 to 4 and in some cases an unexplainable decrease
was noted. The increased conductivity was attributed to excavation-induced creation of new
passages for groundwater flow.
Elsworth and Liu used non-linear finite element modeling to estimate changes in hydraulic
conductivity associated with longwall mining (ELS95). In their modeling, a 140-foot thick
zone of increased horizontal conductivity caused by vertical strains was defined immediately
above a 5-foot thick coal seam. The estimated conductivity increase was about an order of
magnitude.
Bai and Elsworth described modeling studies involving the interrelationship between
subsidence and stress dependent hydraulic conductivity (BAI94). In concept, the rock
mechanics approach was similar to that taken here and described hi Section 9.4.4.2 below.
In the Bai and Elsworth studies, finite element analyses over representative stratigraphy were
used to calculate changes in hydraulic conductivity for various fracture spacings.
9.4.4 Impact of Mining on Hydraulic Conductivity
9.4.4.1 Background Information
Based on the available site information, it appears that one of the potential detrimental results
of mining near the repository could be increased hydraulic conductivity10 in the brittle
10 The terms hydraulic conductivity and transmissivity are sometimes used interchangeably in the text as
indicators of altered flow path resistance. Transmissivity is the product of the hydraulic conductivity and
aquifer thickness. In the examples presented here, the Culebra thickness is assumed to be constant so the
transmissivity is a constant factor of 7.7 higher than the hydraulic conductivity (in metric units).
9-38
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water-bearing strata above the mining horizons. In the analysis discussed here, the focus is
on the Culebra Member of the Rustler Formation which is the most transmissive unit. The
Culebra can potentially provide a lateral conduit to the accessible environment if
contamination from the repository 1440 feet below reaches the transmissive horizon.
According to SNL, the Culebra is a "finely crystalline, locally argillaceous (containing clay)
and arenaceous (containing sand), vuggy dolomite ranging in thickness near the WIPP from
about 7 m (23 ft) .... to 14 m (46 ft) " (SAN92). In its 1992 performance assessment
(PA), SNL chose 7.7 meters as the reference thickness. Using information from 41
boreholes, SNL has calculated that the transmissivity of the Culebra varies by about six
orders of magnitude depending on the degree of fracturing which exists. In the 1992 PA
(SAN92), the median fracture spacing was assumed to be 0.4 m and range between 0.062
and 8 m. Thus, the median number of horizontal fractures through the Culebra thickness
would be 19 and the range would lie between 1 and 124.
If subsidence occurs, it may create a network of both vertical and horizontal strains in the
Culebra. Vertical tensile strains can increase the aperture of existing horizontal fractures;
whereas, horizontal tensile strains can increase the aperture of existing vertical fractures.
Compressive strains would have the opposite effect. Increase hi fracture aperture increases
hydraulic conductivity. This increased hydraulic conductivity can reduce lateral travel time
of radionuclides to the accessible environment at the vertical subsurface extension of the site
boundary.
As noted above, the 1992 PA assumed that flow and transport through the Culebra is through
fractures. In light of this 1992 PA assumption, the following discussion focuses on one
potential theory describing groundwater flow through fractures. The subsequent section
discusses how the fracture aperture increases can be estimated.
Darcy's law relates the movement of water in a porous medium to the hydraulic gradient and
the hydraulic conductivity. The hydraulic conductivity is a measure of the transmissive
capacity of the medium coupled with the density and viscosity of the fluid (water in this
case). The hydraulic gradient is simply the slope of the water table (unconfined aquifers) or
the potentiometric surface for a confined system. The equation for Darcy's law is
—
dl
9-39
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where q is the Darcy velocity (m/yr), K is the hydraulic conductivity (m/yr) and dh/dl is the
hydraulic gradient (dimensionless m/m). Hydraulic conductivity is actually a property of
both the physical media (the aquifer) and the fluid. Darcy's law may also be written using
intrinsic permeability (k) which is a property of the medium alone, as shown below:
where:
k = intrinsic permeability (m2)
p = fluid density (kg/m3)
H — viscosity (Pa-s)
g = gravitational constant (m/s2)
kpgdh
u dl
The advective flow rate for a conservative contaminant (i.e., non-sorb ing and nonreactive)
migrating through a porous medium is computed by dividing the Darcy velocity (given
above), by the effective porosity. The effective porosity for a porous medium is the ratio of
the connected void space divided by the total volume of the medium.
In a fractured medium, Darcy's law still applies, however, the hydraulic conductivity of the
fracture (Kf) is more difficult to determine. If the fractures are conceptualized as a series of
parallel plates (with the fractures being the gaps between adjacent plates), mathematical
equations can be derived to determine the equivalent hydraulic conductivity that would be
used in Darcy's law.
The porosity of the fracture system actually should be viewed as two components, fracture
porosity and matrix porosity. Using the parallel plate analogy, the fracture porosity is the
number of fractures times the fracture aperture (gap thickness) divided by the thickness of
the aquifer. The matrix porosity is the porosity of the blocks of rock between the fractures.
In a fractured system such as granitic rock, the matrix porosity may be effectively zero
because there is no intergranular void space. However, there is some measurable porosity
space within the Culebra matrix (SAN92).
The hydraulic conductivity of a system of horizontal fractures is determined by the fracture
aperture and the spacing between fractures. Given an equivalent hydraulic conductivity of
9-40
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the aquifer (i.e., determined through aquifer testing) and fracture spacing, it is possible to
compute the fracture hydraulic conductivity. The calculation is based upon moving the same
flux of groundwater through the fracture system as through a porous medium. The
derivation of this equation is developed below.
The fracture conductivity equation is derived hi two steps. First, the hydraulic conductivity
for a single fracture is defined and then this is related to the flow rate through the fracture.
The hydraulic conductivity of a single fracture is given as:
K=^ (1)
where:
b = half-fracture aperture (m)
Kf = fracture hydraulic conductivity (m/yr)
This equation is presented in a number of papers by Snow (SNO69) and by Gale (GAL82).
The equation is often rewritten in terms of the full fracture aperture, as follows:
K=^ (2)
where:
w = full fracture aperture (b2 = w2/4) (m)
The second step in computing the aperture from an equivalent porous medium K value is to
equate the flow rates through the porous and fractured systems. The flow through a set of N
horizontal fractures of identical aperture is:
'NWL^ <3>
dl
where:
Qf = flow rate through the fractures (m3/yr)
L = length of fractures perpendicular to flow (m)
N = number of fractures
9-41
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The term (NwL) is the area term in a traditional Darcy's law equation. The equation for
flow through an equivalent porous medium would be:
Q=KDL— (4)
£ dl
where:
Kf =
D =
I j ™~
equivalent porous medium hydraulic conductivity (m/yr)
aquifer thickness (m)
length perpendicular to flow direction (m)
As mentioned above, equation 4 may also be written in terms of intrinsic permeability (k)
and fluid properties, as show below:
dl
(4b)
To compute an equivalent K for the porous medium, the flow rates through the two systems
(porous and fractured) must be equal. Setting equation 3 equal to equation 4 yields:
KD=w3pgN
(5)
with common terms canceling from the equations. This equation can then be rearranged to
give an equation of fracture aperture in terms of an equivalent porous medium hydraulic
conductivity:
w =
12K
pgN
(6)
Finally, to get the equation in terms of spacing between fractures (Df = D/N), the equation
becomes:
w =
Pg
(7)
9-42
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After computing the fracture aperture for a given porous medium hydraulic conductivity
(equation 7), the fracture hydraulic conductivity is computed from equation 2 above. It can
be seen from equation 2 that the fracture hydraulic conductivity (Kf) varies as the square of
the aperture (w) while the equivalent porous medium conductivity (Ke) varies as the cube of
the aperture.
The following example provides an indication of the magnitude of changes which might be
expected in the Culebra hydraulic conductivity resulting from subsidence induced fractures.
For these calculations, it is assumed that the vertical tensile strain produced by subsidence
results in the opening of existing horizontal fractures rather than the creation of new
fractures. The total strain is accommodated by increasing the fracture aperture. Thus, if,
for discussion purposes, there is a single horizontal fracture in the Culebra and subsidence
from potash mining causes 0.03% vertical tensile strain (which is about 10 times the value
calculated in ITC94 for the Culebra from repository subsidence, see Section 9.4.3.1 above),
the total displacement is 2.3 x 10"3 m (7.7 m x 0.0003). If 10 horizontal fractures were
present, then the increase in each aperture would be 2.3 x 10"4 m.
The effect of subsidence on changes in fracture aperture and hydraulic conductivity of the
Culebra for the case of 10 fractures across the aquifer thickness is calculated using the
following assumptions:
aquifer thickness (D) = 7.7m
viscosityOi) = 0.001 Pa»s
density (p) = 1000 Kg/m3
gravitational constant (g) = 9.79 m/s2
equivalent hydraulic conductivity (KJ = 7.0 m/y = 2.24 x 10'7 m/s
tensile strain = 0.03% = 0.0003 m/m
total displacement = 7.7m x 0.0003 strain = 2.3 x 10'3m
The attendant fracture aperture from equation (6) is:
(8)
w =
PgN
Ke = equivalent hydraulic conductivity
w = fracture aperture
p — density
g = gravitational constant
N = number of fractures
/i = viscosity
D = aquifer thickness
9-43
-------
2.24jclQ-7 x 12 x 0.001 x 7.7
w =
1000x9.79*10
w = 5.96 x 10'5 m
For a total displacement of 2.3 x 10"3 m, the displacement per fracture is 2.3 x lO'4 m and
the expanded fracture aperture resulting from the tensile strain (wstrain) is
wstrain = 5.96 x 10'5 + 2.3 x 104 = 2.9 x 10-4 m
To calculate the strain-altered equivalent hydraulic conductivity, KeS,
(2.9xlQ-4)3 x 1000 x 9.79 x 10 (9)
12 x 0.001 jc 7.7
= 2.6;clO-5 mis = 8.2xl02 mly
Values of the equivalent hydraulic conductivity for various assumed values of N within the
range used hi the 1992 PA are summarized below based on 0.03% vertical tensile strain:
N (fractures) Hydraulic Conductivity (m/y)
1 14.8 x 104
10 8.2 x 102
100 4.4 x 101
From this hypothetical example, it can be seem that the change in hydraulic conductivity is
nearly four orders of magnitude for a single fracture and only a factor of six for 100
horizontal fractures through the thickness of the Culebra.
In order to provide a more detailed view of the impact of subsidence on repository
performance, a series of modeling simulations were made. First, the strain distribution in
the Culebra as a function of distance from the face of a potash mine was calculated using a
two-dimensional finite element model (the UTAH2 computer code). Then, this strain
distribution was assumed to be accommodated as increases in the aperture of existing
fractures. Details of these analyses are presented hi subsequent sections.
9-44
-------
9.4.4.2 Strain Analysis
A preliminary analysis was conducted to estimate the effects of simulated mining of potash
near the WIPP site on the hydraulic conductivity of the Culebra Member of the Rustler
Formation. Simulation of longwall mining of potash was done using a two-dimensional finite
element computer program, UTAH2. This program has been in use for many years and is
considered quite reliable (PAR78, PAR91). In response to mining, the adjacent rock mass
moves to a new equilibrium position. Maximum surface subsidence occurs above the center
of a mined panel, but diminishes with distance from the panel center. However, as will be
shown, maximum strains do not occur at the same location as maximum subsidence. Tensile
strains may open existing joints or fractures and fracture opening is assumed to increase
hydraulic conductivity. If tensile strain between existing fractures is assumed to be absorbed
entirely by fractures, then the change in fracture aperture can be calculated. With the
assumption of an initial aperture, the change in hydraulic conductivity can then be estimated
as shown in Section 9.4.4.1.
Finite Element Analysis
The UTAH2 finite element program is a small strain, elastic-plastic computer program that
uses associated flow rules in conjunction with a pressure-dependent yield criterion. Elastic
and strength anisotropy may be independently specified, but one material axis is tacitly
assumed to be normal to the plane of analysis. The form of the yield criterion is J2+Ij=1,
where J2 is an anisotropic form of the second invariant of deviatoric stress and 7; is an
anisotropic form of the first invariant of stress. The isotropic form is a paraboloid of
revolution about the hydrostatic axis in principal stress space. Essential input data include
the elastic moduli as well as the strength parameters, geologic column, mining geometry,
boundary conditions and the premining stress state.
Material Properties
For the isotropic case analyzed here, the strength parameters required are the unconfined
compressive (C0) and tensile (T0) strengths of each material represented in the finite element
mesh. The elastic parameters are Young's modulus (E) and Poisson's ratio (v) for each
material. Specific weights (7) of the various rock types present in the model region are also
needed. The data for the four rock types assumed in the model are given in Table 9-4.
9-45
-------
Sandstone, anhydrite and halite elastic properties were obtained from a subsidence analysis of
the WIPP repository conducted by the IT Corporation (ITC94). Strengths, with the
exception of halite were also obtained from (ITC94). Dolomite properties and halite strength
are averages of about 20 results obtained from a standard handbook (LAM78).
Table 9-4. Rock Properties By Type
Rock Type
Sandstone
Anhydrite
Dolomite
Halite
E(106 psi)
3.8
10.9
9.4
4.5
V
0.21
0.35
0.30
0.25
Q(103.psx)
15.0
13.3
13.3
5.2
T0(103 psi)
5.0
4.6
1.2
3.1
T(pcf)
144
144
144
144
Consideration of strength and elastic modulus properties for the Culebra shows that the strain
at failure under uniaxial compression is 0.14%. Under tension the strain at failure is
0.013%. Rock strength is strongly affected by confining stress, so under multiaxial
compressive stress, the strain at failure should be greater than in the uniaxial case. Tensile
strength is not considered sensitive to confining stress, so tensile strain at failure would also
be insensitive to confining stress. These estimates of failure strain are based on the
laboratory test data summarized in Table 9-4. The rock mass would have different properties
depending on fractures that are present in the field, but absent in the laboratory test samples.
Strains calculated using laboratory data will be lower than strains calculated using field-scale
properties.
Geologic Column
The geologic column used in the analysis was adapted from the ERDA 9 borehole near the
center of the WIPP site (POW78). Table 9-5 gives the depth, formation, and thickness of
the different strata represented in the finite element model.
9-46
-------
Table 9-5. Strata Depth and Thickness
FormatiOE
1. Dewey Lake
2. Rustler
3. Magenta
4. Rustler
5. Culebra
6. Rustler
7. Upper Salado
8. McNutt
9. Potash Seam
10. McNutt
11. Lower Salado
12. Storage Zone
13. Lower Salado
14. Storage Zone
15. Lower Salado
16. Castile
Depth (ft)
0
550
608
632
714
740
860
1,367
1,543
1,553
1,741
2,074
2,178
2,620
2,730
2,836
Thickness (ft):; /• >•
550
58
24
82
26
120
507
176
10
188
333
104
442
110
106
1.66411
As can be seen from Table 9-5, the base of the mesh includes a portion of the Castile to a
depth of 4,500 ft (2836+1664). All strata below the Rustler formation were assigned halite
properties from Table 9-4. The Rustler Formation was assigned anhydrite properties (except
for the Culebra and Magenta which were assigned dolomite properties) and the Dewey Lake
Formation was assigned sandstone properties. This assignment is the same as used in
ITC94.
11 The thickness assigned to the Castile does not include the entire unit, rather it is based on assumptions
regarding the necessary modeling depth required to minimize boundary effects.
9-47
-------
Milling Geometry
The mining panel was assumed to be 10 ft thick12, 3,000 ft long and located near the
middle of the McNutt. However, the center of the panel is assumed to be a line of
symmetry, so only 1,500 ft is explicitly represented in the mesh. As a rule of thumb, the
influence of an excavation extends "one diameter" from the excavation walls. At one
diameter, the stress concentration about a circular hole decreases to within about 15% of the
initial stress state. The "diameter" that characterizes non-circular holes is the long dimension
of the hole. In this case, the "1-D" guideline suggests that panel excavation may noticeably
influence the state of stress 3,000 ft away. Thus, about 3,000 ft was added to the panel
depth (1,543 ft) to obtain a vertical mesh dimension of 4,500 ft. The horizontal dimension
of the mesh extends 5,250 ft beyond the panel edge and is thus 6,750 ft. The mesh and
panel are shown in Figures 9-5 and 9-6 where the scale is 900 ft per inch. There are 4,050
elements and 4,216 nodes in the mesh. The element aspect ratio is 5 or less.
Premining Stress State
The premining stress state was attributed to gravity alone; no tectonic stresses were assumed.
The vertical stress is then simply the average unit weight of rock times depth. Under
complete lateral restraint, the horizontal premining stress is a constant, K0, times the vertical
stress. The constant depends on Poisson's ratio, v, and is therefore different for each rock
type. In fact, K0 = vl(l-v), which ranges from about 0.2 to 0.5 based on the values in Table
9-4.
Boundary Conditions
The centerline of a panel was a line of symmetry; no displacement was allowed normal to
this line. Zero displacement boundary conditions were also specified normal to the mesh
bottom and far side. A zero normal displacement is often represented by a roller. The top
of the mesh coincided with the ground surface and was unrestricted except at the sides.
12 This thickness was selected as a conservative value based on mine workings in the area (Section 9.4.3.2)
and to reflect the possibility of mining on multiple levels.
9-48
-------
CROUHD SURFACE
CULEERA
POTASH
Figure 9-5. Finite Element Mesh Used for Strain Analysis Mesh 4,500 ft by 6,750 ft.
Scale: 1 inch = 900 ft.
Figure 9-6. Half Width (1,500 ft) of Mined Panel. Scale: 1 inch = 900 ft.
9-49
-------
Displacement boundary conditions were also specified on the excavation surface. The panel
roof was specified to "sag" 9 inches per load step; the floor was specified to "heave" 3
inches per load step. Thus, 1 foot of closure occurred during each load step at every pair of
nodes along the panel except at the panel edge where traction boundary conditions, equal but
opposite in sense to the premining stresses, were applied. The amount of seam level closure
is controlled by the number of load steps specified, but is physically limited to a maximum
of 100% of the mined thickness (10ft).
A second physical constraint on seam closure is the amount of subsidence observed at the
surface. The number of load steps was adjusted to meet these constraints. Specifically,
seam level closure (relative displacement between roof and floor) is 70% when 7 load steps
are applied. The corresponding surface subsidence calculated at the panel centerline is
52.5% of the seam thickness. When 9 load steps are applied, seam level closure is 90%,
while surface subsidence is 67.5% of seam thickness. This range of surface subsidence is
considered reasonable for full-extraction potash mining.
Fracture Conductivity Change
As described above in Section 9.4.4.1, the parallel plate model for fracture flow states that
average flow velocity is proportional to the square of the width (aperture) of the fracture; the
volume flow rate (discharge) is proportional to the cube of the aperture (equation 3).
Fracture hydraulic conductivity, Kp is used here to relate flow velocity to hydraulic gradient
and is thus proportional to the square of fracture aperture (equation 2). The relative change
in hydraulic conductivity is (Kf - Kfo)/Kfo where Kfo is the premining fracture hydraulic
conductivity. A purely geometrical calculation gives the relative change. Thus, the relative
change in fracture hydraulic conductivity is (M^-W/J/W/, where w is the fracture aperture
after mining (i.e. wstrain) and w0 is prernining fracture aperture. This ratio is independent of
the units used for hydraulic conductivity such as feet or meters per year.
The post-mining aperture is simply the premining aperture plus the change in aperture, Aw,
induced by mining. This change is the strain, e, integrated over fracture spacing, Df, that is,
Aw = Djt. Fracture spacing was assumed to vary between 3 and 300 inches (ca. 0.08 m and
8 m); initial aperture was assumed to vary from 10"* to 10"2 inches13. Strains are obtained
from the finite element simulation of longwall potash mining.
13 In metric units these apertures are equivalent to 2.5 x 10"6 to 2.5 x 10^ m. This range of apertures
would be associated with equivalent hydraulic conductivities varying from about 6 m/y to about 60,000 m/y. In
SAN92, reported hydraulic conductivities (converted from transmissivities using an aquifer thickness of 7.7 m)
within the WIPP Land Withdrawal Area ranged from 0.026 to 4,400 m/y.
9-50
-------
Results
Two simulations were done. The first case was associated with a subsidence factor (S) of
52.5% (maximum surface subsidence as a percentage of mined panel height); the second case
was associated with a subsidence factor of 67.5%. The results are similar in trend, but differ
quantitatively.
Case 1.
Horizontal and vertical strains in the Culebra formation are shown in Figure 9-7 for this case
(S = 52.5%). The data are strains which are calculated at the centroid of the model
elements in the Culebra. Tensile strain is positive in Figure 9-7. The horizontal axis begins
at the left edge of the finite element mesh, that is, at the center of the mined panel. Mining
extends 3,000 ft, 1,500 ft of which is incorporated into the mesh. Figure 9-7 shows tensile
strain in the vertical direction over the mined panel (between 0 and 1500 ft) and horizontal
tensile strain beyond the edge of the panel (beyond 1500 ft). The peak vertical tension is
about 0.055% (550 micro-in./in) and occurs 1,075 ft from the panel center (i.e., 425 ft
inside the panel edge). The peak horizontal tensile strain occurs 175 ft outside the panel
edge and is 0.0085% (85 micro-in/in). The horizontal tensile strain initially decreases with
distance from this peak and then rises to a broad secondary maximum of about 0.0047% (47
micro-in/in) at 4,275 ft from the panel center after which it decays slowly with increasing
distance.
The horizontal strain changes from tension outside the mined panel to compression inside as
seen in Figure 9-7. The peak horizontal compression occurs inside the panel and gradually
decreases to a minimum at the panel center where the slope of the plot is zero. This trend is
indicative of a panel that is sufficiently wide relative to depth to cause maximum subsidence.
The panel has super-critical width in subsidence terminology. Critical width is usually given
in terms of the angle of draw: Wc = (2H)tan(6). If the angle of draw is 35°, e.g., then
critical width is 1.4H where H is the overburden thickness.
Vertical tensile strains would tend to open horizontal fractures, while horizontal tensile
strains would tend to open vertical fractures. Compressive strains would tend to close
fractures. The magnitude of the vertical tensile strain near the center of the mining panel is
about the same as the horizontal compressive strain outside the mining panel and away from
9-51
-------
(Panel extends 1500 ft +/- from zero)
C/D
0.08
2 4
(Thousands)
DISTANCE FROM PANEL CENTER (ft)
D E-H + E-V
Figure 9-7. Subsidence-induced Culebra strains for subsidence factor of 52.5%. (Panel extends +7-1,500 ft from origin)
-------
the rib. So the change in hydraulic conductivity of horizontal joints over the panel is about
the same as the change in vertical joint conductivity for a substantial distance outside the
mining panel (neglecting peaks near the rib).
Figure 9-8 shows the change in vertical fracture apertures (opening or closing) in the Culebra
as a function of distance from the panel center for three assumed joint spacings (3, 30 and
300 inches). Because vertical fractures or joints respond to horizontal strain, joint closure
occurs over the mined panel where the horizontal strain is compressive. Vertical joints tend
to open outside the mined panel. The magnitude of aperture change increases significantly
with joint spacing. Vertical joint opening which occurs outside the mined panel ranges from
nil to almost 0.03 inches near the rib.
Figure 9-9 shows the aperture change for horizontal joints (which respond to vertical strain).
The peak aperture changes at a 300-inch joint spacing are cut off in the plot. Horizontal
joint opening which occurs above the mined panel ranges from nil to well over 0.04 inches.
Figure 9-10 is a semilog plot of the relative increase in hydraulic conductivity of vertical
fractures, spaced 3 inches apart, that is induced by horizontal tensile strain outside the mined
panel. The relative change depends on the initial fracture aperture; 3 apertures ranging from
10^ to 10"2 inches are assumed in the construction of Figure 9-10. Only fractional increases
occur below the x-axis in Figure 9-10 (i.e., changes are less than an order of magnitude),
while orders of magnitude increase are shown above the x-axis. Figures 9-1 Ob and 9-10c
present similar results at joint spacings of 30 and 300 inches. Generally, the relative
increase is greater for smaller, more widely spaced joints or fractures.
9-53
-------
(V-joint, + opening, 3 spacings)
o
UJ
IT
IT
UJ
V-3
(Thousands)
DISTANCE (tt)
+ V-30
V-300
Figure 9-8. Aperture Change hi Vertical Joints for Fracture Spacings of 3, 30, and
300 inches and Subsidence Factor of 52.5%
o
UJ
o:
UJ
a.
<
0.04
(H-joint, + opening, 3 spacings)
v-3
(Thousands)
DISTANCE (ft)
V-30
V-300
Figure 9-9. Aperture Change in Horizontal Joints for Fracture Spacings of 3, 30, and
300 inches and a Subsidence Factor of 52.5%
9-54
-------
b
•3
(Thousands)
DISTANCE (fl)
A 10-3
(Thousands)
DISTANCE (It)
A 10-3
(Thousa/ids)
DISTANCE (H)
A 10-3
Figure 9-10. Relative Change in Fracture Hydraulic Conductivity for Vertical Joints
with various Fracture Spacings, Subsidence Factor of 52.5%, and Fracture Apertures
of 1Q-4, 10°, and 10'2 in. a) 3 inch b) 30 inch c) 300 inch
9-55
-------
Figures 9-1 la, b, and c show the relative increase in hydraulic conductivity of horizontal
joints over the mined panel. Joint spacings of 3, 30 and 300 inches were used for Figures 9-
11 a, b and c, respectively.
Case 2.
Horizontal and vertical strains in the Culebra formation are shown in Figure 9-12 for this
case (S = 67.5%). The peak vertical tension is about 0.071% (710 micro-in/in) and occurs
inside the panel as seen hi this figure. The peak horizontal tensile strain is about 0.0053%
(53 micro-in/in) and occurs 225 feet beyond the panel edge. With distance, the horizontal
strain becomes compressive, then reverses to tensile, and reaches a secondary maximum of
0.0053% (53 micro-in/in) at 4,375 ft from the panel center. A gradual decrease occurs
thereafter. The trends hi vertical and horizontal strain are similar to Case 1. However,
increasing the subsidence factor increased the peak vertical tension over the mined panel but
decreased the peak horizontal tension outside the mined region. The secondary peaks outside
the mined region changed very little.
Since the horizontal tensile strain did not decay with distance as much as expected (see
Figure 9-12), the strain analysis was repeated with a larger mesh 9,000 ft by 13,500 ft. As
shown in Figure 9-13, with the larger mesh, the horizontal tensile strain decayed to 7 micro-
in/in at 7,025 feet from the panel center.
Figure 9-14 shows the change in vertical fracture aperture (opening or closing) in the
Culebra formation as a function of distance from the panel center for three assumed joint
spacings (3, 30 and 300 niches). Vertical joint opening which occurs outside the mined
panel ranges from nil to about 0.015 niches which is a smaller range than in Case 1 because
of the smaller peak horizontal tensile strain.
Figure 9-15 shows the results for horizontal joints which respond to vertical strain. The peak
aperture changes at a 300 inch joint spacing are cut off hi the plot. Horizontal joint opening
which occurs above the mined panel ranges from nil to well over 0.04 niches.
9-56
-------
£
O
s
(Thousands)
DISTANCE (ft)
a 10-3
(Thousands)
DISTANCE ft)
A 10-3
(Thousands}
DISTANCE (ft)
A 10-3
Figure 9-11. Relative Change in Fracture Hydraulic Conductivity for Horizontal Joints
with various Fracture Spacings, Subsidence Factor of 52.5%, and Fracture Apertures of
104, lO'3, and 10'2 in. a) 3 inch b) 30 inch c) 300 inch
9-57
-------
0.08
' 0
2 4
(Thousands)
DISTANCE FROM PANEL CENTER (fl)
O E-H + E-V
Figure 9-12. Subsidence-induced Culebra Strains for Subsidence Factor of 67.5%
(Panel extends +/-1,500 ft from origin)
K
>-"
z
-0.04
468
(Thousands)
DISTANCE FROM PANEL CEhTTER (ft)
0 E-H + E-V
Figure 9-13. Subsidence-induced Culebra Strains for Subsidence Factor of 67.5%.
(Panel extends +/-1,500 ft from origin.) Horizontal Mesh Extended to 13,500 ft.
9-58
-------
(V-joint, + opening, 3 spacings)
I
u
UJ
en
k
0.04
0.03 -
0.02 -
0.01 -
0 -f
-0.01 -1
-0.02
-0.03 -
-0.04 -
-0.05 -
-0.06
(Thousands)
DISTANCE (ft)
Figure 9-14. Aperture Change in Vertical Joints for Fracture Spacings of
3, 30, and 300 inches and Subsidence Factor of 67.5%
0.04
-0.06
V-3
(Thousands)
DISTANCE (ft)
V-30
V-300
Figure 9-15. Aperture Change in Horizontal Joints for Fracture Spacings
of 3, 30, and 300 inches and a Subsidence Factor of 67.5%
9-59
-------
Comparison with Case 1, at a 30 inch joint spacing, shows greater horizontal joint opening in
this case (somewhat more than 0.02 inches compared with somewhat less than 0.02 inches in
Case 1).
Figure 9-16 shows the relative increase in hydraulic conductivity of vertical fractures, spaced
3 inches, that is induced by horizontal tensile strain outside the mined panel. The gap in the
plot occurs as the horizontal strain outside the panel changes from tension to compression
and then back to tension with distance from the panel edge. The magnitudes of the relative
change in hydraulic conductivity of the joints are similar to the previous case. Figures 9-16b
and 9-16c present similar results at joint spacings of 30 and 300 inches. As before, the
relative increase is greater for smaller, more widely spaced joints or fractures. Relative
fracture conductivity changes for horizontal joints are included in Figures 9-17a, b, and c.
Conclusion
Simulation of full extraction mining of 10 ft of potash at a depth of about 1,500 ft near the
WIPP shows large vertical tensile strains over the mined panel and slowly decreasing
horizontal tensile strains beyond the panel edge. Although generally in the elastic range, the
strains, when integrated between assumed fractures, lead to displacements that are significant
relative to existing fracture apertures.
9.4.5 Consideration of Other Mining Impacts
In addition to subsidence-induced increased hydraulic conductivity of the Culebra, several
other potentially detrimental scenarios were postulated in Section 9.4.2 above. These are
discussed in the context of the information presented here.
9.4.5.1 Solution Mining
As described earlier, solution mining of langbeinite is not technically feasible because the
evaporite minerals which surround the ore are more soluble than the ore itself. Attempts to
solution mine sylvite have not met with success because of the characteristics of the ore
body. Thus, it appears unlikely that this technique will be applied to potash ores in the
region around the WIPP.
9-60
-------
I
I
1
A
^Thousands}
DISTANCE (ft)
a 10-3
o 10-4
(Thousands)
DISTANCE (ft)
A 10-3
-
t
(Thousands)
DISTANCE (ft)
A 10-3
Figure 9-16. Relative Change in Fracture Hydraulic Conductivity for Vertical Joints
with various Fracture Spacings, Subsidence Factor of 67.5%, and Fracture Apertures
of 104, lO'3, and 10'2 in. a) 3 inch b) 30 inch c) 300 inch
9-61
-------
j"X
(Thousands)
DISTANCE (ft)
A 10-3
(Thousands)
DISTANCE (ft)
A 10-3
(Thousands)
DISTANCE (It)
10-3
Figure 9-17. Relative Change in Fracture Hydraulic Conductivity for Horizontal Joints
with various Fracture Spacings, Subsidence Factor of 67.5%, and Fracture Apertures
of IQ-4, ID'3, and 10'2 in. a) 3 inch b) 30 inch c) 300 inch
9-62
-------
9.4.5.2 Change in Flow Direction of Water-Bearing Members if a Vertical Hydraulic
Connection Is Created By Subsidence
As discussed in the Section 9.4.2, if a hydraulic connection did occur, the result could be a
shifting of flow in the Culebra toward the southwest. According to information presented by
Reeves et al. (REV91), the current travel path is toward the southeast and entails a distance
about 3,600 m from the center of the waste area to the southern boundary of the withdrawn
area. If subsidence produced a hydraulic connection between the water-bearing members of
the Rustler Formation and flow shifted toward the southwest, then the travel distance could
be shortened to 2,415 m which is the shortest distance from the southernmost panel in the
waste area to the southern boundary of the land withdrawal area. This would represent a
33% decrease in travel distance to the accessible environment. However, this shift would
also move the contaminant travel paths into zones of lower hydraulic conductivities which
would result in longer travel tunes to the accessible environment (REV91).
9.4.5.3 Formation of Subsidence-Related Surface Depressions Where Water Could
Accumulate and Alter Local recharge Characteristics
As noted in Section 9.4.3.2, the maximum observed surface subsidence over existing potash
mines in the area is 1.5 m. Using what are believed to be conservative factors (from
Sections 9.4.3.1 and 9.4.3.2) in equation 1, including an extraction ratio 90%, a mine height
of 2.6 m (8.5 ft), and a subsidence factor of 0.67, the calculated surface subsidence would be
1.6 m. Subsidence of this order is less than the quoted surface relief in the area of 3 meters.
Thus, topographical depressions where significant surface water could accumulate and alter
local recharge are not likely.
9.4.5.4 Increased Hydraulic Gradient If Significant Flow From Water-Bearing Strata into
Mine Workings Occurs
Flow of water from the Culebra and Magenta Members of the Rustler Formation into open
shafts has been observed for all four shafts at the WIPP site (CAU90). Leakage into shafts
for various area potash mines has also been reported (CAU90). Quoted leakage values for
the open WIPP shafts are:
• construction and salt handling shaft 0.019 to 0.11 1/s (599.0-3469.0 m3/yr)
• waste handling shaft 0.019 to 0.038 1/s (599.0-1198.0 m3/yr)
• exhaust shaft - 0.026 to 0.030 1/s (820.0-946.0 m3/yr)
• air-intake shaft - 0.030 to 0.056 1/s (946.0-1766.0 m3/yr)
9-63
-------
Flows of this magnitude would not persist if the shafts can be adequately sealed after mining
operations have ceased or once the formation is dewatered. The Bureau of Land
Management does not currently have in place specific regulations for sealing abandoned mine
shafts in the KPLA. Rather, abandonment procedures are initiated by the mining companies
and the sealing plans are developed on a case by case basis with the BLM (GRI96, CRA96).
For example, a current operation involves local removal of the shaft liner and replacement
with a concrete plug which extends from the top of the Salado Formation to the bottom of
the Culebra Member of the Rustler Formation. The plug is 16 to 30 feet in length. The
water bearing formations above the plug will be sealed by grouting to prevent the buildup of
water on the top of the plug. Procedures for future sealing operations may be different.
There is no available evidence as to the longevity of these types of seals. It is reasonable to
assume, however, that even degraded seals would somewhat impede flow into the shafts.
Since it is not clear that currently contemplated shaft seals will prevent leakage for long
periods, it is necessary to consider the impacts of leakage on hydraulic gradients and travel
times to the site boundary. To investigate the potential impacts that mining operations may
have on ground water gradients and subsequent contaminant migration rates, a two-
dimensional modeling analysis was performed. The analysis assumes that the system is
confined and under steady-state conditions. The model also assumes that all groundwater
flow is horizontal and occurs within the matrix (i.e, unfractured flow). The Culebra is
represented as a homogeneous and isotropic porous medium at a constant thickness of 7.7 m,
and an effective porosity of 13.9 percent. A series of simulations were performed in which
the hydraulic conductivities (K) were varied from 7 to 500 m/yr to reflect their potential
impact on altering contaminant migration rates (Table 9-6).
Since the rate at which radionuclides are transported by groundwater is directly proportional
to the hydraulic gradient, any perturbances to the gradient will have a commensurate effect
on migration rates. Furthermore, depending upon location, the presence of mining shafts in
the vicinity of WIPP could have either a beneficial or detrimental effect on the performance
assessment. Shafts located upgradient from a hypothetical human intrusion (i.e., borehole)
would tend to lower or even possibly reverse the hydraulic gradients, thus, reducing the
contaminant velocities and subsequent radionuclide releases at the WIPP land withdrawal
boundary. Alternatively, shafts located downgradient from an intrusion would result in
increased gradients towards the shaft which would tend to accelerate groundwater velocities.
9-64
-------
Table 9-6. Summary of Results for Mine Shaft Leakage Scenario
Hydraulic
Conductivity
(m/y)
7
20
50
500
Time to Travel
2 ion without
Shaft (yr)
12400.0
4350.0
1750.0
175.0
Ground-water
Velocity without
Shaft (m/y)
0.16
0.459
1.14
11.4
Time to Travel
2 tan with
Shaft (yr)
9800.0
3430.0
1370.0
137.0
Groundwater
Velocity with
Shaft (m/y)
0.2
0.583
1.45
14.45
Flow rate
into Shaft
(m3/y)
161.0
460.0
1150.0
11500.0
The shortest distance from the southernmost panel of the WIPP repository to the WIPP land
withdrawal boundary is due south, approximately 2400 m. The ambient groundwater
gradient also has a strong southerly component. Therefore, for this modeling exercise, the
hypothetical mine shaft is located 2000 m downgradient from the waste disposal area. This
distance was selected to maximize the effects that would occur if the groundwater gradients
were affected by mining; in that the mine shaft is not so far away as to have little effect on
flow, yet it is not so close as to create a zone of influence in which contaminants flowing
past the mine shaft would actually travel slower due to the diminishing gradient effects that
will occur downgradient of the mine shaft.
To maximize the effect that the mine could have on the hydraulic gradients, the drawdown at
the mine was set almost at the base of the Culebra at 7.7 m, leaving a seepage face of 0.1 m
at the shaft. The flow rate due to this drawdown was then computed by the model (Table 9-
6). For example, where K = 7 m/yr, the calculated flow rate is 161 m3/yr. Because the
drawdown was maximized, this value represents a reasonable upperbound for the volume of
water that would flow into the open shaft at a Culebra transmissivity (hydraulic conductivity
multiplied by unit thickness) of 53.9 m2/yr.
The hydraulic gradient was also computed at selected points upgradient of the hypothesized
mine shaft and compared to the ambient gradient of 0.0032 under current non-mining
conditions. Since the functional relationship between drawdown and transmissivity can be
linearly extrapolated to any value of hydraulic conductivity, the overall effect on gradients
that is imposed by varying hydraulic conductivities is virtually identical. To illustrate this
relationship, the ratio of the gradient under mining conditions to the original gradient of
0.0032 was computed and is shown on Figure 9-18.
9-65
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V
3
eg
CD
n
a
I
O
aJ
fc
0
0
Change in Culebra Gradient Due to Mining
Original Gradient
1000 2000 3000 4000 5000
Distance Upgradient from. Shaft (m.)
6000
Figure 9-18. The ratio of the hydraulic gradient imposed by mining to the ambient gradient.
9-66
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To investigate the effect that this change in gradient would have on migration rates, a
particle-tracking analysis was performed. This type of analysis moves a particle at the same
velocity as the groundwater and the rate is not affected by dispersion, diffusion, or
retardation. The results from this travel time analysis are presented in Table 9-6.
In each case, the contaminant would travel approximately 27% faster over the 2000 m
distance when the hydraulic gradient is affected by a mine shaft placed in a location chosen
to represent mining's maximum expected effect on the hydraulic gradients. The increase for
each of the simulations 27 % above velocities calculated at ambient gradients and is shown in
column 5 of Table 9-6. This increase is small when compared to changes in velocity due to
potential increases in the hydraulic conductivity. As required by the rule, hydraulic
conductivities will be increased by up to a 1000 fold above their current measured values.
To place this travel tune change caused by the mine shaft in perspective, groundwater
velocities for each of the simulations have been recalculated using hydraulic conductivities
that range from 2 to 1000 times then" original values and are shown in Table 9-7. In each
example, the lowest values for the recalculated velocities fall well above the velocity values
that are increased by 27% due to the change in gradient. This velocity comparison indicates
that increases in hydraulic conductivity over the range specified by EPA have far greater
potential effects on groundwater velocities than increases in the velocities caused by altered
hydraulic gradients due to mine shaft leakage. In light of the EPA requirement that DOE
perform analyses that are more stringent in evaluating mining effects than those associated
with an increase hi gradient, it is reasonable to assume that the consequences a 27 % decrease
in travel tune will have on the overall performance assessment will be captured by those
additional analyses.
Table 9-7. Groundwater Velocities at Hydraulic Conductives the Range
from 2-1000 times those values presented in Table 9-6.
:Hy4raulic Conductivity
! ::T (m/y)
Groundwater Velocity
14.0-7000
40.0-20,000
100.0-50,000
1000.0-500,000
0.32-161.0
0.92^60.0
2.30-1151.0
23.0-11510.0
9-67
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9.4.5.5 Damage to Borehole or Shaft Seals by Subsidence
From the information supporting Figure 9-4, it can be shown that the closest approach of the
sylvite reserves (a grade x thickness product of 40) to waste shaft is slightly over 2,500 feet.
The top of this sylvite ore zone (the 10th) lies about 1,900 feet below the surface (GRI95).
Based on a 45° angle of draw, the impacted area from mining the BLM lease grade reserves
would be about 600 feet from the waste shaft at the surface and at proportionately greater
distances below the surface where maintenance of the shaft seal is more important (e.g.,
through the Rustler Formation). Alternatively, if one assumed the most pessimistic angle of
draw (58°) reported for the area in ITC94, the maximum extent of the impacted area would
be 3,040 feet and the disturbed zone would intersect the waste shaft at about 340 feet below
the surface. The juncture is still some 200 feet above the top of the Rustler Formation and
thus shaft seals should not be affected at any critical location in transmissive members of this
formation. .
If all the BLM lease grade reserves within the repository were mined out, a number of
boreholes would be undercut by the mining operations and the sealed area of the borehole
subject to subsidence-induced strains. Some of these impacted boreholes are shown in Figure
9-4. However, the borehole seals between the repository and the mine workings should not
be affected by the mining operations14.
Thus, it is not expected that mining would breach shaft seals at any critical point along the
sealed length and would not affect borehole seals between the repository horizon and the
mine workings (about 430 feet). Consequently, pathways would not be opened to the
repository by a mining related seal failure mechanism which would facilitate release of
radionuclides.
9.4.5.6 Increased Hydraulic Conductivity of the Salado Formation Due to Excavation-
Induced Stresses
As discussed in Section 9.4.3.1, the maximum increase in the hydraulic conductivity of the
Salado Formation due to stress redistribution around underground openings is expected to be
about an order of magnitude and this altered conductivity decreases rapidly as one moves
away from the mined opening. At a distance equal to six times the width of the opening, the
14 Inside the withdrawn area, only four boreholes associated with the WIPP Project (WIPP 12, 13, DOE 1,
and ERDA 9) and two earlier oil and gas holes reached or exceeded the depth of the repository.
9-68
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altered conductivity is only twice that of undisturbed salt. Even with changes of this
magnitude, the salt would remain highly impermeable. In addition, creep should cause the
salt to revert to near the undisturbed state.
9.4.6 Summary
Extensive potash mining operations are being conducted in the vicinity of the WIPP site with
current mine workings less than 1.5 miles from the site boundary (DOE95). Existing potash
leases abut the site boundary around much of its perimeter (SIL94) and its is expected that
current mining operations will be extended to the land withdrawal boundary.
Reserves and resources of both sylvite and langbeinite exist within the land withdrawal
boundary. Based on current BLM lease grade standards (four feet of 4% K2O for langbeinite
and four feet of 10% K2O for sylvite), the langbeinite reserves are within 3,490 feet of the
waste repository footprint and sylvite reserves are within 1,330 feet of the footprint. These
reserves cannot be exploited currently because the WIPP LWA prohibits mining within the
withdrawn area.
At some time in the future, after active institutional controls are no longer practicable and, if
passive institutional controls have failed to warn about the buried hazards, it is a conceptual
possibility that mining of the ore remaining within the withdrawn area could occur. Such a
hypothetical mining operation would probably require development of a new infrastructure
since existing reserves outside the withdrawn area would likely have been depleted prior to
the failure of institutional controls.
The most likely detrimental impact of such future mining would be increased hydraulic
conductivity of the Culebra Member of the Rustler Formation resulting from subsidence-
induced fracturing of the relatively brittle dolomite. This fracturing (or widening of existing
fractures) could reduce the lateral transit time for radionuclides to the accessible
environment. The increased hydraulic conductivity is of no consequence unless a hydraulic
connection exists between the Culebra and the repository 1,440 feet below. Based on current
WIPP scenarios, the hydraulic connection could be created by an inadvertently drilled
borehole which intersected the repository. Thus performance assessment will need to address
the probability and consequence of such a combination of events. Based on studies reviewed
here, it does not appear that other mining-related scenarios will have significant detrimental
effects on the natural and man-made barriers protecting the repository.
9-69
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Simulation of full extraction mining of 10 ft of potash at a depth of about 1,500 ft near the
WIPP shows vertical tensile strains over the mined panel and slowly decreasing horizontal
tensile strains beyond the panel edge. Although generally in the elastic range, the strains,
when integrated between assumed fractures, lead to displacements that are significant relative
to reasonable fracture apertures. This, in turn, can increase the fracture hydraulic
conductivity of the Culebra.
9.5 REFERENCES
AND78 Anderson, R.Y., "Report to Sandia Laboratories on Deep Dissolution of Salt,
Northern Delaware Basin, New Mexico, January 1978.
AXN94 Axness, C. et al., "Non-Salado Flow and Transport Position Paper," (Revision 1)
Sandia National Laboratories, December 15, 1994.
BAI90 Memorandum from J. Bailey, Certified Professional Geologist, New Mexico
State Land Office to M. La Venue, Intera Inc., August 30, 1990.
BAI94 Bai, M. and D. Elsworth, "Modeling of Subsidence and Stress-Dependent
Hydraulic Conductivity for Intact and Fractured Porous Media," in Rock
Mechanics and Rock Engineering, v. 27, no. 4, pp. 209-234, 1994.
BAR93 Barker, J.T. and G. S. Austin, "Economic Geology of the Carlsbad Potash
District, New Mexico," in New Mexico Geological Society Guidebook, 44th
Field Conference. New Mexico and West Texas, pp 283-291, 1993.
BER94 Berglund, J.W., "Direct Releases Due to Drilling," presentation to EPA,
February 23, 1994.
BER95 Berglund, J.W., "Theoretical Reference Manual for CUTTINGS_S: A Code for
Computing the Quantity of Wastes Brought to the Ground Surface when a Waste
Disposal Room of a Radioactive Waste repository is Inadvertently Penetrated by
an Exploratory Borehole," SAND95-XXXX (Draft), New Mexico Engineering
Research Institute, 1995.
BOM94c U.S. Bureau of Mines, "Mineral Commodity Summaries 1994 (Lithium),"
MINES Faxback Document No. 450394, Ober, Joyce A., 1994.
BLM93 Bureau of Land Management, "Preliminary Map Showing Distribution of Potash
Resources", Carlsbad Mining District, Eddy & Lea Counties, New Mexico,
Roswell District, 1993.
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B0092 Booth, C. J., "Hydrologic Impacts of Underground (Longwall) Mining in the
Illinois Basin," in Proceedings of Third Workshop on Surface Subsidence Due to
Underground Mining, pp 222-227, 1992.
CAU90 Cauffman, T.L., A.M. LaVenue, and J.P. McCord, "Ground-Water Flow
Modeling of the Culebra Dolomite, Volume II: Data Base," Intera, Inc.,
SAND89-7068, October 1990.
CHA84 Chaturvedi, L., "Occurrences of Gases in the Salado Formation," EEG-25, New
Mexico Environmental Evaluation Group, March 1984.
CHE78 Cheeseman, R.J., "Geology and Oil/Potash Resources of Delaware Basin, Eddy
and Lea Counties, New Mexico," in New Mexico State Bureau of Mines and
Mineral Resources Circular 159, 1978.
CLA74 Claiborne, H.C. and F. Gera, "Potential Containment Failure Mechanisms and
Their Consequences at a Radioactive Waste Repository in Bedded Salt in New
Mexico," ORNL-TM-4639, Oak Ridge National Laboratory, October 1974.
CON95 Letter to George Griswold from Leslie M. Cone, District Manager, Roswell
District Office, Bureau of Land Management, October 12, 1995.
C0071 Cooper, J.B. and V.M. Glanzman, "Geohydrology of Project Gnome Site, Eddy
County, New Mexico," U.S. Geological Survey, Professional Paper 712-A, 245
p., 1971.
CRA96 Private communication from Cranston, C., Bureau of Land Management,
Carlsbad NM to J. Channell, S. Cohen and Associates, Inc., January 17, 1996.
DAP82 "Natural Resources Study - Waste Isolation Pilot Plant (WIPP) Project,
Southeastern New Mexico," Draft Report, D'Appolonia, Project No. NM78-648-
813A, January 1982.
DOE80 Department of Energy, "Final Environmental Impact Statement - Waste Isolation
Pilot Plant," U.S. DOE/EIS-0026, October 1980.
DOE93 U.S. Department of Energy, "Implementation of the Resource Disincentive in 40
CFR Part 191.14(e) at the Waste Isolation Pilot Plant," DOE/WIPP 91-029,
Revision 1, June 1993.
DOE95 U.S. Department of Energy, "Draft Title 40 CFR 191 Compliance Certification
for the Waste Isolation Pilot Plant," Draft-DOE/CAO-2056, March 31, 1995.
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DOI89 U.S. Department of the Interior, "Inspection and Enforcement Program and
Selected Related Activities, Bureau of Land Management, Report No. 90-18," C-
LM-BLM-26-89, Office of Inspector General, November 1989.
DOI92 U.S. Department of the Interior, "Follow-up of Recommendations Concerning
Inspection and Enforcement Program and Selected Related Activities, Bureau of
Land Management, Report No. 92-1-578," C-IN-BLM-007-91, Office of the
Inspector General, March 1992.
DUP94 Dupree, J.A. and R.W. Eveleth, "New Mexico Annual Report 1992," U.S.
Bureau of Mines, February 1994.
ELS95 Elsworth, D. and J. Liu, "Topographic Influence of Longwall Mining on
Ground-Water Supplies," in Ground Water, vol. 33, no. 5, pp. 786-793,
September-October 1995.
GAL82 T.N. Narasimhan, "Assessing the permeability characteristics of fractured rock,"
Gale, I.E., in Recent Trends in Hydrogeology, Geological Society of America,
Special Paper 189, ed., p. 163-181, 1982.
GRI95 Griswold, G.B. "Method of Potash Reserve Evaluation," in NMB95
GRI96 Private communication from Griswold, George B. to J. Channell, S. Cohen and
Associates Inc., January 17, 1996.
GUZ91a Guzowski, R.V., "Evaluation of the Applicability of Probability Techniques to
Determining the Probability of Occurrence of Potentially Disruptive Intrusive
Events at the Waste Isolation Pilot Plant," SAND90-7100, Sandia National
Laboratories, December 1991.
GUZ91b Guzowski, R.V. and M.M. Grubel, "Background Information Presented to the
Expert Panel on Inadvertent Human Intrusion into the Waste Isolation Pilot
Plant," SAND91-0928, editors, Sandia National Laboratories, December 1991.
HAR93 "Letter to Sandia National Laboratories Re: Sandia Report and WIPP," Hartman,
Doyle, November 22, 1993.
HEN51 Hendrickson, G.E. and R.S. Jones, "Geology and Ground-Water Resources of
Eddy County, New Mexico," New Mexico Institute of Mining & Technology,
State Bureau of Mines & Mineral Resources, Ground Water Report 3, 121 p.,
1951.
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HIL84 Hills, J.M., "Sedimentation, Tectonism, and Hydrocarbon Generation in
Delaware Basin, West Texas and Southeastern New Mexico," in American
Association of Petroleum Geologists Bulletin, v. 68, No. 3, p. 250-267, March
1984.
HUN89 Hunter, R.L., "Events and Processes for Constructing Scenarios for the Release
of Transuranic Waste from the Waste Isolation Pilot Plant, Southeastern New
Mexico," Sandia National Laboratories, SAND89-2546, December 1989.
ITC94 IT Corporation, "Backfill Engineering Analysis Report," August 1994.
JON78 Jones, C.L., "Test Drilling for Potash Resources: Waste Isolation Pilot Plant,
Eddy County, New Mexico," U.S. Geological Survey, Open File Report 78-592,
2 Vols, 1978.
KEL91 Kelleher, J.T., et al., "Overburden Deformation and Hydrologic Changes Due to
Longwall Coal Mine Subsidence on the Illinois Basin," in Proceedings of Fourth
International Symposium on Land Subsidence, pp. 195-204, May 1991.
KIN42 King, P.B., "Permian of West Texas and Southeastern New Mexico," in
American Association of Petroleum Geologists Bulletin, v. 26, No. 4, April
1942.
KIR94 Van Kirk, C.W., "Report Concerning Salt Water Blow-out January 1991 on the
'Bates Lease' Sections 10 and 15, Township 26 South, Range 37 East,
N.M.D.M., Lea County, New Mexico," Colorado School of Mines, September
16, 1994.
LAM78 Lama, R.D. and V.S. Vutukuri, Handbook on Mechanical Properties of Rocks,
Vol. II, Trans Tech Publications, pp 342-344, 411, 1978.
LAV91 "Anomalous Culebra Water-Level Rises Near the WTPP Site," Memorandum
from Marsh La Venue to Distribution, Sandia National Laboratories, January 28,
1991.
MAT92 Matetic, R.J. and M.A. Trevits, "Hydrologic Variations Due to Longwall
Mining," in Proceedings of Third Workshop on Surface Subsidence Due to
Underground Mining, pp 204-213, 1992.
NIC61 Nicholson, A. Jr. and Alfred Clebsch, Jr., "Geology and Ground-Water
Conditions in Southern Lea County, New Mexico," New Mexico Institute of
Mining and Technology, State Bureau of Mines and Mineral Resources, Ground
Water Report 6, 123 p., 1961.
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NMB95 New Mexico Bureau of Mines and Mineral Resources, "Evaluation of Mineral
Resources at the Waste Isolation Pilot Plant (WIPP) Site," Volume 1, Executive
Summary, NM.BMMR-95*1020, March 31, 1995.
OCC88 Oil Conservation Commission, "Case No. 9318, Order No. R-lll-P," Energy,
Minerals and Natural Resources Dept., State of New Mexico, April 21, 1988.
OCD93 "State of New Mexico Energy and Minerals Department Oil Conservation
Division Rules and Regulations," March 1993.
OCD94 "Private Communication from M. Ashley," New Mexico Oil Conservation
Division, Artesia NM, March 2, 1994.
PAR78 Pariseau, W.G. "Interpretation of Rock Mechanics Data", Vol. I & II, U.S.
Bureau of Mines Contract Report, Contract No. H0220077.
PAR91 Pariseau, W.G. "Using Utah2/PC", U.S. Bureau of Mines U of Utah
Workshop on Personal Computer Finite Element Modeling for Ground Control
and Mine Design Notes, Denver, Colorado.
PIC94 Pickens, J.F., "Expert Witness Report by John F Pickens," September 28, 1994.
POP83 Popielak, R.S. et al., "Brine Reservoirs in the Castile Formation, Waste Isolation
Pilot Plant (WIPP) Project, Southeastern New Mexico," D'Appolonia Consulting
Engineers, Inc. for U.S. Department of Energy, TME 3153, 1983.
POW78 Powers, D.W., et al. "Geological Characterization Report, Waste Isolation Pilot
Plant (WIPP) Site, Southeastern New Mexico," ed. Sandia National Laboratories,
SAND78-1596, August 1978.
RAM76 "Water Flows In and Near Waterflood Projects In Lea County," Memorandum
from Ramey, J.D., New Mexico Oil Conservation Division, to John F. O'Leary,
May 5, 1976.
REV91 Reeves, M. et al., "Regional Double-Porosity Solute Transport in the Culebra
Dolomite Under Brine-Reservoir-Breach release Conditions: An Analysis of
Parameter Sensitivity and Importance," Sandia National Laboratories, SAND89-
7069, 1991.
RIC85 Richey, S.F., J.G. Wells and K.T. Stephens, "Geohydrology of the Delaware
Basin and Vicinity, Texas and New Mexico," U.S. Geological Survey, Water-
Resources Investigations Report 84-4077, 1985.
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SAN92 Sandia National Laboratories, "Preliminary Performance Assessment for the
Waste Isolation Pilot Plant, December 1992," SAND92-0700 (Vol. 3), December
1992.
SCA95 S. Cohen and Associates Inc., "Groundwater Flow and Contaminant Transport
Modeling at WIPP," Work Assignment 3-20, November 1995.
SEA94 "Private Communication From James Searls," U.S. Bureau of Mines,
Washington, D.C., January 1994.
SEE78 Seedorff, W.A., "Resource Study for the Waste Isolation Pilot Plant Site, Eddy
County, New Mexico," AIM Inc., August 15, 1978.
SIL94 Silva, M.K., "Implications of the Presence of Petroleum Resources on the
Integrity of the WIPP Site," EEG-55, New Mexico Environmental Evaluation
Group, June 1994.
SMI73 Smith, G.I.,et al., "Evaporites and Brines," in United States Mineral Resources.
Donald A. Brobst and Walden P. Pratt, ed., U.S. Geological Survey Professional
Paper 820, 1973.
SN069 Snow, D.T. "Anisotropic permeability of fractured media," in Water Resources
Research, vol. 5, no. 6, p. 1273-1289, 1969.
USG88 U.S. Geological Survey, "Ground-Water Hydrology of Marshall County, West
Virginia, with Emphasis on the Effects of Longwall Coal Mining," Water
Resources Investigations Report 88-4006, 1988.
WAR90 Warner, D.L. and C.L. McConnell, "Abandoned Oil and Gas Wells - A
Quantitative Assessment of Their Environmental Impact," University of Missouri
- Rolla, under contract to American Petroleum Institute, June 1990.
WEI79 Weisner, R.C., et al., "Valuation of Potash Occurrences Within the Nuclear
Waste Isolation Pilot Plant Site in Southeastern New Mexico," U.S. Bureau of
Mines Information Circular 8814, 1979.
WEI77 Weisner, R.C. et al., "Valuation of Potash Occurrences Within the Nuclear
Waste Isolation Pilot Plant Site in Southeastern New Mexico," U.S. Bureau of
Mines Information Circular 8814, 1977.
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10. Active Institutional Controls
10.1 INTRODUCTION
10.1.1 Regulatory Requirements Relevant to Institutional Controls at WIPP
In recognizing the many uncertainties inherent in the analyses for the containment criteria, as
established in Subpart B of 40 CFR part 191, EPA developed assurance requirements to
guarantee that the implementing agencies act cautiously and take steps to reduce such
uncertainties. The following six assurance requirements are stipulated in §191.14:
• Active Institutional Controls
• Monitoring
• Passive Institutional Controls
• Barriers
• Resource Disincentives
• Waste Removal
Active institutional controls are defined hi §191.12(f) as:
"Active institutional controls means: (1) Controlling access to a disposal site
by any means other than passive institutional controls; (2) performing
maintenance operations or remedial actions at a site, (3) controlling or cleaning
up releases from a site, or (4) monitoring parameters related to disposal
system performance."
Active institutional controls operate sequentially in conjunction with passive institutional
controls to protect and mark the WIPP site. Passive institutional controls are defined in
§191.12(e) as:
"Passive institutional controls means: (1) Permanent markers placed at a
disposal site, (2) public records and archives, (3) government ownership and
regulations regarding land or resource use, and (4) other methods of
preserving knowledge about the location, design, and contents of a disposal
system."
Active institutional controls are applied after "disposal," (i.e., after all shafts to the
repository are backfilled and sealed (§191.02(1)), when the waste has been permanently
isolated with no intent of recovery. According to §191.14(a), "active institutional controls
10-1
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over disposal sites should be maintained for as long a period of time as is practicable after
disposal; however, performance assessments that assess isolation of the wastes from the
accessible environment shall not consider any contributions from active institutional controls
for more than 100 years after disposal." In Appendix C of 40 CFR part 191, guidance is
provided for implementation of institutional controls that states "the implementing agency
will assume that none of the active institutional controls prevent or reduce radionuclide
releases for more than 100 years after disposal. However, the Federal Government is
committed to retain ownership of all disposal sites for spent nuclear fuel and high-level and
transuranic wastes and will establish appropriate markers and records, consistent with
Based on the active institutional controls requirement of 40 CFR part 191, EPA included the
following compliance criterion under §194. 41 (a) of the 40 CFR part 194 regulations:
"Any compliance application shall include detailed descriptions of proposed
active institutional controls, the controls' location, and the period of time the
controls are proposed to remain active. Assumptions pertaining to active
institutional controls and their effectiveness in terms of preventing or reducing
radionuclide releases shall be supported by such descriptions."
Examples of active institutional controls employed for the purpose of restricting site access
include (EPA88):
• a [maintained] security fence and other barriers,
• security guards
• routine patrols
• electronic surveillance
Examples of passive institutional controls include signs, markers, deed restrictions, land-use
controls, records, and legal documents. Passive institutional controls should warn those who
attempt to enter the disposal site vicinity of the hazards associated with activities that would
disturb the subsurface. Furthermore, passive institutional controls require comprehensive
actions that will increase the likelihood that knowledge and information about the disposal
site and its contents are passed on to future generations.
10-2
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10.1.2 WIPP Site Characteristics
10.1.2.1 Site Description
The WIPP site is located in Eddy County, in southeastern New Mexico. The site is 26 miles
east of Carlsbad on a relatively flat, sparsely inhabited plateau with little surface water and
limited land uses. The land is primarily used for grazing. Other land uses within five miles
of the WIPP boundary include potash mining and oil and gas exploration and development
(SAN92).
The WIPP is a controlled site of 10,240 acres, which has been withdrawn from all forms of
entry, appropriation, and disposal including, without limitation, mineral leasing laws,
geothermal leasing laws, material sale laws, and mining laws as described in the WIPP
LWA. Areas designated as subdivisions within the WIPP site boundary include Zones I and
II. Zone I is an area of 35 acres surrounded by a chain link fence. Zone I encloses all the
major surface facilities. Zone II overlies the maximum extent of underground development
and encompasses an area of about 277 acres. The WIPP site boundary provides a minimum
of a one-mile wide buffer area of intact salt beyond Zone II (DOE93).
10.1.2.2 WIPP Facilities
The WIPP site contains surface and underground facilities interconnected by four shafts.
The surface structures accommodate the personnel, equipment, and support services required
for the receipt, preparation, and transfer of transuranic radioactive waste from the surface to
the underground. The underground facility is constructed in a bedded salt formation 2,150
feet (655 m) below the surface. Existing underground facilities include the TRU waste
disposal area, the experimental area, and the underground maintenance and support area
(SAN92).
10.1.2.3 Waste Characteristics
DOE will use the WIPP to receive and dispose of TRU waste. TRU waste are those wastes
containing radioactive elements with an atomic number greater than 92, a half-life greater
than 20 years, and a concentration greater than 100 nanocuries per gram, excluding high-
level waste and/other specific waste types. Some of these wastes are co-contaminated with
10-3
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hazardous constituents, making them mixed wastes. The wastes will be shipped in specially
designed transportation containers and will be packaged in 55-gallon drums and/or standard
waste boxes.
10.1.2.4 Operations
Following receipt and inspection, the waste containers, will be downloaded into the
subsurface repository. Ultimately this repository will consist of eight "panels," each of
which will contain seven separate disposal "rooms" and interconnecting drifts. After an
entire panel is filled, it may be closed to isolate it from the rest of the repository.
DOE expects that waste emplacement will begin in 1998 and continue for a 25-year period
until the regulated capacity of the repository of 6,200,000 ft3 of TRU waste has been
reached. This capacity restriction must also include TRU waste derived from any
decontamination activities during the disposal phase and decommissioning.
10.1.2.5 Closure/Post-Closure Activities
Current DOE plans indicate that prior to closing the waste disposal area, surface facilities
will be decontaminated. Contaminated material that cannot be sufficiently cleaned to be
released as uncontrolled material will be emplaced within the waste disposal area.
The final activities within the repository will be the closing of the waste disposal area and the
sealing of the shafts. Upon completion of this activity, the remaining surface structures will
be dismantled. All surface structures will be removed, except for the concrete Hot Cell
structure and a sufficient quantity of salt tailings to support construction of the permanent
marker berm. Disturbed land will be regraded and planted to return the site to as near its
original condition as is practicable. At completion of the closure phase, DOE will implement
the WIPP active institutional controls program.
10.2 ACTIVE INSTITUTIONAL CONTROLS PROPOSED FOR THE WIPP SITE
As part of the active institutional controls program, DOE has developed a set of design
criteria that describe how the active institutional controls will be implemented. These criteria
are summarized below:
10-4
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• A fence line shall be established to control access to the repository's footprint
area (the waste disposal area projected to the surface). A standard wire fence
shall be erected along the perimeter of the repository surface footprint. The
fence shall have gates placed approximately midway along each of the four
sides.
• An unpaved roadway along the perimeter of the barbed wire fence shall be
constructed to provide ready vehicle access to any point around the fenced
perimeter, to facilitate inspection and maintenance of the fence line, and to
permit visual observation of the repository footprint to the extent permitted by
the lay of the land. This roadway shall connect to the paved south access
road.
• To ensure visual notification, the fence line shall be posted with signs having,
as a minimum, a legend reading "Danger-Unauthorized Personnel Keep
Out" and a warning against entering the area without specific permission of
DOE, or other local authority such as the Eddy County Sheriffs Office.1
• Contractual arrangements shall be developed to ensure that periodic inspections
and necessary corrective maintenance are conducted on the fence line, its
associated warning signs, and the roadway.
• Through direct DOE staffing support and/or contractual arrangements,
procedures shall be established to provide routine periodic patrols and
surveillance of the protected area by personnel trained in security,
surveillance, and investigation.
• Processes will be developed for monitoring and controlling the long-term
testing requirements of the permanent marker system.
• Processes will be developed for implementing the periodic monitoring
requirements of the disposal system's monitoring program.
• Recommendations will be developed for modifications to the active
institutional controls appropriate for access control and surveillance upon
installation of the permanent marker system.
• Guidelines will be developed for recommending mitigation actions to be taken
to address any abnormal conditions identified during periodic surveillance and
inspections.
1 DOE is suggesting use of the Eddy County Sheriffs Department to conduct periodic surveillance of
WIPP active institutional controls. This surveillance would be conducted pursuant to a contract between the
DOE and the Sheriffs Department.
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• Reports of activities associated with the post-disposal active access controls
shall be prepared in accordance with regulatory requirements for submittal to
the appropriate regulatory and legislative authorities.
Details on meeting these criteria were submitted as, "WIPP Active Access Controls After
Disposal Design Concept Description." Summarized below are additional noteworthy items
delineated in the report.
• Access control. Access to an area approximately 2,780 feet by 2,360 feet will
be controlled by a 4 strand (3 barbed and 1 unbarbed in accordance with the
Bureau of Land Management specifications) wire fence. A single gate will be
placed approximately mid-way along each side of the fence for access. The
western gate shall be 20 feet wide; and the remaining three gates shall each be
16 feet wide. Around the perimeter of the fence, an unpaved roadway 16 feet
wide will be cut to allow for patrolling of the perimeter. Patrolling of the
perimeter is based upon the need to ensure that no mining or well drilling
activity is inadvertently initiated which could threaten the integrity of the
repository.
• Surveillance monitoring. Surveillance monitoring will consist of drive-by
patrolling around the fenced perimeter, two to three times per week. During
the course of the patrol, particular note shall be taken of fence integrity, gate
integrity, and locked conditions of each gate. Surveillance should also include
visual observation of the entire enclosure area for any signs of human activity.
• Maintenance and remedial actions. Anticipated maintenance and remedial
action issues during the active control period are minimal and should
encompass issues such as fence/road maintenance, evidence of vandalism,
potential erection of drilling equipment, grass fires, unauthorized entry in
prohibited areas.
• Control and cleanup of releases. DOE intends to complete the
decontamination process and disposal of derived radioactive waste prior to
final closure of the waste disposal area and sealing of the shafts.
• Long-term monitoring. Details describing the establishment of a network of
elevation benchmarks and the development of a data baseline from which to
evaluate disposal system performance is described in the Long Term
Monitoring Design Concept Description (DOE94). (NOTE: Disposal system
monitoring is addressed in§191.14asa separate assurance requirement;
therefore this topic is discussed in detail in Chapter 11).
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10.3 INSTITUTIONAL CONTROLS AT OTHER FACILITIES
For comparison, a review was conducted of active institutional controls proposed or
implemented at other facilities and their corresponding regulations. (It should be noted that,
although the focus of this chapter is active institutional controls, in practice and in the
regulations there may not always be a clear delineation between active and passive controls.)
POE and Department of Defense (DOD) facilities that contain special nuclear material,
NRC-licensed nuclear reactor facilities, low-level waste disposal facilities, uranium mill
tailings disposal sites, and Superfund sites were examined. This review focused on those
institutional controls specifically designed for protection against human intrusion because they
have the most relevance to the WIPP.
10.3.1 Facilities Containing Special Nuclear Material
A number of DOE and DOD facilities must protect special nuclear material. The access
controls at these facilities represent the extreme end of the controls continuum that could be
considered for application at the WIPP. Typically, these controls include continuous
monitoring by armed guards, double rows of chain link fence topped with barbed wire,
motion detectors, infrared detectors, and visual surveillance using remote TV cameras.
These controls are designed to prevent intentional intrusion into critical areas where the
special nuclear material is stored, and to ensure the material is not stolen or sabotaged.
These controls also prevent inadvertent intrusion. Many of the specific control elements in
place at these facilities resemble the proposed controls for WIPP. For example, the fact that
the TRU waste will be over 2,000 feet below the surface should be at least as effective a
control as the fencing arrangement at DOE special nuclear facilities such as Pantex.
10.3.2 Retired Nuclear Reactor Facilities
When a nuclear reactor has reached the end of its useful life, it must be decommissioned in
accordance with the requirements established in 10 CFR part 50. NRC regulations define
"decommissioning" as the process of reducing residual radioactivity to a level that permits
release of a facility for unrestricted use and termination of an NRC license. In effect, this
definition means that, after the radioactivity exceeding NRC limits for unrestricted use has
been removed, no further institutional or administrative controls are required.
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Licensees may request and have been granted exemptions to the unrestricted use requirement.
One interim decommissioning alternative that has been used by several retired facilities is
termed safe storage (SAFSTOR). Safe storage is defined as those activities required to place
and maintain a nuclear facility in such condition that future risk from the facility to public
safety is within acceptable bounds and that the facility can be safely stored for as long as
desired.
During the SAFSTOR period, irradiated fuel assemblies and in-core fission chambers are
stored in the spent fuel pool. The onsite storage of spent fuel requires the continued
operation of numerous plant systems, such as (1) service systems, (i.e., ventilation, spent
fuel pool service, fire protection, and electrical), (2) waste disposal systems, and (3)
monitoring systems, (i.e., stack gas radiation monitoring systems, process water monitoring,
offsite environmental monitoring stations, etc.).
Active institutional controls at reactor facilities in a safe storage condition are extensive and
are, therefore, not limited to protection against unauthorized entry. A permanent plant staff
for the operation of necessary plant systems, preventative/corrective maintenance of
structures, systems, components, and equipment, and onsite/offsite environmental monitoring
must be maintained during the SAFSTOR period. During SAFSTOR, a licensee is required
to maintain a full-time, onsite security force to prevent unauthorized access or deliberate
intrusion into the facility. Additionally, a system of multiple locked physical barriers and
warning signs/signals must be maintained to control access into areas where exposure to
radiation is possible (NRC94).
10.3.3 Low-Level Radioactive Waste Disposal Facilities
The Nuclear Regulatory Commission established regulations under 10 CFR parL.61 to cover
all phases of land disposal of low-level radioactive waste (LLW). A LLW disposal facility
licensed under 10 CFR part 61 consists of the land, buildings, and equipment required for
the near-surface disposal of LLW. These regulations also require the use of a waste
classification system, where high-activity Class C wastes are to be placed deep in the ground
(at depths below 5 meters) or behind barriers to limit human intrusion.
Six commercially-operated LLW disposal facilities, located at Beatty, Nevada; Maxey Flats,
Kentucky; West Valley, New York; Richland, Washington; Barnwell, South Carolina; and
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Sheffield, Illinois, have been licensed and operated in the United States. These facility were
licensed prior to the promulgation of 10 CFR part 61 and use shallow land burial designs.
Thfe Richland and Barnwell facilities continue to operate as disposal facilities for LLW,
whereas the other four sites have closed.
Under the Low-Level Radioactive Waste Policy Act of 1980 and the Low-Level Radioactive
Waste Policy Amendments Act of 1985, states are responsible for the disposal of commercial
LLW generated within their respective boundaries. Since this legislative directive, several
states and regional compacts are in various stages of planning and licensing new LLW
disposal facilities. All new facilities will be licensed under 10 CFR part 61 or compatible
Agreement State regulations. In addition, Envirocare of Utah, Inc., has applied to NRC for
a license to construct and operate a facility to receive, store, and dispose of uranium and
thorium byproduct material.
Institutional control requirements for LLW land disposal facilities, as cited in §61.59,
specifically address control of access, environmental monitoring, surveillance, minor
custodial care, and administration of funds to cover the costs for these controls. The primary
institutional control to protection against inadvertent intrusion is physical security (e.g.,
barriers, fences) to limit site access. Other active controls include periodic inspection of the
site, maintenance of disposal unit covers, revegetation of the disposal area, and maintenance
of the security fence. For example, the site stabilization and closure plan for the LLW
facility operated by U.S. Ecology Inc. in Richland, Washington, has proposed the following
active institutional controls as part of their Site Stabilization and Closure Plan (USE95):
• At closure, security around the facility will be maintained by the existing 8
foot high galvanized chain-link fence, which is topped with three strands of
barbed wire.2
• Two times each year, during the 5 to 54 year post-closure and maintenance
period, a crew of two men and foreman will spend three days each visit
performing miscellaneous maintenance.
• Annually, during the 55 to 100 year post-closure period, a crew of two men
and a foreman will spend three days performing necessary maintenance. Also
during this period, the fence surrounding the facility will be replaced.
2 This fence is much more robust than that proposed for the WIPP site; however the waste at Richland is
shallow lying and hence more prone to disturbance by surface activities.
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Although 10 CFR part 61 specifies that institutional controls cannot be relied upon for more
than 100 years, some of the new LLW disposal facilities are proposing the use of active
controls for longer than 100 years. For example, a minimum of 100 years of active controls
is proposed for new facilities in California and Nebraska (KAR95); and the license
application for a new facility in Illinois contained a 300-year active institutional control
period (NRC93).
10.3.4 Uranium Mill Tailings Disposal
Uranium mill tailings are a byproduct waste that results from the processing of ore to extract
uranium. Historically, uranium mill tailings have been stored in large surface
impoundments. The principal health concern is exposure to radon-222, a radioactive decay
product of uranium.
Long-term stabilization and disposal regulations were developed under the Uranium Mill
Tailings Radiation Control Act of 1978 (UMTRCA) by EPA and NRC and set forth in 40
CFR part 192. In addition, the NRC developed specific licensing and design criteria, which
are addressed hi 10 CFR part 40, to implement EPA's environmental standards.
In accordance with existing regulations, uranium mill tailings must be stabilized for 1,000
years, to the extent reasonably achievable, and, in any case, for at least 200 years without
active maintenance. Therefore, controls for stabilization and safe isolation of the tailings
primarily rely on site characteristics and engineering designs.
Site closure activities, which are intended to reclaim and stabilize the site to such a degree
that no active, ongoing maintenance is required, typically consist of the following:
• dewatering the tailings ponds,
• implementation of a ground water remediation program,
• filling the impoundment area with a sufficient quantity and type of material to reclaim
and stabilize the site (reduce radon to acceptable levels) in an environmentally sound
manner,
• dismantling, disposing, or salvaging mill site buildings and material,
• decontamination of mill site sofls,
• establishment of an appropriate environmental monitoring program for closure and
post-closure needs.
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Institutional controls to protect against inadvertent intrusion are neither explicitly identified or
designated in the regulations for surface remediation of tailings disposal sites, nor do they
provide definitions of, or specific criteria that distinguish between, active and passive
institutional controls. However, provisions in EPA standards and NRC regulations that
contribute directly and indirectly to intruder prevention and protection include (1) transfer of
ownership and control of the site, to a government agency (usually DOE) for long-term
custody, records control, and deed and land-use restrictions; and (2) periodic site inspection
and surveillance, monitoring, and, if necessary, maintenance during the post-closure period.
10.3.5 Superfund Sites
EPA is responsible for remediation of hazardous releases into the environment under the
Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of
1980. This program, commonly referred to as the Superfund Program, involves the
remediation of more than 1,400 contaminated sites. CERCLA requires that Superfund sites
comply 'with the organizational structure, procedures, and criteria specified in the National
Contingency Plan (NCP), 40 part CFR 30<
By the end of 1993, Superfund remedial actions at 21 radioactively contaminated sites had
been determined. Institutional controls were included as part of the selected remedies for 12
of these sites. However, the institutional controls at eight sites were only selected for an
interim period until the final remedy was implemented, which required development of
off site waste disposal capacity.
Institutional controls selected to support temporary remedies include access control, fencing,
waste storage, surveillance, and monitoring. For three sites with permanent remedies,
institutional controls included access controls (typically fences), deed restrictions., leachate
collection and treatment, groundwater monitoring, drilling and pumping restriction, cover
maintenance, and procedural controls. For example, at the Maxey Flats low-level waste
disposal facility in Kentucky, EPA's final remedial action includes institutional controls to
restrict use of the site and to ensure monitoring and maintenance of the site in perpetuity,
since the radioactive and hazardous waste, once stabilized, will remain onsite.
In terms of institutional controls, the NCP is sufficiently general to allow the use of a wide
range of institutional controls, if necessary, to protect human health and the environment.
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Detailed selection of institutional controls occurs as a part of the remedial design, after the
selection of the remedy in the Superfund Record of Decision. EPA or the state agency
negotiates the type and necessary duration of the institutional controls with responsible parties
and affected interests. Since the majority of selected remedies to date ultimately require the
removal of radioactively contaminated materials, the use of institutional controls are typically
passive, such as deed restrictions and ordinances to limit access or resource use.
10.3.6 Applicability to WIPP
Institutional controls defined by Federal regulations at various other facilities are primarily
affected by post-closure conditions and characteristics and accessibility of the hazardous
materials. As discussed above, active institutional controls for restricting site access can
range from a combination of full-time security guards, visual and electronic surveillance, and
multiple locked barriers at facilities in SAFSTOR to facilities, such as uranium mill tailings
disposal sites, where the closure goal is intended to reduce the necessity for and reliance on
active institutional controls.
Due to profound differences in siting, waste characteristics and form, accessibility of
emplaced hazardous materials, and post-closure conditions between WIPP and the above-cited
facilities, the active institutional controls employed at these sites are not directly applicable.
However, the range of active institutional controls at these sites can serve as a basis for
establishing bounding criteria for the controls required by 40 CFR part 194. It should be
noted that no nuclear facility subject to active institutional controls has been hi existence for
more than about half the 100-year post-closure period allowed for active Institutional control
credit in 40 CFR part 194, thus limiting the experience available to determine the adequacy
of these safeguards. One can cite other governmental institutions where active institutional
controls have been in place for more than 100 years (e.g., Sing Sing Prison), but the
applicability of such experience to a geologic repository is questionable.
10.4 ADEQUACY OF PROPOSED ACTIVE INSTITUTIONAL CONTROLS FOR WIPP
As previously defined, active institutional controls refer to the deliberate actions taken to
restrict access to and use of the site.
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The primary considerations in waste disposal siting strategies and design features are their
effectiveness in isolating the waste and protecting against inadvertent intrusion into the
disposal area. Siting features include selecting a location where the benefits of the site
outweigh the detriments. Detriments could be concerns that population growth could affect
the site or that future exploration for natural resources (e.g. hydrocarbons, minerals, water)
could effect repository performance. Design features consist of using natural and engineered
barriers to isolate the waste and minimize/mitigate the effects of human intrusion.
Massive geologic formations between the waste and the earth's surface are undoubtedly the
primary design feature that limits inadvertent intrusion to buried waste. Perimeter fences,
barriers, warning signs, and controlled use of access roads provide a second level of control
measures to protect against inadvertent intrusions. A third level involves surveillance.
Surveillance may be continuous or periodic, conducted through visual inspection by security
personnel, and supplemented by electronic devices.
The adequacy of institutional control measures must, therefore, be assessed in terms of the
effectiveness by which control measures limit intrusions. The effectiveness of institutional
controls can be assessed for a number of intrusion scenarios; however, the active institutional
controls being considered for the WIPP consider drilling as the only intrusion scenario that
could credibly breach the repository. The section that follows describes alternate intrusion
scenarios, and examines them with respect to: 1) probability of occurrence, 2) consequences,
and 3) effect of additional active institutional controls on the probability of their occurrence.
10.4.1 Inadvertent Intrusion Scenarios
An intruder may encounter a closed waste disposal site and, due to a temporary or permanent
breakdown in institutional controls, engage in a variety of activities. Such unintentional
intrusions may be transient, short-term, or even permanent (NRC84). Potential intrusion
scenarios, their likelihood of occurrence, and potential radiological consequences are
discussed below.
10.4.1.1 Recreational
This scenario encompasses numerous plausible activities involving trespassing of hikers,
campers, off-road vehicle operators, etc.
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• Probability of Occurrence - A perimeter fence and warning signs and periodic
security inspections that verify their integrity minimize the possibility of
inadvertent intrusion. Under unusual conditions in which fence and warning
signs are removed but security inspections are maintained, such intrusions
would be limited to less than a four-day time period.
• Potential Consequence The pre-closure decontamination and removal of
surface structures and restoration of land to pre-operational conditions would
ensure that exposure from contact with existing surface materials, including
soil, is below regulatory limits. Thus, for recreational intrusions that do not
significantly modify the site and are of short duration, potential radiological
consequences are insignificant.
10.4.1.2 Agricultural
In this scenario, the inadvertent intruder is assumed to plant crops on the disposal site for
human or farm-animal consumption.
• Probability of Occurrence - The aforementioned institutional controls at WIPP,
which include construction and maintenance of a perimeter fence, warning
signs, and periodic security inspections, preclude the likelihood of the intruder-
agriculture-scenario. Even the transient loss of these institutional controls for
a period of days to weeks is insufficient to support a growing period of weeks
to months required for agricultural crops.
A significant factor in this scenario is that site characteristics, defined by soil
quality and rainfall, would not support agricultural activities. Thus, the
probability of occurrence for this scenario is insignificant during the period for
which active controls will be hi place.
• Potential Consequence - Although crops cannot be affected by waste disposed
at a depth of more than 2,000 feet below the surface, garden crops and animal
forage become contaminated from radioactivity contained in soil as a result of
root uptake and foliar deposition of resuspended soil particles.
The level of residual soil contamination at the WIPP following closure can be
assumed to comply with current standards that limit soil contamination within
the root zone to 5 pCi/g and to 15 pCi/g below the root zone. These values
are likely to represent bounding values for the agricultural intruder scenario
and would cause only minimal impact with regard to human exposure.
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10.4.1.3 Home Construction
This scenario assumes that an intruder inadvertently proceeds with construction of a home on
the disposal site. Construction includes excavation for concrete footers, basement, utilities,
etc. These typical activities should not be expected to involve depths in excess of 15 feet.
One noteworthy exception, however, is drilling for well water, which is discussed separately
below.
• Probability of Occurrence Full implementation of proposed institutional
controls renders this scenario highly improbable. Only with an extended
breakdown in active institutional controls is it conceivable that construction
could inadvertently progress through the initial phase of home construction that
includes excavation of a basement, septic system, and grading of the
construction site.
• Potential Consequences Disturbance of site surface layers that are assumed
not to exceed 15 pCi/g would expose construction workers to low-level
airborne environments and external radiation during this brief period of
construction. Intruder exposure from these pathways is very low.
10.4.1.4 Groundwater Scenario
There are several potential groundwater scenarios depending upon the intended use of the
site. A well may be drilled on behalf of the agricultural intruder scenario, the home-
construction intruder scenario, or for cattle grazing on the open range.
• Probability of Occurrence - Drilling for water can be expected to occur only if
there is a prolonged breakdown of institutional controls. The concern for well
drilling and mining at the WIPP can be addressed by considering that, using
current drilling technology, it typically requires at least 2-3 days for a driller
to setup a deep drilling rig and commence actual drilling operations. To attain
the 655 meter depth that would approach the repository horizon takes at least
another week to 10 days. Patrolling the fenced area 2-3 times weekly would
identify potential drilling activity well before any breach of the repository
could occur.
Active wells exist hi the Dewey Lake Red Beds 3.2 to 3.4 miles south of the
repository (about 1.2 miles south of the southern boundary of the'Land Withdrawal
Area) (DOE93a).
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Beyond these temporal limitations, the improbability of drilling for water is more
likely to be due to common knowledge among local well drillers that there are no
known potable aquifers in the immediate vicinity of WIPP.
• Potential Consequences Deep well drilling could bring water from the
Culebra Member of the Rustier Formation to the surface. The water may be
contaminated and contain radiologically significant quantities of waste only if
the water well drilling activity has been preceded by an intrusion into the
repository.
According to DOE, "water quality of the Culebra in the vicinity of the WIPP is
naturally poor and the waters are not usable for human consumption or for
agricultural purposes," (DOE93a).
10.4.1.5 Drilling for Hydrocarbons
Exploratory drilling for oil and gas is a common activity around the WIPP site. As domestic
and world oil and gas supplies dwindle over the next 100 years, the incentive for exploratory
drilling may escalate.
• Probability of Occurrence As discussed with respect to the water drilling
scenario above, when proposed institutional controls are maintained, the
likelihood of inadvertent commercial drilling for hydrocarbons must be
considered highly improbable. Only with prolonged or sustained breaks in
institutional controls (i.e., greater than 3-4 weeks) could this scenario progress
sufficiently far to pose a radiological threat. Assuming that "rank wildcat"
exploration is carried out at a rate of about 3 x 10"1 drill holes per square
kilometer per year (TRA91),3 for the 277-acre fenced area at WIPP, this
would imply a probability of 6 x 10"2 that a single bore hole would be drilled
inadvertently into the repository over a 100-year period. This probability is
based on a sustained breakdown of institutional controls for the entire 100-year
period.
• Potential Consequence Potential radiological consequences resulting from
exploratory drilling for hydrocarbons are greater than those previously
identified in the water drilling scenario because the deeper hydrocarbon
boreholes are more likely to intercept the buried waste.
3 SAN92 states that, based on guidance in Appendix B of the Standard, "a maximum of 30 boreholes/km2
were allowed in 10,000 years."
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10.4.2 Intentional Intrusion Scenarios
In this chapter intentional intrusions scenarios refer to activities associated with individuals
who willfully and knowingly violate institutional control efforts. Intentional intrusion
scenarios can be further categorized as benign and hostile.
10.4.2.1 Benign Intentional Intrusion Scenarios
There are numerous scenarios that can be labeled as benign intentional intrusion. The
activities associated with benign intrusions generally do not go beyond willful trespass and,
therefore, do not pose a radiological threat. Intruders in this category are likely to include
tourists, curiosity seekers, souvenir collectors, people intent on mischief/vandalism, etc.
• Probability of Occurrence. Population expansions and encroachment by future
communities at the WIPP site will undoubtedly raise the probability and
frequencies of these intrusions so as to make them commonplace. A sincere
desire to impede this type of willful intrusion would require that the proposed
4-strand wire fence be replaced by a more effective fence (e.g., operating
nuclear power plants employ an eight-foot typhoon fence topped with several
coils of razor wire for primary perimeter protection).
• Potential Consequence - In general, benign willful intrusions are likely to have
no radiological significance. Of potential consequence might be a scenario in
which a souvenir hunter by means of a metal detector finds an accessible
contaminated metal object that had failed detection during pre-closure cleanup
efforts.
10.4.2.2 Hostile Intentional Intrusion Scenarios
This classification of intrusion is defined by activities aimed at accessing disposed waste for
purposes of sabotage and/or terrorism. Although the waste is protected by more than 2,000
feet of overlying geological formation consisting of soil, sand, rock, and salt, only drilling
equipment (as used hi drilling water, gas, or oil wells) is needed to penetrate the repository
horizon. Acts of sabotage or terrorism may involve the introduction of chemical and
physical agents, inclusive of explosives, that would impact containment integrity of stored
waste and possibly result in the immediate release of radioactivity to the -surface, as well as
long-term releases to geologic formations surrounding the breached waste.
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Probability of Occurrence The inadvertent intrusion scenarios involving
drilling for water and hydrocarbons, state that drilling activities are likely to
require a period of up to two weeks before the well depth reaches that of the
buried TRU waste.
An act of sabotage or terrorism is technically feasible, but may be logisticaUy
impossible.
Since historical acts of sabotage and terrorism have been few and sporadic and
may be motivated by political, social, and other factors, a quantitative estimate
of probability is inappropriate.
Potential Consequence - The radiological consequences of an act of sabotage
or terrorism is dependent on the methods employed for accessing the
repository and damaging contained waste. However, any successful act may
pose a potentially severe immediate and long-term radiological threat.
10.5 REFERENCES
DOE94
DOE93a
DOE93
EPA88
KAR95
NRC94
"Long Term Monitoring Design Concept Description," Draft Memorandum of
Understanding Between the U.S. Department of Energy and the U.S.
Department of Interior, 1994.
U.S. Department of Energy, "Waste Isolation Pilot Plant Site Environmental
Report for CY 1992," DOE/WIPP/93-017, 1993.
U.S. Department of Energy, "Waste Isolation Pilot Plant Land Management
Plan," DOE/WIPP 93-004, 1993.
Environmental Protection Agency, "Institutional Controls at Superfund Sites:
A Study of Implementation and Enforcement Issues," S. Nicholas EPA/600/9-
89/022, September 1988.
Karl, A., F. Bordell, and J. Shaffner, "Biological Assessment for the Proposed
Low-Level Radioactive Waste Disposal Facility Ward Valley, California "
April 1995
U.S. Nuclear Regulatory Commission, Humboldt Bay Power Plant, Unit 3
SAFSTOR Decommissioning Plan, Revision 1, Docket No. 50-133, NRC
Public Document Room, Washington D.C., July 1994,
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NRC93 U.S. Nuclear Regulatory Commission, "Institutional Controls Used to Protect
Waste Disposal Sites from Inadvertent Intrusion," SECY-93-322, Washington,
D.C., November 1993.
NRC84 U.S. Nuclear Regulatory Commission, "De Minimis Waste Impacts Analysis
Methodology," O.I. Oztunali, NUREG/CR-3585, Washington D.C., February
1984.
SAN92 Sandia National Laboratories, "Preliminary Performance Assessment for the
Waste Isolation Pilot Plant, December 1992," SAND92-0700/1, Albuquerque,
NM, December 1992.
TRA91 Trauth, K.M., Hora, S.C., von Winterfeldt, D., "Expert Judgment on
Inadvertent Human Intrusion into the Waste Isolation Pilot Plant," SAND90-
3063, Sandia National Laboratories, Albuquerque, NM, December 1991.
USE95 U.S. Ecology, Inc., "Site Stabilization and Closure for Low-Level Radioactive
Waste Management Facility," NRC License No. 16-19204-01, Richland,
Washington.
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11. Monitoring
11.1 THE NEED FOR MONITORING
11.1.1 EPA Disposal Standards
Predictive models are used in the design of the WIPP. Computer simulations of the
repository environment are used to provide information to engineers and scientists during the
design of the facility. The computer simulations use analytical, empirical, and statistical
predictive models to simulate various interconnected aspects of the disposal system. These
computational simulations and predictive models reflect the state of knowledge of the
characteristics of the disposal system at the site, and include, for example, chemical,
physical, radiological and biological components.
Analytical techniques are highly useful in predictive modeling due to the conceptual nature of
their formulation. However, such models are based on the current state of knowledge of
highly complex systems and are susceptible to imperfect understanding of individual or
systemic components. A result of this imperfect state of knowledge is the introduction of
uncertainty into the model results. Thus, it is desirable to augment predictive models with
empirical data to lower inherent modeling uncertainties.
Since the predictions associated with long-term compliance have inherent uncertainty, the
final disposal standards issued by EPA in 1985 included a provision requiring monitoring of
disposal systems to assure their compliance. EPA surveyed the capabilities and expectations
of long-term monitoring approaches. As explained in the preamble to the 1985 disposal
regulations (50 FR 38081, September 19, 1985):
Evaluating this information led the Agency to several conclusions:
(1) Perhaps most importantly, the techniques used for monitoring after disposal
must not jeopardize the long-term isolation capabilities of the disposal system.
Furthermore, plans to conduct monitoring after disposal should never become
an excuse to relax the care with which systems to isolate these wastes must be
selected, designed, constructed, and operated.
(2) Monitoring for radionuclide releases to the accessible environment is
not likely to be productive. Even a poorly performing geologic
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repository is very unlikely to allow measurable releases to the
accessible environment for several hundreds of years or more,
particularly in view of the engineered controls needed to comply with
10 CFR part 60 [for facilities subject to regulation by NRC]. A
monitoring system based only on detecting radionuclide releases — a
system which would almost certainly not be detecting anything for
several times the history of the United States - is not likely to be
maintained for long enough to be of much use.
(3) Within the above constraints, however, there are likely to be
monitoring approaches which may, in a relatively short time,
significantly improve confidence that a repository is performing as
intended. Two examples are of particular interest. One involves the
concept of monitoring groundwater sources at a variety of distances for
benign tracers intentionally released to the groundwater in the
repository; this approach can evaluate the delay involved in
groundwater movement from the repository to the environment and can
serve to validate expectations of the performance expected from the
system's natural barriers. Another concept involves monitoring the
small uplift of the land surface over the repository in order to validate
predictions of the system's thermal behavior. Both of these approaches
can be carried out without enhancing pathways for the wastes to escape
from the repository.
Based on these conclusions and the public comments on this question, EPA decided to
include an assurance requirement hi 40 CFR part 191 for long-term monitoring after
disposal: "Disposal systems shall be monitored after disposal to detect substantial and
detrimental deviations from expected performance. This monitoring shall be done with
techniques that do not jeopardize the isolation of the wastes and shall be conducted until there
are no significant concerns to be addressed by further monitoring." (§191.14(b)).
To support post-closure monitoring and to provide a baseline for comparison with future
measurements, the monitoring of parameters which are important to the long-term integrity
of the disposal system must also be performed prior to closure. This type of monitoring can
provide important information about the disposal system and can contribute to a better
understanding of how the disposal system is likely to perform after closure. Furthermore,
such information can be used to verify assumptions (about the disposal system) which form
the basis for PA.
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The word "monitoring" is not defined in 40 CFR part 191. However, monitoring is
considered as a form of active institutional control. According to §191.12(f), "active
institutional control means: (4) monitoring parameters related to disposal system
performance." §191.14(a) requires active institutional controls to be "maintained for a long
a period of time as is practicable after disposal," but contributions of active institutional
controls in PA are limited to a maximum of 100 years. In the preamble to 40 CFR part 191,
EPA noted that many commenters believed "a few hundred years" as originally proposed was
too long a period for active institutional controls to be effective and consequently limited
reliance on active institutional controls to 100 years (50 FR 38080). By inference, reliance
on monitoring for more than 100 years was not contemplated in the final rule.
11.1.2 RCRA Regulations
40 CFR part 264 contains the standards for operating hazardous waste treatment, storage and
disposal facilities. Subpart F specifies the requirements for monitoring. It describes the
monitoring program which must be conducted, the hazardous constituents to be monitored
for, and the applicable groundwater protection standards. The thrust of Subpart F is to
detect any hazardous components of the waste released to groundwater.
Subpart F also contains a waiver to its requirements for monitoring. §264.90(b)(2)(vii) states
that the Administrator can waive the requirements of Subpart F if it is shown to a reasonable
degree of certainty that the facility will not allow hazardous constituents to migrate beyond
the outer containment layer prior to the end of the post-closure care period. Since this
waiver is essentially describing the PA for the site, a waiver to the monitoring described in
Subpart F could be granted. However, if ground-water monitoring wells are employed
during the disposal phase, then the wells must be operated during a thirty-year post-closure
phase, as well (EPA86).
40 CFR part 268 contains the regulations and requirements for land disposal of hazardous
wastes. §268.6 describes petitions to allow land disposal of prohibited waste. The No-
Migration Variance Petition (NMVP) requires a monitoring plan to detect migration at the
earliest practicable time when disposing of prohibited waste in ground. This plan must
include the following information:
• the media monitored,
• the type of monitoring conducted,
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• the location of the monitoring stations,
• the specific hazardous constituents to be measured,
• the implementation schedule,
• the equipment used,
• the sampling and analysis techniques, and
• the data recording and reporting procedures.
Since the monitoring requirements specified in 40 CFR part 268 are general in nature,
sufficient flexibility should be available to develop a monitoring program which complies
with this regulation and the long-term monitoring requirements of 40 CFR part 194.
11.2 PRE-CLOSURE MONITORING
In order for the WIPP to comply with the requirements of 40 CFR part 194, predictive
modeling techniques must be used. EPA has specified that analyses need to be performed to
identify the parameters subject to pre-closure monitoring (§194.42(d)). The required analysis
will define which additional parameters will be subject to pre-closure monitoring. The
information obtained from these monitoring activities will contribute to a better understanding
of how the disposal system is likely to perform after closure.
As discussed above, the objective of the monitoring assurance requirement in 40 CFR part
191 is to detect "substantial and detrimental deviations from expected performance" after
disposal. In order to have a basis for future measurements, an adequate pre-closure baseline
must be established though measurements taken prior to repository closure. This pre-closure
monitoring is not for the same purpose as much of the pre-operational monitoring undertaken
at the WIPP site by DOE and the New Mexico EEG.
11.2.1 Pre-Disposal Monitoring to Support Operations and Closure
EEG was authorized to conduct independent site monitoring under the July 1981 Agreement
for Consultation and Cooperation between the State of New Mexico and DOE ("the
Agreement"). EEG developed a preoperational environmental monitoring plan for the WIPP
site in 1984 (SPI84).
According to the Appendix A of the Supplemental Stipulated Agreement to "the Agreement"
dated December 28, 1982 (SPI84), EEG will conduct:
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• preoperational monitoring
• monitoring during the operational phase
• monitoring during and at least two years after the decommissioning and
decontamination phase
• limited post-operational phase monitoring (i.e., after termination of
decommissioning and decontamination) for not less than five years
For a number of years, DOE has also conducted a program to define the environmental
baseline for the WIPP site (DOE93, DOE94). The Operational Environmental Monitoring
Program (OEMP), which is intended to continue during operation and through
decommissioning.
A major element of the nonradiological monitoring is meteorology which involves
measurement of wind speed, wind direction, temperature, dew point, and precipitation.
Seismic activity is measured routinely at four stations around the WIPP site.
11.2.2 Pre-Closure Monitoring to Support Assurance Requirements
The EPA compliance criteria (§194.42(c)) requires preclosure monitoring of significant
parameters as identified by an analysis by DOE. The analysis will identify parameters that
affect the disposal system's ability to contain waste or the ability to verify predictions abouts
its future performance.
The following is a discussion of several parameters which could potentially be identified as
significant for monitoring. Brine-related parameters, room closure, subsidence, and
geophysical monitoring are discussed because they are examples of parameters that can be
monitored at a mined facility and may already be monitored by DOE to support operational
activities or meet other regulatory requirements. Consideration of the many parameters that
can be monitored in association with a mined facility supported the Agency's decision to
require an analysis of parameters that can be monitored to determine those with significance
to the containment of waste within the disposal system.
11.2.2.1 Brine quantity, flux composition, and spatial disposition
EPA has reviewed DOE and DOE contractor reports on in situ tests and repository
monitoring to characterize brine flow into the repository. This work has been broken down
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into three specific programs:
• Large-scale brine inflow test (Room Q)
• Brine sampling and evaluation program (BSEP)
• Small-scale brine inflow experiments
Large-Scale Brine Inflow Test
The Room Q test which operated from July 1989 until May 1994 was designed to obtain
information on brine inflow into the WIPP from an excavation which approached the scale of
a repository disposal room. Room Q was a horizontal cylindrical excavation 108 meters long
and three meters in diameter sealed with two bulkheads. The rate of brine seepage into the
closed room was measured by various collection methods. During the 1,800-day test
approximately 210 liters of brine were collected.
Brine Sampling and Evaluation Program
The BSEP program, which has been conducted since 1985, collects general repository wide
information on brine inflow. The three program elements include:
• brine flow to boreholes
• brine flow to underground openings
• laboratory measurements of brine chemistry and brine content of rock samples
A total of 119 boreholes have been monitored for various periods of time since 1984. Of
these, 51 never produced brine or stopped producing brine during the monitoring period. As
of September 1994, 105 of the holes had been dropped from the monitoring program due to
contamination from construction activities or because the holes were lost to mine related
events (HOW94). Thirteen boreholes have continuously produced brine with typical yields
being on the order of tens of milliliters per day.
Brine seeps on the walls of the excavated surfaces are also monitored. Mapping studies have
shown that clay seams rather than the halite and anhydrite are the preferential brine sources.
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Small-Scale Brine-Inflow Experiments
This experimental work, which was conducted from September 1987 through June 1993,
involved monitoring brine inflow into 17 boreholes ranging in diameter from 5 to 90
centimeters (HOW94). Some brine inflow was observed in all but two of the holes.
Room Closure
One of the defining characteristics of the WIPP repository is closure of the disposal rooms
due to creep of the salt. Room closure includes any displacement change induced by mining.
Roof sag and floor rise are examples. Displacement is the simplest and most reliably
measured geotechnical variable associated with repository excavation and operation.
Generally displacements are largest near the most recently excavated room and diminish
roughly linearly with distance from a room. A large number of displacement measurements
have been made at WIPP for the purpose of code validation and calibration. These
measurements have been made at the surface of excavated rooms and in boreholes, also
excavated in salt. Creep of salt induces a time-dependency in the displacement field about
the repository. Other rock types, for example, anhydrite, are not considered to creep
significantly. Displacement change in these rock types ordinarily occurs concurrently with
excavation. However, salt creep constrains the motion of adjacent strata that otherwise do
not creep. Displacement measurement by itself is not sufficient for model validation. Stress
measurements are needed in addition to displacement measurements. Other mechanical
measurements and observations are also possible within the scope of pre-closure monitoring
of repository rooms.
Displacement Measurements
Instrumentation for displacement measurement commonly consists of: (1) reference points
directly attached to excavation walls, roof and floor, and (2) reference points anchored in
boreholes (borehole extensometers). The latter provide for displacement measurement within
the rock mass enclosing an excavation. Extensometer lengths of 100 ft (30 m) are common.
Instrument readings may be direct and mechanical, for example, by dial guage.
Reproducibility of readings is generally a few thousandths of an inch (a few hundredths of a
millimeter). The same is true of measurements at excavation walls which are often done by
precision tape measurements between walls and between roof and floor.
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Displacement measurements are not only the most reliable measurements made in the realm
of geotechnical monitoring, they are also the most readily used in model validation and
calibration studies. Almost all computer codes for rock mechanics analysis solve for
displacements as the primary unknowns. For this reason, direct comparison between
displacement instrument readings and corresponding computer code output is easily done.
The quick (instantaneous) displacement response of a rock mass to excavation is largely
controlled by Young's modulus, while the delayed (time-dependent) component is controlled
by rock mass (salt) viscosity. After an advance of the mining face, a period of transient
creep ensues that is followed by steady state creep, provided the interval between advances is
sufficient. Plotting displacement as a function of time, as measured and as modeled, readily
demonstrates the agreement (or lack) between the two. Separation of data into instantaneous
and time-dependent components is possible. Better agreement between model and
measurement may then be obtained by changing Young's modulus and viscosity and
rerunning the model. This circular exercise may eventually calibrate a model that otherwise
simulates geology, excavation geometry, possible bed separation, and transient creep, but
does not really validate constitutive equations at the heart of a model. Displacement data
only feed one side—the strain side—of the constitutive equations. Independent measurements
that feed the stress side of the constitutive equations are needed for model validation. Of
course, only a properly calibrated and validated model has the potential for demonstrating
"predictability." Pre-closure room monitoring should therefore provide for additional
measurements, independent of displacements, that would serve model validation as well as
calibration.
Stress Measurements
Stress changes are the second most common type of geotechnical data collected in monitoring
programs. Stress-change data are independent of direct displacement measurements, although
all instruments move to some degree according to design. Pillars between rooms are usually
the regions of greatest interest for the obvious reason that they must support the overburden
and thereby insure large-scale stability of the mine. Local stability and safety relative to roof
falls, for example, is provided by local support, perhaps by roof bolts or by the intrinsic
strength of the roof strata alone. Borehole stress gages of various types are used to monitor
stress changes in pillars. An elastic response is generally assumed for stress measurement
data reduction. Extension of such procedures to the viscoelastic domain would be necessary
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for current salt creep models used for WIPP analyses. Salt creep data may already be
available from prior in situ stress measurements. When linked to borehole extensometer
measurements, stress monitoring may provide data for technically sound model validation
studies.
Other Mechanical Room Measurements
Other measurements and observations that may be relevant to purely mechanical aspects of
pre-closure repository performance include roof bolt load measurement, and borehole
televiewing for fracture development and bed separation. Off-the-shelf instrumentation is
commercially available for making these observations.
Although not strictly a "room" measurement, observations of displacement and stress change
adjacent to repository shafts are desirable, especially near aquifers above the repository
horizon. Shaft tilt determined from inclinometer measurements would be of particular
interest. Tilt refers to deviation of the shaft line from the vertical that may be induced by
repository mining and is related to the larger issue of subsidence. Inclinometer
instrumentation is relatively expensive.
Subsidence Monitoring
Subsidence can be divided into two broad categories: surface and subsurface ground motion.
Although surface subsidence is usually implied hi discussions of mine subsidence and only
the vertical component of surface displacement is considered ("settlement"), the general
concept encompasses both categories of ground motion. The rock mechanical features of
repository performance are almost synonymous with ground motion, so subsidence is of
considerable importance to rock mechanics analyses that contribute to the demonstration of
computer model predictability for performance assessment. Enlargement of a mine
monitoring plan to include not only surface displacement measurements, but also
measurements within the rock mass between repository room level and the surface, is
desirable. Surface measurements and subsurface observations would serve the twin purposes:
(1) providing physical measurements for model validation, and (2) forewarning of any
threatening departure from predicted performance. Empirical and rational subsidence
approaches, each with advantages and disadvantages are available for subsidence analysis.
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Empirical Surface Subsidence Estimation
There are several well-known empirical and semi-empirical approaches to estimating surface
subsidence (settlement). The best known technique is the procedure described in the
Subsidence Handbook developed by the National Coal Board in the United Kingdom. This
procedure is two-dimensional and for mines using the longwall method that result in almost
100 percent extraction, unlike the room and pillar WIPP plan with about 22% extraction
(SAN92). Stratigraphy and rock properties are ignored. The best known semi-empirical
approach is the influence function technique which is guided by elasticity theory and then fit
to existing surface measurements. Both techniques could be successfully adapted to WIPP
site conditions after calibration against subsidence troughs associated with underground mines
in similar geologic environments. The outcome would be a simple estimate of a WIPP
surface subsidence, an estimate easily done by hand calculation. Analyses of this type have
been conducted by IT Corporation for the WIPP site to support backfill studies (ITC94).
Perhaps the simplest surface subsidence estimation is to neglect volume change of the rock
mass influenced by mining and then impose an angle of draw which defines the extent of
subsidence at the surface. The ratio of average depth of surface subsidence to mining height
is then R/[l+2(HAV)tanS] where R is the area extraction ratio (0.20 for WIPP), H is mine
depth (about 2,000 ft or 610 m), W is mine width, and d is the angle of draw. This rough
estimate indicates surface subsidence of about 1 ft (1/3 m), a realistic estimate in view of
previous surface subsidence estimates at the WIPP site. A more formal volume conservation
approach is possible through a diffusion of voids model. The advantage of empirical
approaches is clearly the ease of calculation.
A serious defect associated with empirical methods is that no account can be made of
subsequent measurements that are not hi agreement with the empirical estimate. For
example, a qualification of plus or minus 10 percent—meaning measured subsidence is
expected to be within ±10% of the empirical estimate—may not be helpful if the deviation is
caused by sinkhole formation. In fact, this claim is made in the Subsidence Handbook of the
National Coal Board only for coal mines hi the United Kingdom where geological conditions
are similar. Coal mine subsidence hi sedimentary basins throughout the world is not the
same and claims to accuracy based on statistical summaries of subsidence surveys hi other
parts of the world are not available.
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Empirical methods tacitly imply subsidence occurs in the form of a smooth trough or
depression over the mined region. The ground surface is assumed to deform continuously,
regardless of tensile strains induced by curvature of the subsidence trough. In fact, large
vertical cracks often form at the surface and subsidence may occur with fault-like motion on
these induced cracks and on preexisting cracks as well. The formation of step-like features
over mined regions is frequently the case and loose, unconsolidated surface burden may mask
the development of cracks in shallow brittle rock strata below. Subsidence may also occur in
the form of chimneys, sink holes, or pipes that propagate by caving from mine level to
surface where a "glory hole" may develop. Subsurface caving may cease and the overlying
strata deform by flexing to form a smooth trough or possibly a stepped trough at the surface.
The presence of major joint systems and faults will also influence subsidence below and at
the ground surface. Experience of nearby mines often provides valuable guidance as to the
type of subsidence expected. With such experience, empirical methods may be useful, but
are generally at a disadvantage when predictability from first principles is desired. For
example, no empirical method is known that will indicate in advance whether trough or
chimney formation is the likely subsidence mode.
Rational Subsidence Estimation
The rational approach to subsidence estimation is through a technically sound computer
model—i.e., one based on fundamental principles. A full account can then be made of the
effects of mining geometry (past and planned), repository stratigraphy (including aquifers),
geologic structure, rock mass properties, and pre-repository stress field. In this regard, the
same computer model used for room and pillar rock mechanics analysis should also be used
for subsidence analyses, if indeed, model calibration and validation are required beyond that
afforded by a small, two-dimensional "strip" model of a generic half-room and half-pillar.
In this regard, there is always a danger of oversimplification of site details that are included
in a computer model. Unless the major features of mine geometry and geology are included,
the computer model will produce results of questionable utility.
Comprehensive computer modelling of mine subsidence is indeed a significant technical
challenge because of the large size of the model region and the amount of detail needed,
especially about rooms and pillars. Nonlinearity in the form of ductile flow of salt and
possibly caving after creep rupture complicates the constitutive description as do strata
interfaces, clay seams, and the presence of other rock types, some of which are water-
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bearing. However, some large-scale three-dimensional finite element modeling has been
done, so there is a possibility for incorporating stratigraphy into an unproved model for
subsidence analysis. The outcome would be a rational forecast of surface and subsurface
displacement, strain and stress field changes as time passes and operations continue during
the pre-closure period. Direct and timely comparisons between measured and simulated
station histories could then be made, and significant deviations could be easily identified for
further investigation. Such a forecast could also be used to plan a network of surface
subsidence stations.
Subsidence Instrumentation
Surface subsidence is often monitored by repeated level surveys over a grid of survey
stations extending beyond the limits of expected surface movement above the mined area. A
rule of thumb for spacing of surface survey points is to limit spacing to l/20th of the mining
depth. This rule suggests a maximum spacing of about 100 ft (31 m) at the WIPP site. The
rule is based on differences between stations required to resolve horizontal surface strains.
Frequency of surface surveys should be appropriate to the mining rate and the amount mined
between surveys. The minimum amount of incremental extraction underground should be
determined for this purpose. Because of salt creep, this amount will be time-dependent.
Surveys more frequent than the rate of extraction multiplied by the minimum required for
detection may be of questionable benefit hi relation to cost. In this regard, the expense of
high precision leveling between stations may not be warranted in view of computer code
scatter and variability in the amount of subsidence forecast by empirical methods. Level
surveys to millimeter accuracy would not be informative when predicted subsidence varies by
tens of centimeters.
A common practice for subsidence surveys hi the mining industry is to use aerial
photography and carry out surveys annually. During the early years of repository operation,
quarterly surveys may be desirable. Subsidence determination from aerial photographs hi the
form of present and past contour plots may be substituted for leveling.
Subsidence instrumentation could be installed hi boreholes below ground surface as well as at
the surface above the repository. The first would require reconsideration of boreholes as
threats to repository integrity. Some boreholes are planned for hydrologic monitoring. This
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suggests that boreholes of limited depth that do not penetrate the repository horizon or the
zone of influence of the repository rooms may be acceptable with proper seals.
Several types of instrumentation could be installed in boreholes for monitoring subsurface
subsidence. Two borehole techniques used successfully hi the past are cross-hole seismic
monitoring and time domain reflectometry. The principal objective of both is to detect
caving above the mining horizon. Cross-hole seismic wave velocity decreases dramatically
as the caved zone propagates upwards between instrumentation holes. An array of
geophones in boreholes and at the surface would allow for general monitoring of "rock talk,"
that is, micro-seismic events associated with rock mass motion induced by repository mining
and operation. The location, amplitude, frequency, and other characteristics of the mining-
induced seismicity could be helpful in identifying the onset of anomalous rock mass behavior.
Investigation of the phenomena during the pre-closure phase would be helpful in determining
applicability during the post-closure phase when direct access to the repository is not
possible. The second borehole method involves installation of an electrical conductor that
subsequently breaks with advance of the cave zone. Signal time to the break is measured and
the cave height determined by time domain reflectometry. Establishment of successful
borehole monitoring procedures during the pre-closure phase would be particularly
advantageous during the post-closure phase.
Geophysical Monitoring
Several geophysical phenomena can be utilized to characterize subsurface features which are
dependent on the physical composition of the subsurface materials and include, but are not
limited to: conductivity, magnetic susceptibility, dielectric permittivity, radioactivity, density,
rigidity and morphology. These physical parameters control geophysical observations which
can be exploited to monitor target objects and phenomena. Several geophysical sensors are
available to directly or indirectly measure subsurface geophysical phenomena and are
described below.
Passive geophysical techniques are used to measure fields caused by the presence of
anomalous bodies, not those caused by external sources. These observations include the
Earth's magnetic field, gravitational fields, and temperature, for example. Active
geophysical techniques are based on the measurement of the response of the media to
external sources of excitation. Examples include ground penetrating radar, induced
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conductivity, acoustic and seismic methods, and nuclear activation techniques. Active
geophysical methods are distinguished by having a source and receiver, whereas passive
methods only have receivers.
In order to evaluate the performance of the facility, a monitoring system will need to be
highly sensitive to changes in subsurface conditions. The sensitivity and subsequent
effectiveness of monitoring techniques can be significantly enhanced through the use of
differential measurement techniques. With differential methods, redundant, identical
measurements are carried out over regular time intervals. The complex features contained in
many types of geophysical data are thus reduced through numerical comparison of previous
measurements. The collection and analysis of monitoring data with identical instrumentation
and configuration will serve to enhance the observation of temporal changes in repository
conditions.
Various geophysical phenomena can be exploited in order to evaluate the state of the WIPP
repository after closure. Existing sensors are available to make highly accurate and relevant
measurements which can be used to evaluate the performance of the facility. Further, the
geophysical sensor industry is very active in developing new and more sensitive instruments
which will likely be applicable to post-closure monitoring. For the purpose of current
planning, however, only existing sensor technologies need be considered.
11.3 POST DISPOSAL MONITORING
11.3.1 Jeopardizing Waste Isolation
11.3.1.1 Introduction
Data collected post-closure could be used to compare the predicted repository conditions to
the actual repository conditions. Subsequently, deviations in conditions from those of the
predictive modes could be used to validate or dispute the validity of the predictive constructs,
and to further quantify expected future performance.
To explicitly address the potential detrimental effects of environmental monitoring, the EPA
stated in §191.14(b) that any post-closure monitoring be conducted with technologies that do
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not jeopardize the isolation of the waste. There are several ramifications of this requirement
which affect the conceptual and operational aspects of post-disposal monitoring.
11.3.1.2 Ramifications of Non-Threatening Monitoring
The explicit requirement to perform all monitoring activities without jeopardizing the
integrity of the repository has physical, conceptual, and logistical ramifications. First, this
requirement may limit or exclude the use of direct monitoring techniques within the
repository or within boreholes placed hi the strata above the repository. It is likely that this
requirement will limit most or all sensor deployments to the surface, and impose strong
reliance upon remote monitoring technologies.
Second, the spatial separation between the monitoring sensors and the repository, in excess
of 2000 feet, increases the need for highly sensitive instrumentation. High sensitivity is
required due to the decrease in amplitude of most geophysical observations with increased
sensor-to-source distance. The need for enhanced sensitivity may require monitoring plans to
incorporate advanced or innovative techniques, such as differential measurement strategies.
Third, the requirement of safe and remote monitoring at WIPP may make the evaluation of
important repository performance parameters difficult, ambiguous or impossible to perform.
In such cases, observable parameters may be identified which correlate with performance-
sensitive un-observable parameters. As such, an evaluation of critical performance
parameters should be performed which includes the ability of available monitoring
technologies to make accurate parameters estimates. In cases where significant performance
parameters cannot be determined through monitoring, correlated parameters should be
identified, if possible.
Direct measurements of important parameters within the repository after closure can be
performed only if there are no detrimental effects to the repository caused by the monitoring
activity. Presently, technologies are being developed which may allow for monitoring
directly within the excavated rooms without a physical connection between the repository and
the surface.
Post-closure monitoring must be comprised of existing and available sensing and data
collection technologies. While a new technical development may provide additional options in
the future, present requirements must be met with present technologies.
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The restrictions and goals of post-closure monitoring lead to the likely use of surface or near
surface monitoring techniques. Given the parameters and correlated parameters of interest
discussed above, the employed surface techniques will predominantly exploit geophysical
measurement and monitoring technologies as described in Section 11.3.2.
11.3.1.3 Bore Hole Techniques
The placement of sensors within bore holes located above the repository represents a
potential monitoring advantage compared to surface deployed monitoring equipment. This
potential improvement in monitoring performance comes from the geometrical aspects of
borehole techniques. First, borehole sensors, by their very nature, can be located closer to
the repository. Depending on the maximum safe depth of the borehole, significant
improvements in target proximity can be achieved. This is particularly important in
repository monitoring, where measurement of small variations in background conditions are
desired. Physical phenomena such as temperature, electromagnetic field strength,
gravitational field strength, and elastic waveform amplitudes decay rapidly with distance.
Thus, monitoring capabilities may be significantly improved through the deployment of down
hole sensors.
Second, the effectiveness of many geophysical techniques is diminished due to the
contamination of geophysical data by noise. The detection of features in geophysical data
which can lead to the estimate of targeted performance evaluation parameters is based on
both the strength of the observational feature as well as the degree of noise in the data. The
important factor in environmental monitoring is thus captured in the signal to noise ratio
(SNR). Subtle important features in monitoring data may be used for parameter estimation if
noise levels are sufficiently low to achieve a high SNR. Bore hole sensors, by their isolation
and distance from the surface, can achieve extremely low noise conditions.
Third, the geometry of geophysical monitoring arrays, through the deployment of bole hole
sensors, can have a positive impact on the post-closure monitoring program. The relatively
planar feature represented by the ground surface above the repository limits geometrical
observational considerations. Bore holes, conversely, can be optimally positioned to
maximize observation resolution. As mentioned previously, cross-hole seismic measurements
made in bore holes straddling the repository may be highly effectual in monitoring
deformational changes in repository structures.
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11.3.2 Geophysical Methods
In this section a series of possible monitoring techniques are discussed which are separated
into three categories depending on the repository attribute targeted by the monitoring activity.
These categories include; direct measurements, geoelectrical properties, and geophysical
properties. The associated technologies which can be used in these areas are listed below:
• Direct Measurements
Surface Subsidence
Direct Repository Monitoring
Groundwater Monitoring
• Geoelectrical properties
Electromagnetic
Resistivity
• Geophysical properties
Seismic
Gravity
Magnetic
11.3.2.1 Surface Subsidence.
Pre-disposal subsidence monitoring was discussed in Section 11.2.2.3. This is a simple
technology by which subsurface repository characteristics are estimated through the collection
and analysis of data describing the changes in the surface topography above the repository.
Relative motions of surface can be upward or downward, and are referred to as uplift and
subsidence, respectively. Typically, a reference point within the study area is assumed to be
constant, and all motions are calculated relative to this pseudo-stationary point. Such surveys
are relativistic, and do not attempt to describe the regional uplift or subsidence trends
overall. The technology utilizes relative vertical height measurements between benchmarks
located on the ground surface. Various conventional methods are used to make these
measurements throughout a network of stations over regular time intervals. Through the
simple processing of these data, a deformational history of the surface can be determined.
With existing technologies vertical motion of the order 0.01 inch can resolved.
Since subsidence can be caused by a variety of factors including mining, water extraction,
dissolution and hydrocarbon production, the determination of the cause of subsidence
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observations can be highly non-unique. Further, the spatial and temporal resolution of the
parameters affecting the surface observation may be lower than is required to make useful
conclusions regarding the repository performance.
While subsidence monitoring is attractive due to operational and financial considerations
(subsidence monitoring is simple and inexpensive), the measurement of the deformation
history of the ground surface over 2000 feet from the repository, may not have the required
sensitivity to assess desired performance characteristic parameters.
11.3.2.2 Direct Repository Monitoring
The most effective way to monitor post-closure repository performance is to make direct
measurement of the repository. Unfortunately, no viable technologies are presently available
to connect the sub-surface repository sensors systems to the surface without direct physical
connection. The risk of jeopardizing repository integrity by creating a borehole to the
repository or by establishing a cable connection from the surface are unacceptable. Thus,
direct repository monitoring must be accomplished without direct connection.
Currently, both industrial and governmental groups are developing technologies which may
facilitate remote communication between the ground surface and the repository. Both very
low frequency (VLF) and ultra low frequency (ULF) electromagnetic propagation methods
are being developed. If this technology matures significantly, then it may provide a means to
deliver direct repository measurements for performance evaluation.
11.3.2.3 Ground water Monitoring
Groundwater monitoring is a form of direct measurement of environmental conditions which
may be useful for the evaluation of the WIPP repository performance. Groundwater
monitoring consists of the selection of borehole locations, drilling of the boreholes,
ground water well installation and sampling.
Prior to installation of the monitoring wells, depth and diameter of the well should be
established to meet the specific monitoring needs of each location. Specification of adequate
well depth and diameter depends on the purpose of the monitoring system and the geologic
system it is monitoring. Wells of different depth and diameter can be employed in the same
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groundwater monitoring system. Varying the depth interval covered by the well screens in
several wells, or in one multi-level well, can help to determine the vertical distribution of
hydraulic heads and the levels at which contaminants are present. The quality of the
groundwater may vary with depth due to several factors, such as the density of the
contaminant-water solution, lenses or layers of varying permeability, and geologic features
that may form barriers diverting fluid flow. A fully penetrating well cannot be used to
quantify or vertically locate a contamination plume, since groundwater samples collected in
wells that are screened over the full thickness of an aquifer will be representative of average
conditions across the entire screened interval. However, fully penetrating wells can be used
to establish the existence of contamination in an aquifer.
The decision concerning the depth of placement and length of the well screen is based on:
aquifer depth, thickness, and uniformity; head distribution and estimated flow in the aquifer;
permeability of the aquifer; specific yield of existing wells; anticipated depth, thickness, and
characteristics (e.g., density relative to water) of potential contaminants; expected fluctuation
in groundwater level; the expected presence of volatile organic compounds; and the type of
borehole geophysical logs expected to be deployed.
Sampling strategy decisions, including the amount of flushing a well should receive prior to
sample collection and the selection of sampling devices, depend on the intent of the
monitoring program and the in situ hydrogeologic conditions. Programs to determine overall
quality of water resources may require long pumping periods to obtain a sample that is
representative of a large volume of that aquifer. The pumped volume may be specified prior
to sampling so that the sample can be a composite of a known volume of the aquifer.
Alternately the well can be pumped until the parameters such as temperature, electrical
conductance, and pH have stabilized. Sampling instrument selection is dependent of several
factors including cost, power requirements, sample isolation requirements, decontamination
requirements, sample volumes, etc. Bailers, suction pumps, gas-lift samplers, and
submersible pumps may be applicable.
Several factors will contribute to the potential usefulness of groundwater monitoring
techniques in providing data for the evaluation of repository performance. These include,
but are not limited to: the number, type, and depth of wells located above and around the
repository; the chemical and radiological analyses performed on extracted samples; and the
time interval between subsequent sampling events.
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One groundwater monitoring technique which has been employed at other locations involves
the use of synthetic tracers. The State of New York Department of Ecology (NYDEC) is
interested in using synthetic tracers to "spike" individual landfill cells to facilitate the
detection and identification of future leaks. To support this objective, the NYDEC sponsored
a detailed tracer study, designed to evaluate the relative characteristics of numerous tracers in
the laboratory and field. Work included laboratory tests, several natural gradient tests and
recirculation/injection tests. Numerical simulations were performed to analyze the collected
data. Similar technology has also been used to detect and identify leaks from large
petroleum tank farms.
Since the travel time for non-sorbing tracers is shorter than for radionuclides subject to
chemical retardation, such tracers could conceptually provide a leading indicator for
radionuclide release from the repository. This approach is consistent with the concepts
outlined by EPA in 1985 and discussed in Section 11.1.1 above. Depending upon the results
of ground-water flow calculations, various tracers might be placed in the repository at
closure and down stream monitoring for tracer migration in the marker beds near the
repository horizon be conducted at the site boundary. This might be particularly useful in
assessing any detrimental effects caused by waterflooding of vicinity oil fields to enhance
recovery.
11.3.2.4 Electromagnetic Conductivity
Electromagnetic (EM) waves can be used to determine electrical properties of subsurface
materials within the earth. Techniques based on propagating electromagnetic energy into the
earth and measuring responses to that input have been used extensively in geophysical
exploration and characterization applications for several years.
When an alternating current is applied to a looped transmitter antenna, it acts as an
alternating magnetic dipole source. The resulting alternating magnetic field, the primary
field, induces current flow in subsurface conductors. Good conductors, like buried metal
objects or saline ground water, produce strong induction currents that decay more slowly
than induced currents in poor conducting materials. The induced current in the buried
conductor thus produces a secondary field observed at the surface.
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Since variation in magnetic permeability in the ground is small, for a fixed frequency and
fixed transmitter-to-receiver coil distance, the fields produced by the induced electric currents
will be proportional to the conductivity of the subsurface. The most common field
measurement for these systems is the ratio of the primary to secondary magnetic field.
Different relative orientations of transmitter to receiver coils respond differently depending
on the spatial geometry of the conductive bodies. In particular, the quadrature component of
the secondary field is considered most appropriate for measuring broad anomalous conductive
layers like contaminant spills. In-phase component measurements, on the other hand, are
more sensitive to the effects of local highly conductive objects like buried metal objects.
Depth of penetration and resolution of the EM induction depends on several parameters,
including average ground conductivity, source power, source frequency, and antenna spacing.
Systems used in a depth-sounding configurations, with expanding source to receiver spacing,
can estimate the depth to anomalous conducting bodies. In a profiling mode, the
source-to-receiver spacing is fixed and the area is surveyed on grid lines. This type of
survey can produce 2-D surface anomaly maps.
EM systems can be used to make estimates of the conductivity of the subsurface. In turn,
these estimates can inferentially be related to subsurface geophysical and geochemical
properties such as porosity, permeability, and concentrations dissolved electrolytes, for
example. The EM response of the subsurface is highly sensitive to water content and thus
may be useful in the mapping of aquifers and brine bodies. Highly conductive metal objects,
such as pipes, metallic debris and structures are readily detected and mapped.
EM techniques have operational limitations inherent to active geophysical techniques,
including the accurate positioning of source and receivers. EM techniques have the further
weakness of having restricted resolution due to the averaging of the media conductivity
between source and receiver. The operational concerns at WIPP must be investigated to
determine the specific usefulness of EM techniques in assisting with repository performance
evaluations.
11.3.2.5 Resistivity
Like electromagnetic conductivity methods, resistivity techniques represent well established
methods for determining the electrical properties of the subsurface. Resistivity is the inverse
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of conductivity, and is thus affected by the same factors described for the electromagnetic
conductivity technique above.
The resistivity technique utilizes a series of electrodes on the surface. Two electrodes are
energized to establish a electric current in the earth between them and two other electrodes
are used to measure the potential developed by the first pair. By varying the number, type,
and position of the electrodes, as well as varying the input currents, subsurface resistivity
parameters can be estimated.
Resistivity methods can be used to determine the thickness of electrically distinct strata as
well as the location of aquifers or brine layers.
11.3.2.6 Seismic Techniques
Seismic techniques exploit well understood elastic wave propagation phenomena and can be
used for a variety of applications ranging from the determination of the depth, thickness and
composition of geologic structures, to the identification of subsurface fracture zones or voids.
The subsurface is mapped by modeling measured seismic observations in terms of the travel
times and shapes of the recorded waveforms. The paths, velocities and amplitudes of the
waves are controlled by the distribution of the elastic moduli of the media, including the bulk
modulus (volumetric response), shear modulus (torque response), and Young's modulus
(stretch response).
Oil and gas industry seismic techniques are generally based on the recording and
interpretation of reflected and/or the refracted waves generated by controlled sources.
Various forward and inverse waveform modeling procedures have been established to process
data for interpretation of the sub-surface geologic structure. Data are collected on geophones
located either on the surface or in boreholes in one, two, or three-dimensional
configurations. The wave propagation energy source is generally a dropped weight or
controlled explosive. A wide variety of source and receiver configurations can be utilized to
exploit various wave propagation effects.
Seismic methods based on the observation of reflected waves and refracted waves (bent due
to Snell's Law) are regularly used to map hydrocarbon reserves, to locate aquifer boundaries,
and make estimates of subsurface velocity and density parameters. These highly developed
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methods may be applicable to the evaluation of the performance of the WIPP site after
closure.
Additionally, seismic surveillance techniques may assist in determining the post-closure
repository performance. These methods are based on the continuous monitoring of ground
motions on the surface or in boreholes located above and around the repository. Best known
for detecting and locating earthquakes, seismic networks can also be used to locate and
characterize microseismic perturbations within the repository.
11.3.2.7 Gravitation
The gravitational method involves systematic measurements of small deviations in the earth's
gravitational field. Observations of the gravitational forces, described by Newton's Law, are
sensitive to mass and density variations in the lithosphere. Gravity surveys can be used to
detect and map structural changes in the subsurface, such as faulted or bending strata.
Gravimeter data is collected on a grid and interpreted to provide a broad picture of the
density distribution of the media.
At WIPP, it is unlikely that gravity variations will provide sufficient spatial or temporal
resolution of parameters which describe the repository performance. However, this
technique may be useful in constructing a comprehensive and accurate understanding of the
substructure when combined with the data from other sensors.
11.3.2.8 Magnetization
Like the gravity method described above, the magnetic method takes advantage of naturally
occurring force fields. An additional similarity is that, with the magnetic method, small
deviations hi the force field, relative to the primary field, are required. Anomalies in the
magnetic field of the earth provide an effective geophysical observation to be exploited for
the detection and characterization of underground buried ferrous objects. This phenomenon
is based on the fact that some ferromagnetic minerals (magnetite, ilmenite, pyrrhotite) have
magnetic susceptibilities more than 10,000 times that of non-ferromagnetic materials.
Materials composed of ferromagnetic minerals have elevated bulk magnetic susceptibilities
which can alter the magnetic field. Variations in the magnetic field can be measured and
modeled to determine the relative distribution of ferromagnetic materials below the surface.
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Clearly, magnetic techniques are only useful in detecting and characterizing ferromagnetic
objects, typically those made of iron or steel. At WIPP, the magnetic method, by itself, will
not provide a robust or comprehensive technique for assessment of the repository post-
closure performance. However, a magnetic field survey may provide important information
about the distribution of subsurface materials which may simplify, through modeling, the
interpretation of effects observed in other data types.
11.4 CONCLUSION
As technological advances continue, the options for monitoring will expand. However,
fundamental aspects regarding the measurement of the environment at WIPP will remain
indefinitely. These aspects are related to the basic physics of the repository and include, for
example, the location of the repository structures, the employed construction techniques and
materials, the backfill composition, and the type, location, and quantity of naturally
occurring brine. An evaluation of critical repository performance parameters suitable for
monitoring must be performed to define the monitoring requirements and this evaluation is
included in the compliance criteria of 40 CFR part 194.
Monitoring at WIPP may include measurements of the repository itself as well as
measurements of the surrounding environment. Sensors may be deployed to perform a
variety of geophysical, meteorological, and radiological measurements, and may include, for
example, seismometers, magnetometers, ground penetrating radar, pulsed induction sensors,
conductivity sensors, resistivity sensors, acoustic devices, bench mark leveling devices,
global positioning system equipment, groundwater sampling devices, radiation sensors, and
meteorological instruments.
11.5 REFERENCES
DEA94 "Conclusions After Eleven Years of Studying Brine at the Waste Isolation Pilot
Plant," Deal, D.E. and R.A. Bills in Waste Management '94. Tucson. AZ.
March 1994.
DOE93 "Waste Isolation Pilot Plant Site Environmental Report for Calendar Year
1992, U.S. Department of Energy, DOE/WIPP 93-107, 1993?
DOE94 "WIPP Site Environmental Report for Calendar Year 1993, U.S. Department
of Energy, DOE/WIPP-94-2033, September 1994.
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DOE95 "Draft Long-Term Monitoring Concept Descriptions," U.S. Department of
Energy, March 1995.
DOI91 "Ultralow Frequency Electromagnetic Fire Alarm System for Underground
Mines, Report of Investigations," RI 9377:1991, U.S. Department of the
Interior, Bureau of Mines.
EPA86 "RCRA Ground-water Monitoring Technical Enforcement Guidance
Document," OSWER-9950.1, U.S. Environment Protection Agency, 1986.
HOW94 "Salado Formation Fluid Flow and Transport Containment Group - White
Paper for Systems Prioritization and Technical Baseline" (Revision 1),
Howarth, S. et al., Sandia National Laboratories, September 14, 1994.
ITC94 "Backfill Engineering Analysis Report," IT Corporation, August 1994.
SAN92 "Preliminary Performance Assessment for the Waste Isolation Pilot Plant,
December 1992," Sandia WIPP Project, Sandia National Laboratories,
SAND92-0700, Vol. 3, December 1992.
SPI84 "Proposed Preoperational Environmental Monitoring Program for WIPP,"
Spiegler, Peter, Environmental Evaluation Group, EEG-26, November 1984.
STO85 "SANCHO - A Finite Element Computer Program for the Quasistatic, Large
Deformation, Inelastic Response of Two-Dimensional Solids," Stone, C.M.,
R.D. Kreig, and Z.E. Beisinger, Sandia National Laboratories, SAND84-
2618, April 1985.
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12. Passive Institutional Controls
12.1 INTRODUCTION
In developing the 40 CFR part 191 rule, EPA recognized that the quantification of risk over
long periods of time was subject to considerable uncertainty and consequently introduced
assurance requirements into the rule to qualitatively address this uncertainty. One of these
assurance requirements deals with passive institutional controls. Passive institutional controls
are designed to reduce the probability of inadvertent human intrusion into a repository by
conveying information about location, design, and hazards of the WIPP. Preliminary PA
runs conducted by SNL have shown that human intrusion presents the most serious problem
in demonstrating compliance with the disposal standards (SAN92). Human Intrusion can
disturb the natural and engineered barriers used in the repository to contain the waste.
In this chapter the necessity of incorporating passive institutional controls in the WIPP
compliance criteria is discussed in terms of their ability to reduce the likelihood of human
intrusion.
12.1.1 Regulatory Background
This sub-section defines passive institutional controls from a regulatory point of view. The
regulations of principal interest are 40 CFR parts 191 and 194. NRC rules involving site
markers and records are included to furnish perspective on a similar regulatory approach.
EPA Regulations
Passive institutional controls (PICs) are defined in §191.12(e) as follows (50 FR 38085):
(1) permanent markers placed at the disposal site,
(2) public records and archives,
(3) government ownership and regulations regarding land or resource use, and
(4) other methods of preserving knowledge about the location and contents of a
disposal system.
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PICs are one of the assurance requirements specified in §191.14. The assurance
requirements are designed to provide additional confidence that the containment requirements
(§191.13) are realized. Substantial uncertainty is inherent in the long-term, 10,000-year
prediction of disposal system performance necessitated by the Containment Requirements.
The Assurance Requirements balance the quantitative uncertainties involved in calculating the
magnitude of radioactive releases to the accessible environment. As noted in the preamble to
the 40 CFR part 191 Standards, "Each of the assurance requirements was chosen to reduce
the potential harm from some aspect of our uncertainty about the future" (50 FR 38072).
The specific regulatory requirement stipulating the use of passive institutional controls is
included as §191.14(c), which states:
"Disposal sites shall be designated by the most permanent markers, records,
and other passive institutional controls practicable to indicate the dangers of
the wastes and their location."
The role of passive institutional controls is to reduce the probability of inadvertent human
intrusion into the repository during the regulatory time frame. Therefore, the possibility that
institutional knowledge of the repository will be lost must be considered.
The compliance criteria related to passive institutional controls are included in §194.43 of 40
CFR part 194. This section requires the compliance certification application to include
"detailed descriptions of the measures that will be employed to preserve knowledge about the
location, design, and contents of the disposal system" and estimates of "the period of time
passive institutional controls are expected to endure and be understood."
Conceptually, some credit could be allowed hi PA for the use of PICs. In theory, providing
such credit might create an incentive for enhanced PICs. EPA proposed allowing credit for
PICs and solicited public comment on such in the 40 CFR part 194 (60 FR 5766) proposal.
In finalizing 40 CFR part 194, EPA consulted the WIPP Review Committee of the National
Advisory Council on Environmental Policy and Technology (NACEPT) on three issues,
including PICs. The Committee agreed that PICs would be likely to decrease the possibility
of inadvertent intrusion into the WIPP but expressed concern about the ability of a rigorous
method by which to determine the appropriate reduction due to PICs in the future. Some
members of the Committee stated that, if credit were to be approved, the size of the credit
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should not reflect that PICs would be effective for more than a small fraction of the 10,000-
year regulatory time frame.
NRC Regulations
Though WIPP is not subject to NRC licensing regulations, NRC regulations covering
disposal of high-level wastes in geologic repositories similarly require the use of site markers
and records (10 CFR part 60). License applications must contain a SAR which includes:
(8) A description of the controls that the applicant will apply to restrict access and to
regulate land use at the site and adjacent areas including a conceptual design of
monuments which would be used to identify the controlled area after permanent
closure (§60.21c).
When the repository is ready for permanent closure, a license amendment must be obtained
which provides:
(2) A detailed description of the measures to be employed - such as land use controls,
construction of monuments, and preservation of records - to regulate or prevent
activities that could impair the long term isolation of emplaced waste within the
geologic repository and to assure that relevant information will be preserved for
future generations. As a minimum, such measures shall include:
(1) Identification of the controlled area and geologic repository operations area
by monuments that have been designed, fabricated, and emplaced to be as
permanent as practicable; (§60.51a).
12.1.2 General Background
A 1984 study conducted by the Human Interference Task Force for the Office of Nuclear
Waste Isolation (ONW1)(HUM84) concluded that long-term communication is the primary
method for reducing the likelihood of human intrusion at nuclear waste repositories (GIL85).
Vehicles for such communication can involve both markers and records. The two are closely
intertwined. A limited record can be inscribed on a marker, a marker can designate the
location of an on-site vault containing records, or a marker can specify off-site archives
where records are located.
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The ONWI Task Force asserted that, for messages on markers to be communicated, they
must be detectable, durable, comprehensive, conveyed at several levels of sophistication and
imparted by several techniques. The logic diagram developed by the Task Force to provide a
framework for modeling future communication, taken from ONWI 84, is presented in Figure
12-1.
Assuming the markers survive and the messages inscribed on them remain legible, several
scenarios can be postulated. If society either continues to advance technically or remains
static, the message may be understood.1 However, it is likely that the inscribed message
will be understood at the site only if the message is periodically updated. Sebeok relates a
widely accepted generalization in the field of semeiotics (communication through signs);
namely, all natural language and human communication systems change over time (SEB84).
An example can be drawn from the evolution of the English language. General
comprehension of Middle English (ca. 1,100 - 1,500 A.D.) is limited and general
comprehension of Old English (ca. 400 1,100 A.D.) is virtually non-existent. It has been
estimated that only about 12% of basic English words and an even lower percentage of
complex vocabulary items will exist in 12,000 A.D. (GIV82). The Nordic Committee for
Nuclear Safety Research (NKS) provides an interesting example of this English language
change over the past 600 years. The quotation is interesting because it illustrates the changes
hi the English language and typifies the difficulty that society may have in just a few
centuries understanding English of today. The following is a quotation from "Sir Gawain
and the Green Knight" written hi about 1375 A.D. (JEN93):
The stele of a stif staff the sturne hit bi gripte
That was wounden with iron to the wandes ende,
And al bigraven with grene in gravios werkes.2
1 In the context used here, society refers to present day U.S. society.
2 The grim man gripped it by its great strong handle, which was wound with iron all the way to the end, And
graven in green with graceful designs.
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Reduce the likelihood of
human interference
OBJECTIVE
Protect future societies
against human interference
in nuclear waste repositories!
1
Reduce the consequences
of human interference
Likelihood can be reduced
by site related considera-
tions to reduce incentives
Likelihood can be reduced
by effective communication
over a long time period
Likelihood can be reduced
by effective communication
over a long time period
Natural and engineered
barriers to waste migration
presently the focus of
siting and design studies
would also mitigate
against human induced
releases
K)
Effective communication
requires message
comprehension
The communication system
must be designed to elicit
the desired response
Communication must
continue over long
time periods
(durability)
Messages must be inter-
prelabte over long time
periods
Messages must be within
the reader's level of
understanding
Messages must appear
relevant to the reader
The communication system
must warn of the conse-
quences of interference
activities
16
Long-lasting techniques
must be used
Communication system
must be easily detected
A variety of transmission
techniques must be used
Messages must be perceived
by several sensory techniques
(human/instrument)
20
Remote and local
perception must be
possible
Messages must contain
sufficient information to
produce the desired result
Redudancy in individual
techniques must be used
Redundancy in techniques
to enhance observation/
perception must be used
Figure 12-1 Human Interference Logic Flow
-------
The message may not be understood by societies of the future. Such future societies possibly
would not have the technology or the motivation to intrude into a deep geologic repository.
However, a future society might take actions which could represent indirect interference with
a geologic repository. One example would be the development of large scale irrigation or
reservoir projects which could disrupt ground water flow patterns (HUM84). Large scale
irrigation has been employed for millennia. Durant noted that extensive irrigation was used
by the ancient Sumerians beginning about 4,000 B.C. This was the cornerstone upon which
the Sumerian culture was built (DUR54, p. 124). If society regresses and then advances to a
state of civilization akin to today's society, the new society might not understand any
message which survives on the site markers. Language continuity may be lost. Future
linguists might be able to decipher the marker inscriptions, but tuning and knowledge of such
decoding at the repository site might not be contemporaneous with the intent to explore for
resources. The roles of markers and messages and the extent to which they may be expected
to persist will be discussed more fully in subsequent sections.
The following sections discuss the four categories of passive institutional controls identified
in the 40 CFR part 191 standards: (1) permanent markers, (2) public records and archives,
(3) government ownership and regulations, and (4) other methods of preserving knowledge
about the disposal system. It should be noted that records can be located either on-site or
off-site. Since on-site records are closely tied to markers, they will be discussed with
markers in Section 12.2, while off-site records will be considered in Section 12.3.
12.2 PERMANENT MARKERS
Permanent markers are the foundation of any passive institutional control strategy. This
section includes examinations of archeological monuments to provide an historical
perspective on the persistence and understood purpose of various monuments. Examples of
monuments and megaliths which have endured and whose purpose is clearly understood are
described. Ancient structures which have not persisted and/or been understood are described
here to the extent possible. Programs conducted by various government agencies related to
the marking of radioactive waste repositories are also discussed. The merits of a no-marker
strategy are briefly recounted, although such a strategy would not be acceptable from a
regulatory viewpoint. As mentioned previously, the discussion of permanent markers also
includes the messages embodied in or contained on the markers.
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This section focuses on permanent surface markers. Underground markers are considered in
Section 12.5.
12.2.1 Archeological Analogues
12.2.1.1 Introduction
EPA has long recognized that the study of ancient markers or monuments may provide
insight into the effectiveness of passive markers. At the very least, WIPP planners can
learn about materials and forms of construction which are expected to last for very long
periods of time. Beyond this, it is conceivable that the study of ancient monuments can
provide information on how best to record messages and build markers in order to convey
clearly a message to future civilization.
In 1982, DOE's Human Interference Task Force (HITF) engaged The Analytic Science
Corporation (TASC) to develop recommendations of marker design based on a study of
selected archeological sites. The resulting technical report by Maureen F. Kaplan (KAP82)
considered the pyramids of Egypt, Stonehenge in England, the Nazca Lines in Peru, Serpent
Mound in the United States, the Acropolis of Athens, Greece, and the Great Wall of China.
Kaplan classified potential messages regarding the WIPP into four levels:
• Level I: Attention getter, i.e., "something is here."
• Level II: Attention getter and warning, i.e., "something is here and it is
dangerous."
• Level III: Basic information, i.e., what, who, when, why, what actions to avoid,
and where to find information.
• Level IV: Full record of information, i.e., plans, drawings, environmental impact
statements, etc.
Kaplan pointed out that "the medium of the message may determine the level of information
the marker can convey." An earthwork, for example, can convey little beyond a Level I or
perhaps a Level II message. On the other hand, the media usually employed to convey Level
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IV messages (paper, plastic, metal, electronic media) are not nearly as likely to survive the
millennia as is an earthwork.
Kaplan emphasized the importance of identifying the audience to whom a message is
addressed and the undesirable actions to be warned against. She went on to discuss such
actions. She concluded that "the primary emphasis in the marker system design should be on
detection by sight," and noted that "the distance at which the message is detectable may be
determined, in part, by whether it is desirable to actively call attention to the site or to warn
people once they have decided to investigate the area." She discussed various possible
marker designs and message contents, stressing that because "Level III and Level IV
information may only be carried by the written word," it is important to incorporate written
text into on-site monuments and to store records elsewhere as well.
Kaplan's study served as the primary basis for the analysis of monuments that appears below.
Among her important observations were:
• Monuments that require no maintenance survive best;
• Monuments made of stone or earth survive best;
• Metals are not suitable marker materials; since they tend to be recycled;
• Markers should be shaped and sized to minimize their potential for reuse;
• The majority of ancient monuments were meant to be detected by sight, at ground
level;
• If the component parts of a marker are small, and the public is not excluded from
the vicinity of the marker, the chance that the marker's message (to say nothing
of the marker itself) will survive are relatively slim.
A logic diagram formed from these observations led to the conclusion that the WIPP marking
system should be comprised of durable, megalithic, monolithic stones with engraved
symbols. Kaplan proposed that if the markers were to be visible from the air, an earthwork
should be incorporated into the design. A basic design was proposed that consisted of outer
rings of monoliths conveying Level I and II information. The outer rings surround a tumulus
over a vault in which Level IV information would be stored. Immediately surrounding the
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vault are megaliths conveying Level III information. Kaplan concluded with a discussion of
media in which Level IV information can be encapsulated and of potential designs for the
monoliths.
Other HITF reports relevant to this study include Communication Measures to Bridge Ten
Millennia (SEB84), Communications Across 300 Generations: Deterring Human Interference
with Waste Deposit Sites (TAN84), Reducing the Likelihood of Future Human Activities That
Could Affect Geological High-Level Waste Repositories (HUM84), and Expert Judgement on
Markers to Deter Inadvertent Human Intrusion into the Waste Isolation Pilot Project
(TRA93). The last report will be addressed in the conclusion of this section.
When developing the 40 CFR part 194 proposal, EPA examined a range of ancient
monuments wider than that previously investigated by DOE. Published discussions of
twenty-five ancient monuments and classes of monuments were examined in an attempt to
answer the following seven questions:
• What message(s) were the monument's creators attempting to convey?
• Were they trying to convey this message to future civilizations or to then: own
people?
• What has been involved hi interpreting the message by modern scholars?
• How sure are we that we have the message right?
• If the message had been "Don't dig here because it is dangerous," is it likely that
we would have gotten the message before digging there?
• What physical and environmental characteristics have permitted the monument to
withstand the ravages of time and vandalism?
• What physical and environmental characteristics have permitted the monument to
convey its meaning clearly through the millennia?
In many, if not most cases, the published literature does not contain explicit answers to all of
these questions. Answers often had to be inferred from the evidence.
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In addition to previous monument studies connected with the WIPP (e.g., KAP82), results
from other studies were used. No new archeological research was conducted which would
most likely would have been redundant.
Below, four examples from the 25 archeological analogues listed below are discussed with
reference to the seven questions. Summary observations are made at the end of this sub-
section. The 25 archeological examples studied are: pyramids of Egypt, Egyptian funerary
and temple monuments, monuments of Mesopotamia, Great Wall of China, pyramids and
related monuments of Mesopotamia, Adena and Hopewell mounds, Mississippian mounds,
Wisconsin effigy mounds, John East mound, Stonehenge, West Kennet long barrow, Knowth
passage tomb, Avebury stones, Maltese temples, Easter Island statues, Nan Madol, Nazca
lines, Intaglios of the California, Arizona, and Nevada deserts, Chacoan roads, Lascaux
caves and similar paleolithic painted caves, Australian rock art, African rock art, Adamgarh
painted caves, general rock art in the Western United States, and North Fork and Jeffers
petroglyphs.
12.2.1.2 Monuments of Mesopotamia
Duration: Up to ca. 6,000 years so far.
Description: The ancient Sumerians, Babylonians, Assyrians, and other civilizations of
Mesopotamia (roughly, modern Iraq) constructed major urban centers with extensive
fortifications and religious and secular buildings. Emblematic of the Sumerians, and to some
extent, their successors was the ziggurat, virtually a constructed mountain made of brick,
topped with a religious structure. Some ziggurats rivaled the Egyptian pyramids in scale.
Most Mesopotamian buildings were made of mud brick, so their upper parts have tended to
collapse, forming mounds or "tells." Foundations and lower rooms are often preserved
within these tells, as are the remains of older buildings that were covered by later
construction. Fired clay tablets containing written material in cuneiform script are commonly
found in Mesopotamian tells, as are elaborately carved statuary and bas-relief panels
portraying rulers, wars, rituals, and aspects of daily life (WOO63; MAL65).
• What message(s) were the monument's creators attempting to convey?
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Cuneiform writing, both on tablets and on monuments, transmitted historical data, religious
observations, proclamations of law, and political propaganda. Tablets also contain more
humble writings such as inventories, financial accounts, textbooks, and student's essays.
Huge structures such as palaces and ziggurats were presumably intended to impress the
viewer and, in the case of the ziggurat, to convey a sense of religious awe.
• Were they trying to convey this message to future civilizations, or to their own
people?
Some Mesopotamian monuments were rather explicitly addressed to the future. For
example, around 500 B.C. the Persian emperor Darius had the following inscribed in
cuneiform script in three languages (Old Persian, Elamite, and Akkadian) on the Rock of
Bihistun along the caravan road between Babylon and Ecbatana:
Saith Darius the King:
Thou who shall hereafter
Behold this inscription
Or these sculptures,
Do thou not destroy them
(But) thence onward
Protect them as long
As thou shalt be in good strength.
Having at the tune some three thousand years of rising and falling civilizations to look back
upon, and having extensive contact with cultures other than his own (e.g. the Greeks),
Darius may certainly have contemplated the idea of communicating with other civilizations in
the future. However, Mesopotamian monuments appear to have been designed to
communicate information, ideas, directions, and impressions to both the people of the
communities in which they existed and to surrounding contemporary groups.
• What has been involved hi interpreting the message by modern scholars?
Archeological research, including major excavations, has been conducted extensively in
Mesopotamia for over a century. One major breakthrough occurred in the late 1830s, when
Henry C. Rawlinson was successful in translating the cuneiform script on the Rock of
Bihistun. An excavation uncovering the library of Nineveh in the 1850s produced some
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25,000 tablets that provided a rich source of messages from the ancient Assyrians.
Increasing scholarly fluency in reading cuneiform script has been the key to interpreting such
messages.
• How sure are we that we have the message right?
Scholars can be quite certain that they understand straightforward written messages correctly.
Religious parables and the propaganda of warriors and rulers, however, are less firmly
understood.
• If the message had been "Don't dig here because it is dangerous," is it likely that we
would have gotten the message before digging there?
Because of increased fluency hi reading cuneiform script, this message would certainly be
understood. However, many excavations took place in Mesopotamia before scholars became
familiar with cuneiform. Moreover, since most cuneiform records are buried, it is
impossible to read their messages without digging.
• What physical and environmental characteristics have permitted the monument to
withstand the ravages of time and vandalism?
Since the mud-brick structures of the Mesopotamian cities have collapsed, they have not
entirely withstood environmental degradation. They have also been the victims of extensive
vandalism. However, as upper walls have settled down over lower rooms and stabilized, the
lower rooms have been protected. Monuments carved on hard stone have survived well, as
have tens of thousands of fired clay tablets of cuneiform script.
• What physical and environmental characteristics have permitted the monument to
convey its meaning clearly through the millennia?
The fact that cuneiform texts were both inscribed on stone monuments and imprinted on
fired-clay tablets has been the key to understanding the messages of the ancient
Mesopotamian.
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12.2.1.3 Intaglios of the California, Arizona, and Nevada Deserts
Duration: Up to 10,000 years (DAV80).
Description: The intaglios consist of gigantic geometric, anthropomorphic, and zoomorphic
figures scratched into a "desert pavement." This hard surface is created by soil deflation in
arid areas or by aligning rocks which become cemented in as part of the pavement. While
they are most commonly found in the California desert west of the Colorado River, others
have been discovered east of the river in Arizona, in Nevada, and elsewhere.
• What message(s) were the monument's creators attempting to convey?
There is no widespread agreement about what messages the intaglios were intended to
convey. They have been interpreted as reflecting shamanistic symbols, messages about
water, astronomical observation points, maps, gaming facilities, and "random, perverse
behavior" (DAV80, HUD79, RAV85).
• Were they trying to convey this message to future civilizations, or to tfieir own
people?
There is no evidence that the builders of the intaglios were trying to communicate with future
civilizations.
• What has been involved in interpreting the message by modern scholars?
Contemporary scholars have used vertical and oblique aerial imaging, intensive surface
survey and mapping, and test excavation to interpret the intaglios.
• How sure are we that we have the message right?
There is no general agreement among scholars about the message of the intaglios.
• If the message had been "Don't dig here because it is dangerous," is it likely that we
would have gotten the message before digging there?
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Because of the hardness of the desert pavement, the intaglios are not particularly conducive
to digging. On the other hand, there is nothing in their character that discourages digging.
• What physical and environmental characteristics have permitted the monument to
withstand the ravages of time and vandalism?
The extreme aridity of the desert and the stability of the desert pavement have tended to
preserve the intaglios. Their remoteness has also been crucial in their preservation.
• What physical and environmental characteristics have permitted the monument to
convey its meaning clearly through the millennia?
The intaglios convey a message with only Level I complexity.
12.2.1.4 Australian Rock Art
Duration: ca. 25,000 years so far (possibly 35,000 years).
Description: Australian rock art includes paintings, engravings, and peckings. Subjects and
styles vary with time and location and contain both simple and complex representations.
Most are polychromatic. Many pictures overlap or are superimposed on one another. Some
of the art includes symbols that appear to convey information on direction, movement, the
act of speaking, and events in the time known today as "the dreamtime." Australian rock art
is found in rock shelters and overhangs. Aboriginal Australians continue to create rock art.
• What message(s) were the monument's creators attempting to convey?
The messages encoded in Australian rock art have been the subject of intense debate.
Chalapka asserts that the art portrays a wide variety of human experiences which reflect the
artist's physical, social, and cultural environment. For example, many motifs relate to
hunting, plant gathering, and swamp life. Dominant plant and animal figures may represent
the local subsistence base. Paintings reflect economic and socio-cultural activities, as well as
"mythic" events and spirits (CHA84, GRA93). Some intent to record contemporary events,
and possibly to influence them, is suggested by such sites as the Emu Dreaming and Pig
galleries on the Cape York Peninsula. At this site, "a half dozen white men with rifles are
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depicted. In the Pig Gallery, birds are shown standing atop the bodies of two of the men,
beaks thrust into their armpits" (GRA93).
• Were they trying to convey this message to future civilizations, or to their own
people?
The intended recipients of the messages are also the subject of considerable debate. There
is, however, nothing to suggest an intention to communicate with future civilizations.
• What has been involved in interpreting the message by modern scholars?
Since aboriginal Australian culture remains alive today, ethnography, oral history, studies of
folklore, and consultation with Aboriginal experts have been primary bases for what is
known about Australian rock art. Archeological research and comparative analysis of artistic
elements have also made important contributions.
• How sure are we that we have the message right?
The messages of Australian rock art are not thoroughly understood.
• If the message had been "Don't dig here because it is dangerous," is it likely that we
would have gotten the message before digging there?
Although much Australian rock art is very abstract, much of it is also highly
representational. Often both the external and internal characteristics of animals are
portrayed. It is conceivable that a representational pictograph warning against digging would
be understood. It is also conceivable that a pictograph representing a human's internal
structure becoming afflicted would be understood as well.
• What physical and environmental characteristics have permitted the monument to
withstand the ravages of time and vandalism?
Although the painted pictures are especially fragile, they have been protected from the
weather and preserved by their location in rock overhangs. Some paintings may have also
been maintained over the years, because they have continued to figure in the ceremonial life
of Aboriginal communities.
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Chalapka identifies five variables that affect the survival of the Australian rock art:
1. Degree of protection. The less moisture present, the greater the isolation from
humans and animals, and the greater the distance from the exterior of the shelter,
the more likely the paintings will survive.
2. Type and matrix of host rock. The harder and more stable the rock, and the
greater its ability to absorb pigment, the more likely the art will survive.
3. The properties of the pigment. The pigment's ability to be absorbed is a variable
of survival. Hematite-based pigments survive best.
4. The method of application. If the pigment is applied directly to the rock, rather
than onto a clay base, the painting will survive longer.
5. Climate at the time of execution. If the climate is dry, the pigment is more likely
to be absorbed and has a better chance of having a protective mineral coating
form over it.
• What physical and environmental characteristics have permitted the monument to
convey its meaning clearly through the millennia?
To the extent that Australian rock art has conveyed its meanings, it has done so because of
the physical survival factors listed above. Also, Aboriginal Australians are still living today
and are able to interpret the paintings.
12.2.1.5 Rock Art in the Western United States (General)
Duration: ca. 10,000 years so far.
Description: Many Native American groups of the western United States produced both
pictographs (painted rock art) and petroglyphs (pecked, ground, scratched, or incised rock
art). Such art, particularly pictographs, has been produced hi recent tunes and is probably
being produced today. However, the rock art tradition dates back at least several millennia,
and rock art sites up to 10,000 years old have been reported (HED83). Styles and motifs
vary widely from tribe to tribe, region to region, and through time, but geometric,
anthropomorphic, zoomorphic, and abstract styles are common.
• What message(s) were the monument's creators attempting to convey?
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Much of the rock art in California, particularly pictographs in the Chumash area along the
south coast, is thought to represent astronomical phenomena (e.g. HED83, HUD78). Some
sites are ethnographically associated with specific rituals, such as the initiation of girls into
womanhood (e.g. TRU88). Some petroglyphs are identified as trail markers and other
mnemonic devices, vision quest location markers, and hunting ritual depictions (e.g.
MUR87). Several scholars see representations of sexual parts and acts in rock art as part of
fertility rituals. Others identify elements of rock art as vehicles to telling and remembering
of origin stories and other traditions. Rock art may also serve as markings for social group
boundaries (HED83). Petroglyphs in the form of Hopi clan symbols hi the Southwest are
thought to have marked the route to sacred sites, and as reflecting the journeys of the Hopi
ancestors recorded hi tribal tradition (JUD50). In short, western North American rock art
probably was intended to convey a wide array of messages, most of which are not completely
understood today.
• Were they trying to convey this message to future civilizations, or to their own
people?
Rock art was probably intended to communicate messages with the artist's fellow religious
practitioners, or hunters, other members of the artist's family, clan, or tribe, and (hi the case
of boundary markers and some trail markers) with other groups. Some rock art may have
been intended for communication with the supernatural. There is no evidence to suggest an
attempt to communicate with future civilizations.
• What has been involved hi interpreting the message by modern scholars?
Interpretation of rock art in the American west has involved ethnographic consultation,
excavation of rock art sites, photography, rubbings and tracings, and a wide range of
comparative analyses.
• How sure are we that we have the message right?
There is little agreement about what messages are embedded in western American rock art.
In a few cases, the testimony of ethnographic consultants has generated a fair amount of
certainty about the meaning of particular rock art sites (e.g. TRU88).
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• If the message had been "Don't dig here because it is dangerous," is it likely that we
would have gotten the message before digging there?
In cases where rock art is representational, it is possible that such a message could be
transmitted. However, the highly abstract, presumably symbolic, forms that dominate much
western American rock art would not be likely to convey such a message readily.
• What physical and environmental characteristics have permitted the monument to
withstand the ravages of time and vandalism?
Pictographs survive best when well protected from exposure to the elements. Petroglyphs
often survive on exposed surfaces, but do erode and weather. Petroglyphs deeply pecked or
polished into very hard rock, like granite, survive best. Pecked and polished petroglyphs
survive more readily than scratched or engraved forms. Burial under sand or silt can
preserve petroglyphs in almost pristine condition (TUR94). Since human vandalism is a
major cause of petroglyph destruction, remoteness from human settlements promotes
survival.
• What physical and environmental characteristics have permitted the monument to
convey its meaning clearly through the millennia?
On the whole, the meaning of western North American rock art has not been clearly
conveyed.
12.2.1.6 Summary Observations on Archeological Analogues
The review of archeological analogues to the WIPP marker system suggests that there are
few specific analogues which address all the questions posed in this section. There is little
evidence to suggest that the monuments, structures, and other markers discussed above were
explicitly intended to convey much - if any - information into the distant future. Even
highly permanent and seemingly message-rich structures like the various pyramids and the
megalithic structures of England and Europe have contemporary functions such as housing
the dead, supporting temples, making astronomical observations, and impressing the
population with the power of the government or the awesomeness of religious rites. To the
extent messages were encoded in the structures, they seem for the most part to have been, or
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at least are today understood as, fairly simple Level lor II messages, on the order of "the
pharaoh who built this structure was very powerful."
Some of the monuments were certainly intended to mark historical events for future
reference. The stelae of the Maya and other Mesoamerican civilizations are an example, as
are the painted tombs of the Egyptian pharaohs, Australian rock art, and such Mesopotamian
monuments as the Rock of Bihistun. With the exception, perhaps, of markers like the Rock
of Bihistun, there is little reason to think that these monuments were meant to convey
information to future civilizations that their creators thought would be substantially different
from their own. There is little reason to believe that anyone who created a monument in the
past designed it with an eye to communication across thousands of years of cultural and
linguistic change.
Some monuments do communicate warnings against disturbance, but these have been notably
ineffective. Tombs explicitly marked as protected by supernatural sanctions have routinely
been looted by treasure seekers and excavated by archaeologists. In almost all cases of
archeological analogues, the value of what was contained or hidden was vastly understated,
and the threat or hazard was overstated or imaginary. Of course, no one in the past warned
against deep drilling, so the possible effectiveness of such a warning cannot be directly
assessed.
The review of archeological analogues has produced some information relevant to the WIPP,
however, which may be summarized as follows:
1. The "monuments" that have managed to convey fairly detailed information over
millennia (e.g., paleolithic rock art, Australian rock art, Egyptian tomb paintings and
carvings, Mesopotamian cuneiform inscriptions, and Mesoamerican stelae) have done
so because they have been inside something very protective, like a cave, rock shelter,
or excavated tomb, or because they have been buried.
2. In some cases of European paleolithic and Australian rock art, images have survived
for possibly up to 35,000 years. These images are often portrayed hi the very fragile
medium of paint.
3. Certain graphic images communicate clearly through the millennia. Anthropomorphic
and zoomorphic figures painted on cave walls during the paleolithic period hi Europe
and Africa and by Australian aborigines and prehistoric Native Americans are easily
recognizable as such today. Geometric figures that presumably had symbolic meaning
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when they were produced, such as intaglios, the Nazca lines, and prehistoric rock art,
convey their meaning much less reliably, if at all.
4. Detailed (ca. Level IV) information on ancient monuments has for the most part
survived to the present because it was inscribed or otherwise written on the
monuments themselves (e.g., Mayan stelae, Egyptian tombs, Mesopotamian
monuments), embodied in a literature that has survived or been recovered (Great Wall
of China, pyramids of Egypt, Mesopotamian monuments), or embodied in the oral
history and cultural practices of a population that remained resident in the vicinity
(Mesoamerican pyramids and other structures).
5. The other major source of information on ancient monuments has been archeological
research. This has almost always involved excavation. Excavations conducted in and
around such monuments have never reached depths that would compromise waste
containment if conducted at the WIPP site. Future excavations could be taken to such
depths if archaeologists believe that there is something important to be learned and do
not realize the danger of deep excavation.
6. Most ancient monuments have attracted people to dig into or around them in search of
treasure or recyclable material, or for purposes of scientific research. An exception is
the Chacoan road system, which has not enticed or suggested that it would be
profitable to dig. This is because the roads do not convey the impression that
anything would be gained by digging hi their vicinity.
12.2.1.7 The WIPP Markers Panel Report
In late 1993, DOE published the opinions of two expert panels assembled by SNL to offer
advice about the WIPP marker system (SAN93). The recommendations of the panels can be
summarized as follows:
Team A
1. Warnings should be conveyed through a gestalt sense of place, through written
languages and scientific symbols, and through the use of the human face with
expressions.
2. The WIPP and a buffer area should be surrounded by earthen berms, jagged and
threatening in shape, to create a threatening sense of place.
3. Within the "keep" created by the berms there should be multiple "message kiosks"
containing Level II messages in some seven languages - the languages of the United
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Nations plus a local indigenous Native American language. The messages should be
inscribed on a granite wall protected by a partially encircling "mother wall."
4. Also, a world map should be constructed within the "keep" showing other disposal
sites, together with the original WIPP buildings, which should be left to decay. The
map should be visible from the tops of the berms.
5. Level IV information should be contained in concrete rooms buried at the four
corners of the berm system, designed to permit access but preclude the removal of the
information. The information should be inscribed on redundant layers of stone tablets
in multiple languages, each tablet being too large to be removed intact through the
entry way, which would be blocked by a sliding stone plug.
6. Construction should employ materials that are too large, too difficult, and too
worthless to tempt recycling or relocation to museums.
7. Message design should include the use of inscribed pictographs3 of human faces
expressing shock and disgust.
TeamB
1. The marking system should employ both surface and buried markers.
2. The messages must be truthful.
3. The outer extent of the marker system should be visible from the center.
4. The area marked should be surrounded by berms, which should not include a buffer
area. The berms should be spiked with materials having anomalous properties (e.g.
magnetic, radar reflective, dielectric).
5. The warning messages should be conveyed in number of ways so that if one message
is not completely understood, the message in another form can be used to fill in the
gaps. Messages should be conveyed in multiple languages, scientific symbols, and
pictographs. They should be inscribed on stone monoliths and buried in "tune
capsules."
3 The word "pictograph" is often used to mean painted rock art, as opposed to incised, pecked, or inscribed
"petroglyphs." Following common WIPP practice, the term is used in a more general sense here, to mean pictures
on stone — in this case, inscribed.
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6. The original WIPP buildings should be left in place for future archeological study,
which will preserve knowledge of what was done there.
7. Detailed information should be stored off-site.
8. The marker system should include a map showing other disposal sites around the
world, and perhaps an international radiation warning symbol.
9. In the center of the marked area there should be a granite structure containing detailed
information about the WIPP and its dangers, in both textual and pictograph form,
inscribed on large stone slabs.
10. Testing marker designs for durability and cross-cultural understanding should be
undertaken between now and the time of implementation.
EPA reviewed the report of the Markers Panel and its recommendations as an aid in what
types and to what extent PICs should be required at the WIPP.
12.2.2 NRC Studies
No NRC work on markers has been uncovered in the development of this background
information document.
12.2.3 NASA Studies
Four NASA deep space probes which have left the solar system contained symbolic messages
for other possible civilizations hi the universe. These included the Pioneer 10 and 11 and the
Voyager 1 and 2 spacecraft. A gold-anodized plaque on Pioneer uses various symbols to
depict the position of our sun relative to various pulsars, the relative size and physiognomy
of male and female compared to the spacecraft, and the track of the spacecraft from Earth
past Jupiter (NGS75).
Each Voyager spacecraft included a copper disc (protected by an aluminum cover) with
greetings from Earth people hi 60 languages. The record also contained samples of music
from different cultures and ages ranging from the 1958 recording of "Johnny B. Goode" by
Chuck Berry to Beethoven's Fifth Symphony and various Earth sounds (NGS90, NASA 77).
The record also contained digitized pictures describing the blue planet. Instructions and
equipment were included in the spacecraft for retrieving the contained information.
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12.2.4 NEA/OECD Studies
In 1990, a Working Group on Future Human Actions at Radioactive Waste Disposal Sites
was established by the Nuclear Energy Agency (NEA) of the Organization for Economic
Cooperation and Development (OECD). This group's purpose was to review and summarize
work of OECD member countries regarding treatment of future human actions on post
closure safety assessments of geologic repositories (NEA93). In discussing site marker
systems, the Working Group noted that some markers have already survived for 5,000 years
and, consequently, the task of devising markers which will persist and be understood
"appears daunting but feasible." The Working Group described surface marker studies
conducted by DOE in the United States and Agence National pour la Gestion des Dechets
Radioatifs (ANDRA) in France. (DOE work will be documented in Section 2.5 below.)
The French approach is based on concern that on-site messages may be misunderstood. .Site
markers may arouse curiosity and thereby encourage intrusion. The elements of this
approach as presented by the Working Group are:
• Markers should not be located directly above the repository, but rather 10 to
20 km away. This distance should limit direct intrusion due to curiosity and
allow the markers to be located in the same political and geographical region as
the repository itself.
• Markers should be sufficiently large enough to be recognized by any people
living hi the vicinity of the site.
• Markers should contain redundant messages indicating the exact position of the
disposal site. This information should be in a form recoverable by a civilization
knowing the basic elements about radioactivity; otherwise, the situation would be
equivalent to that for a marker built directly above the repository. Thus, the
messages and particularly the location of the disposal site should be encoded, for
example, using the symbols and quantities used in nuclear physics.
The Working Group concluded as follows:
"In summary, the Working Group considers that marker systems can form a
useful part of a system of warnings to future generations about the location and
contents of the repository. While well-designed markers may be durable and
interpretable for long periods of time, the Working Group notes that it will be
difficult to take credit for marker longevity for periods much beyond one
thousand to several thousand years from repository closure and decommissioning."
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The NEA Working Group also met with the WIPP Markers Panel to hear presentations on
the expert judgments rendered by the Panel and to audit the probability elicitation session of
one of the Markers Panel teams (TRA93).
12.2.5 DOE Studies
DOE has done a significant amount of work on markers for geologic waste repositories.
This work falls broadly into two categories:
• Studies in the early 1980s under the aegis of the Office of Nuclear Waste
Isolation (ONWI) directed toward repositories for high-level waste (KAP82,
HUM84)
• Studies beginning in the late 1980s and continuing today on markers for the
WIPP site (HOR91)
Some of the concepts developed in the ONWI work were subsequently considered by DOE
for application at a low-level waste disposal site being studied at Hanford, Washington
(ADA86). Further information from these two DOE programs is presented in the following
sections.
12.2.5.1 ONWI Program
A Human Interference Task Force operating under the direction of ONWI conducted
extensive studies on reducing the likelihood of human-initiated processes and events affecting
geologic waste repositories. The multi-disciplinary Task Force included members with
expertise in law, sociology, political science, nonverbal communication, nuclear physics,
environmental science, archeology, climatology, linguistics (and semeiotics), behavioral
psychology, materials science, nuclear waste management and engineering (HUM84, SEB84,
TAN84).
Assuming a remote, flat, non-glacial site, the Task Force proposed that granite markers
about seven meters high and spaced at intervals no greater than 1000 meters should be used
to define the perimeter of the site. The monolithic triangular pyramids would be inscribed
with appropriate messages and warnings in several languages, supplemented with symbols
and pictograms. The site would also be centrally marked with three large triangular
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monuments defining the location of three granite storage vaults where site records would be
stored.
The Task Force suggested that these markers could be supplemented with earthworks and
anomalies capable of being detected by remote sensors. Recommended earthworks included
an arrow-shaped central plaza about 100 meters across and several meters high, on which are
located the central markers and storage vaults described above and a segmented berm located
several hundred meters from the central plaza, as shown in Figure 12-2. To enhance the
durability of these earthworks, the Task Force suggested that they be covered with 0.15 to
0.30 meters of an aggregate asphalt mixture. This proposal was based on the fact that
natural asphalts have been used since antiquity. This protective layer would also hinder
growth of vegetation on the earthworks. Because various materials would be used in
constructing the central plaza, it should create an anomaly which would remain recognizable
by several remote sensing techniques.
The Task Force reached no conclusions as to the effectiveness of markers and other measures
to reduce the likelihood of human intrusion. The Task Force noted that any such conclusions
must be based on site-specific analysis. These analyses should use probabilistic assessment
techniques to estimate the effectiveness of a highly redundant system even though some
elements may fail. The Task Force observed that the expected longevity of earthworks,
monuments and vaults was a subject requiring further investigation.
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KJ
K)
0\
Periphery Is Stabilized With
Indigenous Vegetation
Document Vaults (3)
Central Markers (3)
— Rsinforced Concrete
Base Mat
Raised Earthwork Plaza
with 15-30 cm Surface of
Aggregate Asphalt Mixture
Peripheral Earthen Berm
Figure 12-2. ONWI Conceptual Design of Site Markers for Geologic Repository
-------
12.2.5.2 WIPP Program
In a 1979 study conducted by the WIPP Technical Support Contractor (WTAC) for DOE,
crude probabilistic estimates were made of the time over which monuments and records
would persist beyond the period of effectiveness for active institutional controls (BRA79).
The probabilistic estimates were based on considerations such as the observations that records
of land ownership and transferral in the IJ.S. can be traced back to the 1700s, and grave
stones in U.S. cemeteries are still legible after about 300 years. From this type of anecdotal
evidence, it was estimated that there was a 50% probability that markers would persist for
200 years beyond active institutional controls. There was also a 50% probability that
accessible records would be available for 110 years after markers were lost. The study
assigned a 95% probability to the postulate that active institutional controls would last for
100 years. It was also assumed to be likely that economically recoverable hydrocarbon
resources would be depleted by the time that active and passive institutional controls were
lost. However, it was presumed that a 50% probability of drilling at the WIPP still existed
for a 50 year period beyond loss of institutional controls. Based on these assumptions, the
time of isolation was calculated to be at least 450 years with 50% probability. Probabilities
of various isolation periods are summarized in Table 12-1.
Table 12-1. Isolation Times Prior to Drilling in the WIPP Site Area
:: Probability8
0.50
0.90
0.95
0.98
0.99
Time of Isolation
(years) •
450
300
250
200
150
' Probability that Time of Isolation beyond active
institutional controls is at least this duration.
The credit for active and passive institutional controls estimated in the WTAC study is
relatively short compared to the 10,000 year containment period required by 40 CFR part
191. More recently, in support of its WIPP-related PA activities, DOE and SNL organized
four teams of experts (the Futures Panel) to provide judgments on the probability of future
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human-initiated processes and events. The Southwest Team of experts cited an instance in
their summary report where marking of a radioactive location near WIPP has been vitiated
(HOR91, p. D-15). Under the aegis of the Plowshare Program (see 12.2.5.3 for more
detail), an underground nuclear detonation was conducted in 1961 at the Gnome Test Site,
about six miles from the WIPP. The test was performed in a bedded salt formation at a
depth of about 1,250 feet, as compared to the WIPP repository depth of 2,150 feet. The
Gnome Site was marked with a single monument which, according to the Southwest Team,
already shows signs of weathering and has shifted from its original location.
WIPP Markers Panel.4 Using a formal process similar to that involved in setting up the
Futures Panel, SNL organized an expert judgment panel called the Markers Panel 1991. The
panel served the following two purposes:
• Qualitative to develop design guidelines for markers and messages needed to
communicate with future societies concerning the location and dangers associated
with buried TRU wastes at WIPP.
• Quantitative - to estimate the probability that markers would survive for the
required time period and convey the intended message.
The Markers Panel was divided into two multi-disciplinary teams with expertise in such areas
as anthropology, materials science, architecture, linguistics, environmental engineering,
astronomy, semeiotics, archeology, and communications. Each team developed a total
marker system design which addressed architectural design, material properties, linguistics,
message levels, message media, and other marking components (e.g., international symbols
and standards). Based on the conceptual marker systems, each team was asked to estimate
probabilistically:
• the durability of the marker system for various periods up to 10,000 years
• the ability of future societies to understand the message embodied in the marker
system
The teams were asked to assume three different levels of technology in future society; higher
than current levels, at current levels, and lower than current levels. In assessing these
1 See also Section 6.2.1.7'.
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probabilities, Team A divided the regulatory future into time periods of 200; 500; 1,000;
5,000; and 10,000 years after closure of the WIPP repository while Team B considered
periods of 500, 2,000 and 10,000 years after closure. Each member of Team A developed
individual probability estimates while Team B developed a consensus estimate. Results of
these expert elicitations, reproduced from Reference SAND 93, are included as Tables 12-2
and 12-3.
Team A's estimates that the marker system would persist for 500 years ranged from 0.85 to
0.99. The lowest probability estimates are for a society where a high level of technology is
dominant. Some Team A members felt that such a society might be able to remove the
entire WIPP repository markers. Presumably, they would understand the consequences of
such action. Team B estimated the probability of marker persistence after 500 years to be
0.9, independent of the state of technology.
While both teams estimated that there was a high level of probability that the marker system
would persist for a considerable period, it should be noted that the Panel was directed to
exclude cost as a factor in the conceptual marker system designs (TRA93, p. F-21). The
extent to which the elicited probabilities would be reduced if a cost perspective was included
is unknown, but interviews with some of the members of the Markers Panel suggest that this
should not have a significant effect. The materials proposed for the marker systems are of
intrinsically low cost. What is not clear is whether the design and construction are also of
relatively low cost. They may not be.5 One member of Team A observed that a successful
marker system at WIPP "will have to be one of the greatest public works ventures hi
history." This statement was predicated upon more elaborate earthworks than the Team A
consensus recommendation. If DOE is to take credit for a marker system capable of
persisting for several millennia, DOE must be prepared to make a concomitant obligation for
the construction of such a system. A scenario developed by one of the Futures Panel teams
posed the possibility that a future bureaucracy functioning in the year 2020 might become
embroiled in a major debate on closure costs and choose to authorize only a modest marker
system (BEN91). The scenario, as outlined from the perspective of someone recounting the
debate 25 years hence, is as follows:
5 The NKS Working Group KAN-1.3 indicated that the structures proposed by the WIPP Markers Panel might
cost tens of millions of dollars (NKS93).
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"The markers recommended by a panel of experts convened by the now
defunct Department of Energy in 1990 are widely viewed as extravagant in
view of the fact that the WIPP repository has not been used to capacity and is
such a controversial topic. It now seems unlikely that the site could ever be
forgotten, its potential hazard is thought to be less than originally foreseen,
and it seems politically dangerous to advocate large sums of money for it in
view of the pressing current social problems which followed the costly
conventional weapons buildup of the 1990s. After protracted debate lasting
several years, Congress finally appropriates money for the markers, although
design compromises must be made because it is not enough to pay for the
extensive marker system envisioned in 1990."
In reviewing such a scenario, EPA recognized a more probable future argument in which an
elaborate system of PICs had been described in a long ago compliance application because
they were, "the right thing to do." The Department of the future was tasked with justifying
the expenditures of limited funds so that they could provide the most protection for the public
and the environment. Without a documented and quantifiable benefit to justify implementing
such a system, more pressing needs with quantifiable benefits could take priority.
In eliciting expert judgment as to the probability that the message contained on/in the
markers would be correctly interpreted by future intruders, various levels of technology and
various time periods were considered in a manner similar to that used to assess the
probability of marker persistence. In addition, five conceivable motivations for intrusion
were appraised including drilling for water, mineral exploration, drilling to create injection
wells for waste disposal, archeological investigation, and other scientific investigation.
Generally, the panelists grouped the first three intrusion modes together and treated the last
two as a second group. Probabilities estimated by the two panels for intrusion driven by
mineral exploration are summarized in Table 12-4 (TRA93).
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Table 12-2. Probabilities of Marker System Persisting Team A
Expert
Ast
Brill
Goodenough
Kaplan
Newmeyer
'
Sullivan
Dominant
Technology
High
Medium
Low
High
Medium
Low
High
Medium
Low
High
Medium
Low
High
Medium
Low
High
Medium
Low
Years After Closure
200
.99
.99
.99
.99
.99
.99
.99
.99
.99
.9S-.99
.9S-.99
.95-. 99
.90
.95
.95
.90
.95
.95
500
.98
.98
.98
.98
.98
.98
.98
.98
.98
.9S-.99
.9S-.99
.9S-.99
.85
.90
.90
.85
.90
.90
1,000
.95
.95
.95
.95
.95
.95
.90
.95
.98
.90-.95
.90-.95
.90-.95
.70
.85
.85
.80
.85
.85
5,000
.75
.75
.75
.70
.70
.85
.85
.90
.95
.80
.80
.90
.65
.80
.85
.70
.80
.80
10,000
.50
.60
.60
.50
.50
.80
.70
.75
.80
.70
.70
.85
.60
.60
.65
.50
.70
.70
Source: Table 5-1 TRA93
Table 12-3. Consensus Probabilities of Marker System Persisting - Team B
Dominant
Technology
High
Medium
Low
Years After Closure
500
.90
.90
.90
2,000
.85
.80
.70
10,000
.85
.60
.40
Source: Table 5-2 TRA93
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Table 12-4. Probability of Correct Interpretation of Marker Message Intrusion by Mineral Exploration
Expert
Ast
Brill
Goodenough
Kaplan
Newmeyer
Sullivan
Team B
200 Years
Technology =
Ha
.99
.99
.99
.99
.99
.95
M
.99
.99
.99
.98
.99
.95
L
.98
.95
.99
.95
.90
.80
500 Years
.90
.90
.80
500 Years
Technology =
H
.98
.95
.95
.98
.90
.90
M
.95
.95
.95
.90
.85
.90
L
.70
.90
.70
.70
.80
.60
2,000 Years
.90
.85
.70
1,000 Years
Technology =
H
.95
.95
.90
.97
.80
.85
M
.90
.95
.90
.85
.70
.85
L
.50
.70
.50
.65
.50
.40
10,000 Years
.99
.80
.30
5,000 Years
Technology =
H
.90
.95
.65
.95
.70
.70
M
.20
.95
.60
.80
.60
.70
L
.10
.60
.15
.50
.40
.10
10,000 Years
Technology =
H
.90
.95
.50
.90
.50
.40
M
.20
.95
.40
.75
.30
.40
L
.05
.50
.02
.02
.20
.01
a The levels of technology being more advanced than today (H), similar to today's level (M), and less advanced than today (L).
Source: Table 5-4, TRA93
-------
Turning again to Team A's 500 year estimates, it can be seen that the probability of correct
interpretation of the marker message ranged from 60 to 90% for a technologically less
advanced society to 90 to 98% for a technologically more advanced society. On the same
basis, the Team B probability estimates ranged from 70% to 90%. Higher probabilities were
assigned to understanding the marker message by potential intruders seeking archeological or
other scientific knowledge.
If one averages the individual Team A estimates and then averages Team A and Team B
estimates, the probability of correct interpretation of the marker message by a future society
conducting mineral exploration with the same level of technology as today is 0.91 at 500
years after closure. Performing the same type of averaging for marker system persistence,
the probability that the system will persist at the current level of technology after 500 years
is 0.93. The probability that the markers will persist and that their message will be correctly
interpreted is therefore 0.84.
The Markers Panel recommended to DOE that additional study is warranted in three areas:
• Durability of marker materials under the WIPP site conditions including the
mechanism for attaching or inscribing messages and the interaction of wind, sand,
and water with marker materials and configurations
• Interpretation of graphic or pictorial messages that are independent of culture
• Interpretation of written messages that are independent of culture
EPA did not review the Markers Panel Team's results in order to apply numerical values to
credit for PICs or even develop a credit methodology. Instead, EPA noted the great
variability and uncertainty in the efforts of the two teams.
12.2.5.3 Project PLOWSHARE and Related Tests
The Atomic Energy Commission (AEC) established the PLOWSHARE program in June
1957, under the technical direction of the Lawrence Radiation Laboratory (LRL), now
Lawrence Livermore National Laboratory (LLNL). The program consisted of 27 nuclear
detonations conducted at the Nevada Test Site (NTS) and other sites in Colorado (2) and
New Mexico (2) from 1961 to 1973. The nuclear tests were all underground, either shaft or
cratering shots, and had yields of no more than 200 kilotons. The PLOWSHARE nuclear
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detonations were designed to explore nonmilitary applications of nuclear explosives. The
primary potential use envisioned was in large-scale engineering projects such as canal,
harbor, and dam construction, the stimulation of oil and gas wells, and mining.
The 1963 atmospheric nuclear test ban treaty caused cancellations of many of the plans, such
as those for dredging canals and excavating harbors. Other factors contributing to the failure
of PLOWSHARE to fulfill its goal were changes in national priorities, Government and
industry's disinterest in the program, public concern over the health and safety aspects of
using nuclear power for civil applications, and shortages in funding. Several other
underground nuclear tests (Vela Uniform Events and weapons tests) were also conducted
away from the NTS. A total of 11 underground nuclear tests were conducted at locations
other than the NTS since the beginning of testing through December 1973.
Comments on a few of these tests relevant to human intrusion and PICs are discussed below.
Project Gnome:
Project Gnome, a shaft detonation, was fired on December 10, 1961, at a site 40
kilometers southeast of Carlsbad, New Mexico. The site was in the Salado formation
of the Delaware Basin. This geologic formation consisted primarily of halite (rock
salt), with minor traces of anhydrite, polyhalite, silt, and claystone. The top of the
salt formation was approximately 710 feet below the site surface. The device was
buried 1,184 feet underground in bedded rock salt at the end of a 1,116-foot hooked
tunnel meant to be self-sealing. A shaft 1,216 feet in depth and ten feet in diameter
ended hi a station room connected to the tunnel. The detonation, which had a yield
of 3.1 kilotons, resulted in an underground dome-shaped chamber 60 to 80 feet high
and 160 to 170 feet hi diameter.
All Gnome site decontamination and decommissioning activities were completed and
terminated on September 23, 1979. A concrete and bronze monument was erected at
the Gnome surface ground zero location as an historical marker. The following
wording was inscribed on two bronze plates:
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Historical plate:
United States
Atomic Energy Commission
Dr. Glenn T. Seaborg, Chairman
Project Gnome
December 10, 1961
The first nuclear detonation in the Plowshare Program to develop
peaceful uses for nuclear explosives was conducted below this spot at
a depth of 1,216 feet in a stratum of rock salt. The explosive,
equivalent to 3,100 tons of TNT, was detonated at the end of a
horizontal passage leading from a vertical shaft located 1,116 feet
southwest of this point. Among the many objectives was the
production and recovery of useful radioactive isotopes, the study of
heat recovery, the conduct of neutron physics experiments, and the
provision of seismic source for geophysical studies.
Restrictive plate:
No excavation and/or drilling is permitted to penetrate Section 34,
Township 23 South, Range 30 East, New Mexico Principal Meridian,
at any depth between the surface and 1,500 feet. (DOE81)
If Section 34 is leased, a "special stipulation" is to be put into the lease by the U.S.
Geological Survey (USGS). This stipulation would require the drilling operator to
abide by the lease stipulation to protect the area between the surface and 1,500 feet
below the surface; no exceptions are to be allowed. The BLM is to ensure that no
drilling or excavation will occur.
Project Rulison:
Project Rulison was an experiment co-sponsored by AEC and Austral Oil Co. to
determine the technical and economic feasibility of using nuclear explosives to
stimulate the flow and recovery of natural gas from the Mesa Verde formation in the
Rulison Field, Garfield County, Colorado. The test, conducted near Rifle, CO on
September 10, 1969, consisted of a 43 kiloton nuclear explosive emplaced at an 8,426
foot depth. Production testing began in 1970 and was completed in April 1971.
Cleanup was initiated in 1972 and wells were plugged in 1976. Some surface
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contamination resulted from decontamination of drilling equipment and fallout from
gas flaring (burning). Soil was removed during the cleanup operations.
Project DRIBBLE:
Project DRIBBLE was comprised of four explosive tests, two nuclear and two gas. It
was conducted in the Tatum Salt Dome area of Lamar County, Mississippi, near the
communities of Baxterville and Purvis, under the Vela Uniform Program. The
purpose of Project DRIBBLE was to study the effects of decoupling on seismic
signals produced by explosives tests. The first test, SALMON, was a nuclear device
with a yield of about 5.3 kilotons, detonated on October 22, 1964, at a depth of 2,710
feet. This test created the cavity used for the subsequent tests, including STERLING,
a nuclear test conducted on December 3, 1966, with a yield of about 380 tons, and
the two gas explosions, DIODE TUBE, conducted on February 2, 1969, and HUMID
WATER, conducted on April 19, 1970.
The nuclear tests resulted in the release of radioactive elements into the salt rock.
Although most of these radioactive elements remain in the salt dome, some
contamination in the form of radioactive drill cuttings, drilling muds, and water was
brought to the surface during the drilling of boreholes into the shot cavity. Today,
the Tatum Dome Test Site has largely returned to its original state. Except for the
monument the U.S. Department of Energy erected over the location of the actual
subsurface detonations (called Surface Ground Zero), there is little indication of any
of the past activities at the site. The areas where soils were excavated have been
backfilled and seeded and now have a well established cover of vegetation. Wildlife
is abundant at the test site and the area is used for timber production and hunting.
Although the residual levels of radioactivity remaining at most of the test sites are not
extensive enough or high enough to pose an imminent or substantial risk to the environment
or public health, DOE, in conjunction with EPA and the involved states, has continued to
investigate and monitor the sites. These continued investigations are driven, in part, by the
environmental laws of the states and federal governments and, in part, by the concerns that
have been raised by the public with respect to the safety of the sites. For example, U.S.
Senator Lott identified four major issues related to the Tatum Dome Test Site in an October
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1989 letter to the U.S. Department of Energy. One of those issues was: "The control of
access to Surface Ground Zero"(DOE93b).
In addressing Senator Lett's issues, DOE has committed to conducting additional
investigations and studies at the Tatum Dome Test Site. As such, DOE will conduct a
Remedial Investigation at the site to determine if the use of the nonradioactive hazardous
substances at the site has resulted in contamination of the soil, ground water, or surface
water. A feasibility study will also be performed to determine what measures can be taken
to reduce risks associated with the site. One of the measures that will be evaluated is the
need for fencing of the site to maintain institutional control over access to the facility.
The routine annual site visits by the Environmental Monitoring Systems Laboratory include
groundwater, air, and biological sampling. Results are reported annually in EPA's Off-site
Environmental Monitoring Report.
As to how much radioactivity was released and how much is left, the Nevada Environmental
Restoration Project (ERP) is presently sponsoring a determination of this for all underground
tests at the NTS. This work is being performed by the LLNL and the LANL. To date,
similar work for NTS off-site tests has not been funded. For the NTS on-site tests, ERP
personnel are calculating inventories of fuel, fission products, and activation products initially
and at the end of 1992. The work and calculations for the NTS on-site PLOWSHARE are
probably complete by now, however, the results are classified, but DOE/NV expects to
declassify summaries (DOE94a).
All NTS off-site PLOWSHARE sites have been decommissioned. This has included
plugging wells into the cavities and cleanup of surface structures and waste sites. Remedial
investigations/feasibility studies are being conducted. Monuments were erected at Gnome
and the Tatum Dome sites warning against excavating at the sites. Similar warnings were
attached to the property deeds. Status of site markings at the other NTS off-site locations is
not well documented and is currently being investigated (DOE94b).
12.2.6 No Marker Strategy
§191.14 requires sites to be "designated by the most permanent markers, records, and other
passive institutional controls practicable to indicate the dangers of the wastes and their
location." In spite of this regulatory admonition to mark the WIPP site, some have argued
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that a no-markers strategy is preferable and some have argued that markers should be located
away from the site of the repository.
The presence of markers on which the meaning of the inscriptions has been lost might create
a desire by future archaeologists to understand the function of the site. This desire could
result in intrusion. The markers could constitute an, "attractive nuisance", encouraging
intrusion. Kaplan suggests that a marker with a so-called Level I message which serves
solely as an attention-getter might have this result (KAP82). She also suggests that such a
situation could be avoided if a Level II message is conveyed which is both an attention-getter
and a warning ("something is here and it is dangerous"). Kaplan points out that an
earthwork marking a site may convey only a Level I message. But, if the earthwork were in
the form of a recognizable hazardous/radioactive waste symbol, the marker could reduce the
likelihood of human intrusion rather than serve as an invitation to future intrusion.
However, Givens remarks that prohibitions, curses, and other dire warnings have been
"sorry failures" in deterring tomb robbers (GIV82).
The no-marker strategy was rejected by the WIPP Marker's Panel and the NKS Working
Group KAN-1.3, but the working group report did not elaborate on the basis for their
decision (JEN93). It should also be noted that the recent NAS report on the Yucca Mountain
high-level waste repository briefly states a conclusion about the desirability of marking a
repository site (NAT95, p. 108), namely:
... passive markers could attract the curious and actually increase the risk of
intrusion. Nonetheless, we conclude that the benefits of passive markers outweigh
their disadvantages, at least in the near term.
As discussed below, two of the four teams on the WIPP Futures Panel recommended that
consideration be given to not marking the site, or at least not employing surface markers, out
of concern that markers would draw attention to the site and possibly encourage curiosity
seekers (below-grade markers could still be detected by and warn off technically advanced
societies). However, both teams on the Markers Panel rejected that advice and concluded
that the site must be marked in some manner, including surface and sub-surf ace markers, to
reduce the risk of inadvertent intrusion.
The French national waste management authority, ANDRA, has expressed similar concerns
to those of the two Futures Panel teams. ANDRA has suggested that markers might be
misunderstood by persons in a future society that lacked a knowledge of radioactivity,
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engender curiosity in such persons, and, if the society was sufficiently advanced to have deep
drilling capabilities, lead to human intrusion into the repository. ANDRA proposes an
alternative marking strategy in which markers would be placed off-site and information about
the location of the repository encoded so that it could only be recovered by a civilization
understanding basic elements of radioactivity. This strategy will be discussed in greater detail
subsequently.
The WIPP Markers Panel did not consider the use of off-site markers nor (apparently) review
the ANDRA proposal.
12.2.6.1 Futures Panel
The 1991 Futures Panel for the WIPP site (HOR91) consisted of four separate review teams
(Boston, Southwest, Washington A and Washington B). On the issue of markers, two
teams—Boston and Southwest—recommended that consideration be given to not marking the
WIPP site. The Boston Team recommended as follows (HOR91, pp. C-68 to 69):
The marker panel should consider the possibility of not marking the site. There is at
least some reason to believe that markings of any kind will be attractive to a future
society and draw special attention to the region of WIPP. Most of the potential
intrusions we studied would, if truly inadvertent, be extremely unlucky to penetrate the
repository by chance. For example, without knowledge of the specific location of the
transuranics at WIPP, a future wild cat driller would have an extremely small chance
of hitting the wastes stored at WIPP. We ask that the marker panel at least consider
whether the small risk of a coincidental penetration is more or less favorable than
attracting attention to the site with permanent markers. (Another panel on hearing this
recommendation suggested subsurface markers - no markers on the surface - but clear
markers underground near the site.)
This recommendation was based in part on various intrusion scenarios the team postulated.
For example, the team postulated one scenario in which institutional memory of WIPP is lost
but local folklore holds that something valuable was dumped in the area years ago. "Treasure
hunters" locate the markers and interpret them as warning people to stay away from the
treasure, confirming their conviction that they choose the correct site to dig for the treasure.
The markers had an effect opposite to that intended by the WIPP designers: the
treasure hunters understood the warnings as confirming their selection of the site as
containing something valuable and they started to excavate. As they encountered
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additional warning markers on different levels, they became increasingly convinced
that they had picked the correct location, (p. C-45)
The Southwest Team Report within HOR91 made similar recommendations (p. D-44):
No-marker strategy? Consider a "no surface marker" strategy, or a "soft" marker
which erodes in a few centuries, to meet short-term marking needs. Hidden
markers could still be placed underground. This avoids attracting curiosity
seekers, yet the hidden markers below can warn off high technological societies.
The risk lies in the Seesaw Scenario, since wildcatters in a reviving era receive no
warning at all.
• To Mark or Not to Mark: The crucial decision confronting the Marker Panel is
whether to use surface markers at all. A "soft" surface marker which erodes in a
few centuries will cover the short-term possibilities, and then avoid curiosity
seekers in the far future. High technologies will still be able to sense the buried
markers.
Much of the Egyptian legacy came from King Tut's tomb, the only major
unviolated burial site. It was covered by the tailings of a later tomb. Unmarked,
it escaped the grave robbers.
But not marking the WIPP imposes ignorance on our descendants, who may wish
to avoid the site but could no longer locate it well. Also, low-tech wildcatters in
re-emergent technological societies would have no warning.
12.2.6.2 Markers Panel
The 1993 "Markers Panel" report for the WIPP site (TRA93) was based on two expert panel
reports: Team A (AST92) and Team B (BAK92).
Section 1.3 of AST92 addressed the question - "Should the site be marked?" - and
concluded:
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We... feel that it is essential that the WIPP site be marked in some manner, and cannot
agree with the conclusions of two of the Futures panel teams and other authors
[ROC77]6 which suggested not marking it. We take it as uncontroversial that all
people have an inherent right to understand as far as possible the forces that might
profoundly affect their well-being...
Similarly, the preface to BAK92 states that the panel reached a unanimous opinion that:
The site should be marked, on the assumption that leaving it unmarked would pose
greater risks to the future. Current mining activities in the area, alone, would make
the choice of not marking extremely risky for present-day (i.e., living) humans, and
cumulatively more dangerous for those living between now and 12,000 A.D. At
present the WIPP is in an area of active oil production, gas production and potash
mining...Surface and buried markers should be used in tandem to enhance message
redundancy...Only the land directly above the waste panels themselves — about a l/2
square-kilometer area — should be marked. (A) this would put the marker system on a
cognitive scale better geared to human perception than one spread thinly over 16
square miles. (B) Additionally, it would reduce confusion that could arise from boring
beneath a marker system beyond the panels and uncovering nothing unusual....
AST92 stated as one of its criteria for the marking system that:
The site must be marked. Aside from a legal requirement, the site will be indelibly
imprinted by the human activity associated with waste disposal. We must complete
the process by explaining what has been done and why. (p. F-ll)
6 ROC77 states: "Intelligent life is notoriously incautious in indulging its curiosity. Construction of a large
concrete mausoleum, for example, would almost guarantee that concerted efforts would be made to breach it by
intelligent, but uninformed, life. On social grounds, such a method is held to be quite reversible. Additional
irreversibility cannot be provided by warning messages, symbols, or labels. We cannot assume that even a society
that has the technology to undo rather irreversible storage will know enough about radioactivity to proceed
cautiously, or that it will be able to decipher a message it cannot read... Indeed, the presence of such an
indecipherable message would only arouse additional interest. 'Interesting' geological formations such as salt domes
are equally likely to draw attention. The society that drills into them may know nothing of radiological hazards,
but still be sufficiently advanced technologically and scientifically to be curious about the formation itself and its
possible contents... A condition for site location that aids irreversibility is that it be as uninteresting as possible,
and so draw no attention for other reasons." (p. 27)
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In other words, the site's presence will be detectable whether or not it is marked. The report
states that:
so much buried metal and radioactive material will leave a 'signature' that scientists of
the future will have no difficulty in detecting. What we need to do, of course, is to
'complete' the marking by letting them know why it is there. Also, it is projected that
after settling of the excavated and filled salt deposits, ground levels will be depressed
by at least a half foot. Even today's geologists and archaeologists can detect such a
depression; those of the future will presumably be able to do so even more readily, (p.
F-24)
12.2.6.3 ANDRA Off-Site Marker Concept
Discussions in an OECD Nuclear Energy Agency working group on human intrusion inspired
work in France on a different design for marker systems (NEA95). The design was motivated
by concern that on-site messages may not be correctly understood, and that the markers may
arouse curiosity and actually increase the risk of human intrusion.
A paper by the French national waste management authority, ANDRA, describes a concept for
off-site markers based on the reasoning outlined below (RAI93). (The paper notes that a
report published in 1987 by a special working group in France recommended a study of the
potential benefits as well as disadvantages of marking nuclear waste repositories (GOG87).
The ANDRA paper notes that some of the participants in this working group defended the
concept of marking the surface of a repository site, but others remained unconvinced, which
was the basis for the study.)
In RAI93, ANDRA considered three options:
1. Markers built directly above the disposal site
2. Leaving the site unmarked and allowing its existence to be forgotten
3. Markers built off-site
The ANDRA researchers explained the justification for off-site markers as follows:
...if messages are misunderstood or not understood at all, the marker may stir up
curiosity as a probe of an ancient civilization or valuables. It could therefore increase
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the probability of human intrusion in the sense that it is only useful to prevent
inadvertent intrusion and not intrusion due to curiosity. Other possibilities have thus
to be thought out.
We propose a different marking system that presents the advantages of a marker but
do (sic) not increase the risk of intrusion: the marker is not located directly above the
disposal but several kilometers away. It is monumental enough to resist to erosion and
to be known by any people living around, but it is far enough from the site to prevent
any intrusion due to misunderstanding of the messages and therefore to curiosity.
We think that 10-20 kilometers away from the site would be a good distance since the
marker and the site have to be in the same geographical and political region, (p. 217)
The ANDRA authors propose to describe both the contents and the location of the repository
in messages written on the marker. However, they would encode the message about the
location so that the information would only be recoverable by a civilization knowing basic
elements about radioactivity (such as by encoding with symbols and values commonly used in
nuclear physics). In this manner, a civilization unable to understand the message regarding
the dangers of the repository would not be lured to dig there.
The core of ANDRA's analysis is an evaluation of the probability of human intrusion
resulting from inadvertence or misunderstanding of markers, as a function of technological
level, for each of the three marker concepts. The analysis is summarized as follows:
• At a low technological level, there is a low potential for human intrusion
regardless of the marker concept because humans would not have the capability to
drill deep into the repository.
• At an intermediate technological level, humans would have the capability to drill
deeply but would not have an understanding of radioactivity.7 Because of the
capability to drill, the risk of inadvertent human intrusion is higher than at the low
technological level even if the site is not marked. If the site is marked, there is an
additional risk that the message will be misunderstood and will actually encourage
7 The authors suggest that this technological level corresponds to the period from the Roman era, when 400
m-deep mines were dug, to the end of the 19th century, when radioactivity was discovered.
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human intrusion. (This analysis assumes that the net effect of on-site markers for
this technological level is negative: the probability that the markers will be
misunderstood and will encourage intrusion is much greater than the probability
that the markers will be understood and will prevent inadvertent intrusion.)
• At a high technological level (equal to or greater than our own), the encoded
messages would be understood. Either an on-site or off-site marker will prevent
inadvertent intrusion. The risk that an on-site marker will be misunderstood and
encourage intrusion is also eliminated, making the effectiveness of on-site and off-
site markers equal. Both marker concepts are preferable to no marker because
they prevent inadvertent intrusion.
In effect, ANDRA separates the analysis of the potential for human intrusion into different
cases depending on technological level, and finds very different results for intermediate and
high technological levels. The ANDRA researchers agree that markers prevent inadvertent
intrusion if civilization is sufficiently advanced to understand them, but feel that such
prevention can be equally achieved through off-site markers that would not invite a less
advanced civilization to explore the site out of curiosity.
EPA did not reconsider the "no markers" option because markers are an established and
codified requirement. The same arguments that were used to support the "no markers"
argument could be applied to the "no credit" argument. However, the Agency found that the
primary responsibility was to require that the site be marked with markers that had the highest
probability of surviving, being understood, and therefore protecting the public and the
environment.
12.3 PUBLIC RECORDS AND ARCHIVES
This section discusses the use of public records and archives as passive institutional controls.
Such controls are designed to increase the probability that institutional knowledge of the
WIPP repository will not be lost by future societies. This section deals specifically with
records and archives located away from the repository site. With regard to the types of
records to be considered for archival purposes, the NKS Working Group has suggested the
following (JEN93):
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• Geographical location of the repository
• Chemical and physical properties of the waste
• Design of the repository including physical shape and barriers
• Background information and data used in the final safety (risk) assessment
• Various background materials including the final safety assessment, laws and
regulations, general information from and about society, and operational records of
the repository
12.3.1 Regulations
12.3.1.1 WIPP Land Withdrawal Act (LWA)
Under the terms of the WIPP LWA, 16 square miles of land around the WIPP site are
withdrawn from all forms of entry, appropriation, or disposal under the public land laws.
According to Sec. 3(c), of the LWA, Land Description, the boundaries of the land withdrawn
for the WIPP site are described on a map, issued by BLM which is entitled "WIPP
Withdrawal Site Map," dated October 9, 1990, and on file with the New Mexico State Office
of the BLM. Under the LWA, the Secretary of the Interior is required to publish a
description of the withdrawal area in the Federal Register and to file copies of the legal
description of the withdrawal area and the site map with the U.S. Congress, the Governor of
the State of New Mexico, the Secretary of Energy, and the Archivist of the United States.
On November 24, 1993, BLM published a description of the WIPP in the Federal Register as
required by the LWA (57 FR 55277). The notice is included as Appendix 12A. BLM also
submitted the required documentation to various governmental organizations on November 16,
1992. A sample transmittal letter is included as Appendix 12B. (While this information was
supplied to the Archivist of the United States and presumably has been filed, the existence
and location within the Archives have not been uncovered in spite of numerous inquiries.)
12.3.1.2 40 CFR part 194
§194.43 specifies that the compliance certification application include a detailed description of
plans for the "placement of records in the archives and land record systems of local, State,
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and Federal governments, and international archives, that would likely be consulted by
individuals in search of unexploited resources." The records must identify:
• The location of the controlled area and the disposal system
• The design of the disposal system
• The nature and hazard of the waste
• Geologic, geochemical, hydrologic, and other site data pertinent to the
containment of waste in the disposal system
• The results of tests, experiments, and other analyses pertinent to the containment
of waste in the disposal system
12.3.2 Historical Perspective on Use and Survivability of Records
Although records describing the WIPP site have been filed with various government agencies,
the key question to be asked is "Will these records persist and for how long?" An essential
element in the efficacy of records as a component in a passive institutional controls strategy is
survivability of the records. Written or pictographic records have endured for almost 6,000
years. Sumerian inscriptions have been documented on stone dating from about 3600 B.C.
and on clay tablets from about 3200 B.C. Egyptian hieroglyphics on various monuments date
from about the same time. The famous Code of Hammurabi was inscribed on diorite in about
2100 B.C. (DUR54). Written Chinese, which dates to the Shang dynasty (1766-1123 B.C.),
has remained substantially unchanged over the millennia.
Various investigators have suggested that records be located in numerous locations and
include several vehicles such as maps, land-use records, geological surveys, and archival
facilities (TAN84, GIL85, GIV82). This redundancy in record keeping will increase the
probability that records will survive at some location. Even so, Tannenbaum has observed
that storage materials may not last for the required 10,000 years; therefore, records must be
periodically reproduced and perhaps translated into contemporary language (TAN84). Such
reproduction will require the existence of a "responsible institutionalized authority" to
periodically undertake this reproduction and revision. Since the availability of long-term
record-keeping materials and the existence of a responsible authority to maintain records for
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the entire regulatory period are questionable, the future availability of repository records in
useful form for the entire regulatory period is also questionable.
Gillis mentions a downside to record-keeping redundancy (GIL85). Dispersal of the
information to ensure its survivabiliry may reduce detectability by persons at the site for
whom it is most relevant.
A second key question regarding records is whether they will be understood if they do
survive. These are the same key questions which must be answered with regard to markers.
The example of the Rosetta Stone, a three language monument used in unlocking Egyptian
hieroglyphics, is often cited as an example of how maintenance of records in several
languages will promote future understanding (KAP86). However, as Givens has pointed out,
translation may be accomplished "at the expense of years of sometimes painful
decipherments" (GIV82). He also notes that some ancient scripts remain undeciphered even
today, citing Mayan, Indus Valley, Minoan Linear A, Germanic and Turkish runes, and
certain African scripts. With regard to the use of icons to covey messages, Givens observes
that even the most simple and obvious iconic signals left by antiquity may not be completely-
understood by a future society.
Seboek also takes issue with use of the Rosetta stone as a "success story" vis-a-vis language
redundancy as an aid to communication with future societies (SEB84). He observes that,
although the stone's importance was instantly identified as a possible key to deciphering
hieroglyphics upon its discovery in 1799, its "mere existence .... did not make solution
automatic." The puzzle was not solved until 1822. Kaplan takes a more optimistic view of
the deciphering of the Rosetta stone, noting that one month after the stone was made available
to scholars, the Greek text had been translated and presented at a scientific meeting (KAP82).
This certainly argues for a multiple language approach.
The NBA Working Group cited historical examples, drawn from French experience, of lost
records (NEA93). In the civil war of 1870, a fire in the Tuileries Palace destroyed archives
relating to Paris. In World War II, many records were destroyed during bombing raids.
These examples reinforce the need for worldwide, redundant record storage. Despite past
problems, the Working Group felt that it was reasonable to assume maintenance of records for
500 years.
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The Nordic Working Group KAN-1.3 commissioned studies on German archives in the 20th
century and the Vatican archives since their inception (JEN93). Both studies provided
examples of major losses of archival information as well as successful attempts to protect and
shelter the information. In the German study, the interesting observation was made that many
losses occurred after World War II. The German people, driven by extreme poverty, found
paper in the poorly guarded archives to be useful for such basic needs such as fuel or
wrapping groceries. The NKS Working Group deduced from these studies that loss of
archival information is often engendered by forces different from or external to the institutions
which created the archive. Accordingly, they concluded that "an international and
internationally respected archive would represent a robust strategy." The Working Group
suggested that IAEA might be a candidate archive manager.
12.3.3 Survivability of Land Ownership Records
12.3.3.1 Introduction
EPA disposal regulations include a requirement that "[djisposal sites shall be designated by
the most permanent . . . passive institutional controls practicable to indicate the dangers of the
wastes and their location."8 The definition of PICs includes "public records," which
incorporate state and federal land records. The issue presented in this section is whether
existing state and federal land records will effectively delineate the WIPP so as to provide the
most "permanent . . . passive institutional controls practicable."
If the benchmark of "practicable permanence" is the time period over which cumulative
releases are to be limited, there are innumerable complexities and uncertainties. Use of
historical analogues is largely inadequate to determine whether land records of the WIPP
withdrawal would survive for even a significant portion of the 10,000 year period following
disposal. Writing is believed to have been developed in Mesopotamia as cuneiform only
5,000 to 6,000 years ago. Slightly later, hieroglyphics were developed in Egypt. Chinese
script is the oldest writing still in use. The fact is, however, that writing itself has only been
in existence for about 5,000 years. Based upon historical precedent, it would be sheer
speculation to conclude that any written record would last for 10,000 years.
40 CFR § 191.14(c).
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12.3.3.2 Historical Land Records
Land ownership records were maintained by different civilizations for a variety of reasons. In
most instances, preserving an accurate list of current land owners was not a high priority. In
many civilizations, alienation of land (i.e., disposal) was not widespread. In some societies,
(e.g., 6th century B.C. Athens) alienation of land from the owning family was largely
prohibited. The legendary Doomsday Book, compiled in 1086 A.D. at the behest of William
the Conqueror, is the greatest land record of medieval Europe. (It is currently displayed at
the Public Record Office in London.) This book was not intended to be an ongoing record of
land transfers. It was used as a means of settling feudal controversies that had arisen from the
Norman conquest so that the King could obtain needed assurances from all his feudal tenants
of substance.
From at least the 2nd century B.C., the Romans had a Cartesian system of land records that
was very precise. To the extent that the records have survived, they demonstrate a concern
with obtaining an accurate "snapshot" of land ownership from time to time, rather than a
record of land transfers. This is consistent with the purpose for which the records were
maintained, namely obtaining the correct amount of tribute from each landholder in the
provinces.
Some historical land records showing a "snapshot" of land ownership have survived, largely
by historical accident. There is no indication that the keeper of the records considered them
to be of lasting significance. Thus, there is no way of judging, for example, how durable
Roman provincial land records would have been had there been any premium placed upon
their lasting existence.
Ironically, the ancient people with the greatest sense of history, the Jews, had no enduring
need for land records because of political domination by others and a nomadic existence.
This irony was heightened two millennia later in England. In medieval England, an effective
system of recordation and registration of security interests in lands was developed by the
Exchequer of the Jews. The registration of sealed contracts before officials at the Jewish
Exchequer in certain towns in England was an effective recording system and provided
necessary notice to all concerned. It thus fulfilled the chief objectives of livery of seizing9
9 Ancient ritual involving the giving of notice of land ownership.
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and was an extremely successful incursion into the feudal system. The system was too
successful to suit those who had a considerable interest in the maintenance of the feudal
system and ended with the expulsion of the Jews from England in 1290.
12.3.3.3 State Land Records in the United States
In this country, real estate recording is usually done at the county level under state law. The
systems generally resemble the type first used in the Massachusetts Bay Colony in the 1600s.
Documents which may affect title to real estate are presented for recording. Recording gives
legal priority over possible conflicting interests. Otherwise, title may be lost to a subsequent
transferee who has recorded his deed or other document affecting title.
As a rule, recording is not a prerequisite to legal validity between the parties to a land
transaction. Deeds and other instruments may create interests in property even if they are not
recorded. Furthermore, recording a void instrument will not normally make it effective.
Unless there is a land registration system in effect, which is unusual in this country,
acceptance of an instrument for recordation is not an official determination that the instrument
is legally effective.
Recording systems tend to use either a grantor-grantee or tract index, the former being older
and more prevalent. In a grantor-grantee index, instruments are indexed alphabetically
according to the grantor's and grantee's surnames. The grantee index is used to search from
the present back to the beginning of the recording system to establish the chain of owners.
The grantor index is used to find adverse recorded conveyances made by or through each
owner during the time that the owner was the apparent or actual owner of the property in
question. This type of index is inexpensive to maintain but difficult to use.
A tract index organizes instruments according to the property they affect. Instruments
affecting each segment of land are indexed for that parcel. A tract index is easy to use, but
more expensive to maintain than a grantor-grantee index.
Recording system in this country do not necessarily show who is the actual owner of a
particular piece of property. Unrecorded interests may be valid and recorded interests may be
void. It would be unusual for these records to show a withdrawal of public land.
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A system of title registration, as opposed to recordation, is used in a few parts of the United
States and throughout the United Kingdom and Scandinavia. Perhaps the most popular
system of land registration is the so-called Torrens system, which is modeled after a title
registration method used for ships. Under this system of land registration, the government
actually determines that a valid conveyance has been made. In other words, each time a
document is registered, there is an administrative determination analogous to the judicial
determination made in a quiet title action10. Registration of a conveyancing document is a
prerequisite to its validity in most of those jurisdictions having a registration system.
Therefore, a registration system is much more likely than a recordation system to show who is
the actual owner of a particular piece of property. However, governmental action, such as a
reservation or withdrawal of public lands, would not necessarily be reflected in a registration
system.
12.3.3.4 Public Land Records
Public land records are maintained by the BLM. These records generally consist of a master
title plat, a use plat, mineral lease plats, an historical index, a serial register, and case files.
The master title plat is a copy of the official survey or a composite of several surveys. It
contains references to all patents, reservations, withdrawals, rights-of-way, and similar actions.
The references on the master title plat have generally consisted of weighted lines and
abbreviations. The use plats show what uses are being made of public lands. Exceptions are
grazing leases. Oil and gas leases appear on a separate use plat. Other mineral leases appear
on separate plats.
The historical index is a chronological narrative of reservations, withdrawals, and other
actions that have affected the use or title to public lands. Much of the same information will
appear on the master title plat and the use plat.
The serial register is an index to all filings made with respect to a particular application, such
as an offer to lease. In the serial register, an offer to lease, for example, would be the date of
10 Quiet title actions are lawsuits that are brought to settle land ownership disputes. The court decides who owns
the land and , after all appeals have been exhausted, that decision is controlling. The point being made here is that
under a system of land registration, every time a document is registered, the administrative body that registers the
document has to decide whether the document is valid and whether there has been a conveyance, i.e. who owns the
land. This is similar to what a court does in a quiet title action, and very dissimilar to what most county
"recording" offices do when they simply accept for filing, a document that is given them, and make no
determination concerning its validity or who actually owns the property.
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issuance of the lease, approval of any assignments, and any applications for extension. There
is a serial register sheet for each filing.
The case file contains the original instruments that have been filed for an oil or gas lease, the
land office copy of the lease or application, and related correspondence. Case files are listed
under the serial number for the particular offer, application, or lease.
In addition to the records listed above, there are other public land records that are available,
but these are of solely historic significance because they do not necessarily contain current
information. These are the tract books and the plat books. The plat book contains plats
arranged by township and range numbers. The tract book contains entries affecting lands by
land description.
Because the WIPP is located on public lands, the durability and reliability of the public land
record system is relevant to determine the extent to which there would be permanent, passive,
institutional controls. The United States Supreme Court attempted" to define away much of
the problem of ownership of public lands by excluding from the definition of "public lands,"
lands that were subject to the claim of a third party, whether or not that claim was valid.12
If lands to which any questions exist regarding title are excluded from the definition of
"public lands," administration of the public lands becomes easier.
12.3.3.5 Past Problems With Public Land Records
It is possible to describe particular circumstances in which the public land records system has
not proven to be a reliable indication of land use, title,13 or description. These have been
11 The Court has not been consistent in its definition of "public lands." To the extent, however, that they
attempted to define away the problem, as stated in the text, they probably succeeded in alleviating some of the
burdens associated with administering the public lands, i.e., the federal land management agency's problems, but
did little to solve the problems of the claimants who may otherwise have had a valid claim to public lands.
12 Newhall v. Sanger, 92 U.S. 761, 763 (1875); Bardon v. Northern Pacific R.R., 145 U.S. 535, 545 (1892).
13 Traditionally the public lands were equated with the public domain obtained by the United States from the
1780s until 1867. This definition has been expanded over the years and now includes interests in land and "acquired
lands." 43 U.S.C. § 1702(e). But see, Columbia Basin Land Protection Association v. Scheslinger, 643 F.2d. 585,
601 (9th Cir. 1981). Acquired lands were lands that had been in non-federal ownership and, subsequently, were
granted or sold to the United States by an individual or a state.
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recurrent problems that can be examined to see whether or not the public land records are a
passive institutional control that is permanent and practicable for purposes of describing the
WIPP withdrawal.
The previously mentioned approach taken by the Supreme Court over a century ago to
alleviate the problem of third party claims is indicative of one problem that has beset the
public land record system and the administration of the public lands for many years. In the
Treaty of Guadalupe Hidalgo in 1848, the United States obtained lands that had previously
been owned by Spain. The United States was confronted with claims of title based upon
Spanish grants. Some of these were recognized by the United States; others were not.
However, this is not a problem that should have any bearing on the public land records
accurately and enduringly depicting the WIPP withdrawal.
Another issue with the public land records system that was ultimately recognized by Congress
when it enacted the Federal Land Policy and Management Act ("FLPMA")14 in 1976, was
inaccurate surveys.15 Some of the original surveys of the public lands erroneously omitted
entire tracts of land. Omission of these lands from the surveys was due to simple error,
laziness, or, in some cases, outright fraud on the part of the surveyors. Generally, it has been
held that title to these omitted lands is in the United States and that they are subject to
administration under applicable public land laws. Although this could be a problem in
administering the WIPP withdrawal and identifying the lands withdrawn based upon existing
public land records, this is extremely unlikely.
14 43 U.S.C. §§ 1701-1782.
15 Section 211 of FLPMA authorizes the Secretary to convey omitted lands and unsurveyed islands to states or
their political subdivisions without regard to the acreage limitations contained in the Recreation and Public Purposes
Act. In some circumstances such lands could also be conveyed to an individual occupying and developing such
lands for a period of five years prior to January 1, 1975. 43 U.S.C. § 1721.
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Until 1976, when FLPMA was enacted, there were no reliable records of unpatented mining
claims. The holder of the mining claim did not have to file any notice of the location16 with
the federal government. A valid, unpatented mining claim is a property right and BLM was
faced with a serious problem in administering the public lands without any reliable indication
of where these claims were. Since enactment of FLPMA, notice of all mining claims, past
and future, must be filed with BLM. If the required notice is not filed, the claim is deemed
abandoned. Consequently, this is no longer a problem.
A situation that BLM has occasionally encountered arises from administering interests, e.g.,
oil and gas rights, in land that has not been surveyed under the System of Rectangular
Surveys. In Texas, for example, surveys are by metes and bounds and do not always close,
which may cause the description to be inadequate. This, however, is not a problem in New
Mexico and should not affect the WIPP withdrawal.
Managing 200 years of paperwork relating to the public lands17 is complicated. This is
especially true in light of the roughly 3,000 public land laws that have been enacted, repealed,
or amended at various times in our history. Furthermore, BLM public land records are
maintained at various offices in the western states and in an eastern state office. This amount
of paperwork could affect the viability of BLM public land records as suitable passive
institutional control, i.e., the most permanent passive institutional control practicable.
Fortunately, as discussed below, the BLM public land records are being converted to disks
and a software has been developed that will make such records readily accessible.
16 There is no reason to delve into the refinements of the Mining Law of 1872. At the risk of
oversimplification, a mining claim is "located" when the miner stakes out his claim and complies with the Mining
Law of 1872. No action is required by BLM or the Department of the Interior in order for this claim to be valid.
The miner is entitled a patent of the lands containing his claim for which he must make a nominal payment, but he
is not required to obtain a patent. If a miner does "proceed to patent" and the Department of the Interior finds that
there has not been a valid discovery, not only will the miner not receive the patent, but also the claim will be
invalidated.
17 One of the first problems confronted by the Continental Congress was how to dispose of the western lands
that the original states had ceded to the new federal government. The Continental Congress responded by enacting
the Land Ordinance of 1785, which established the rectangular system of survey and subsequently, in 1787, the
Northwest Ordinance, which provided for new states to be admitted into the union on an equal footing with the
existing states when certain conditions were met.
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With regard to land withdrawal, it should be noted that since 1976 the Secretary of the
Interior has had authority under FLPMA to withdraw federal lands18 from "settlement, sale,
location, or entry" under the general land laws. Prior to 1976, the Secretary had withdrawal
authority delegated to him by the President.
Although there are two cases19 that suggest otherwise, the well-understood rule is that a
withdrawal has no effect upon discretionary disposals, such as oil and gas leases, but simply
upon non-discretionary disposals, e.g., settlements, sales, locations, and entries. This is
consistent with how withdrawals were defined for the 200 years preceding FLPMA.
The rationale behind this definition is that lands do not need to be "withdrawn" from the
operation of the mineral leasing laws, for example. The Secretary of the Interior can simply
refuse to issue a lease. On the other hand, when someone can acquire rights in the land
without the Secretary or any other official doing anything, such as "locating a mining claim"
under the Mining Law of 1872, a withdrawal may be necessary to prevent the lands from
passing out of federal ownership or control. Mineral leasing would not be an immediate
concern to a knowledgeable BLM employee simply examining New Mexico land records with
a view towards a possible withdrawal of those lands from settlement, sale, location, and entry.
Furthermore, Congress itself withdrew the lands comprising the WIPP.20 There are
procedures established in FLPMA that may have uncovered any oil and gas leases and any
other "natural resource uses and values of the site and adjacent public and non-public
18 The term "federal lands" is broader than "public lands" and includes land administered by agencies other than
BLM.
19 Mountain States Legal Foundation v. Andrus, 499 F. Supp. 383 (D. Wyo. 1980). The court in Nat 'I Wildlife
Federation v. Watt, 571 F. Supp. 1145 (D.D.C. 1983) assumed that the withdrawal provisions hi FLPMA could
be used to prohibit mineral leasing. Neither opinion is particularly cogent in this regard. The terms used in
FLPMA — "settlement, sale, location, and entry" — contemplate the transfer of title, not the issuance of a mineral
lease. Udall v. Tollman, 380 U.S. 1 (1965). For a Court of Appeals decision, albeit pre-FLPMA, that recognizes
the distinction between a refusal to issue a lease and a withdrawal, see Duesing v. Udall, 350 F.2d 748 (D.C.Cir.
1965), cert, denied, 383 U.S. 912 (1966).
20 Section 2(22) of the WIPP Land Withdrawal Act defines "withdrawal" to mean an area of land, rather than
in its usual sense, which is removing that land from settlement, sale, location, and entry.
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lands"21 had those procedures been followed. This does not suggest that Congress may not,
or should not, exercise its virtually limitless power over public lands and withdraw such lands
when the need arises. However, when Congress does so, procedures that Congress itself has
established in FLPMA may be ignored, and existing uses of the lands to be withdrawn are not
fully assessed or understood.
12.3.3.6 Automation of Public Land Records
BLM is in the midst of an effort to have its public land records placed in a computerized,
"user friendly" data base. The focus of these efforts is the Automated Land and Minerals
Records System (ALMRS), which combines records of ownership, authorizations, and use
affecting the public lands. This information will come from BLM's 400,000 land and mineral
case files'. Also included will be data on legal land parcels defined in the Public Land Survey
System (PLSS). These data will form the Geographic Coordinate Data Base (GCDB), which
will incorporate the PLSS and will tie map information to points on the ground by latitude
and longitude.
GCDB will allow ALMRS data to be overlaid on discrete parcels and will provide immediate
access to data on land ownership, use, and authorizations. This data base will probably be
operational by the mid-1990s. In addition, by the year 2000, resource data will be integrated
with the ALMRS data to depict the resource values and management concerns relevant to
each parcel of public land.
12.3.4 Contemporary Examples of Lost Government Records
The following sections describe several contemporary examples where records have been
"lost" or at least were unknown or unavailable to those in need of them on a timely basis
(e.g., from archives that would likely be consulted by individuals in search of unexploited
resources).
21 43 U.S.C. § 1741(c)(2). Although the Congressional veto provision in 204(c)(l) of FLPMA, 43 U.S.C. §
1741(c)(l), is probably unconstitutional, Immigration & Naturalization Serv. v. Chada, 462 U.S. 919 (1983), there
is no reason why Congress cannot direct the Secretary to provide it with certain information concerning a proposed
withdrawal as it did in section 204(c)(2).
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12.3.4.1 Oil and Gas Leases Near WIPP
The New Mexico Environmental Evaluation Group (EEG) conducted a detailed analysis of oil
and gas leases in the vicinity of the WIPP. EEG described a situation in which the U.S.
Department of Energy failed to document the presence of a producing oil and gas well under
the southwest corner of the WIPP Land Withdrawal Area (EEG92). A brief history of this
loss in institutional knowledge, derived from the EEG report, is described below.
May 1952 - Conoco obtains an oil and gas lease NMPM Lease # NM 02953 covering
all of Section 31, T22S, R31E. (This lease lies wholly within what is now the WIPP
land withdrawal boundary described in 12.3.1.1 above.)
February 1959 - Conoco lease is divided. The southern half assigned to Richardson
and Bass as Lease NM 02953-C.
November 7, 1976 - Bass files for a permit to drill a well on Lease 02953-C.
December 10, 1976 - Energy Research and Development Administration (ERDA),
now DOE, files notice in the Federal Register of intent to withdraw 17,000 acres of
public land including Section 31 T22S, R31E for waste disposal site.
January 20, 1977 - U.S. Geological Survey (USGS) approves Bass application to drill
on Section 31.
February 9, 1977 and December 7, 1977 - ERDA files suit condemning both oil and
gas leases on Section 31 from surface to a depth of 6,000 feet.
February 12, 1979 - The court condemns leases on Section 31 to 6,000 feet and
awards damages to Conoco, Bass and others.
1980 - DOE issues the Final Environmental Impact Statement which does not show
existence of leases on Section 31.
December 11, 1981 - Bass files an application to drill hole on Section 6, T23S, R31E
deviating into Section 31.
December 16, 1981 - The USGS district office transmits the drilling request and notes
"drillsite not considered politically sensitive area."
September 13, 1982 - Well deviating under Section 31 is completed and tested.
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August 4, 1987 - DOE signs a second modification to the Consultation and
Cooperation Agreement with State of New Mexico committing to a prohibition of
"slant drilling from under the site from within the site or from outside the site," even
though DOE did not have the right to make such a commitment.
May 1990 DOE reaffirms commitment to prohibit slant drilling under the WIPP in
Final Safety Analysis Report.
October 26, 1990 DOE signs Memorandum of Understanding with Bureau of Land
Management (BLM) under which "BLM will prohibit directional drilling underneath
the WIPP site boundary, except as may be required for the development of two leases
located under Section 31; drilling may be allowed below 6,000 feet of the surface."
As matters currently stand, the rule does not preclude lease holders from drilling additional
holes under Section 31 within the WIPP site, as long as the holes penetrate the site at more
than 6,000 feet below the surface.
Some observations which may be distilled from the above chronology include the following:
• In spite of the fact that detailed records as to the existence of the oil and gas
leases within the WIPP site existed, DOE did not acknowledge such existence over
a ten-year period.
• Divided, poorly defined, and/or conflicting responsibilities among various
governmental agencies and various levels of organization within the agencies
contributed to selective instances of loss of institutional knowledge.
The existence of records, per se, does not guarantee that institutional knowledge will be
retained at the locations and be available to those who "need to know." The significant events
surrounding this oil and gas lease chronology cover a period of less than 20 years. If this
example were typical, it would be inappropriate to assume that records alone can provide
adequate information so that adverse actions do not occur over a period of many years.
The example suggests that considerable attention must be directed to the issue of how records
are to be retained. The mere existence of records may not be adequate.
It could be asserted that the example described here is a record-keeping anomaly and thus is
not a relevant criticism of systematized long-term record-keeping, the goal of which is to
ensure that basic knowledge of the WIPP repository is retained. The argument that "If a good
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record-keeping system was in place, this would not happen," has some merit. It is certainly
necessary to establish a system which defines the kinds of records that should be retained, the
retention locations, the materials used for record keeping, and their potential availability to
those who need them on a timely basis.
Similar concerns about record keeping were echoed by one of the Futures Panels assembled
by SNL. The Washington A Team enumerated the following potential issues with records
(HOR91, pp E-7 to E-10):
• Records are inadequate
• Records exist but are not accessible to intruders
• Records are accessible but not understood
• Records are accessible and understood but ignored
• Records are accessible and understood but information is lacking regarding effects
of nearby activities such as large scale mining or water withdrawal
In a total record-keeping system, steps can be taken to deal with all of these issues, except the
case where records are accessible and understood, but ignored. The burden of that type of
conscious action lies on future society, not current society. Nevertheless, the duration of the
effectiveness of the best record-keeping system is unknown.
12.3.4.2 Lost AEC Records
Another recent example of the failure of records to maintain knowledge of waste burial
operations pertains to low level nuclear waste buried on U.S. Air Force controlled land under
the authority and purview of the AEC.22 This example in no way establishes or suggests
that the sites in question pose an immediate or long term risk to human health or the
environment. Neither is there any implication of negligence on the part of individuals or the
federal government. It is merely intended to illustrate the institutional and social processes
that can contribute to the success or failure of passive controls.
22 The information presented here is taken from a document entitled Burial of Radioactive Waste in the USAF
(USA72 and revisions).
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Most of the sites in question were created in the 1950s, under the auspices of the AEC and in
accordance with accepted industry waste disposal standards. The waste materials consisted of
radioactive electron tubes, solid and liquid waste from weapons maintenance, radium oxide
paint, and medical research wastes. Some burials were made in accordance with specific
AEC (now Nuclear Regulatory Commission) licenses.
"Guidance on constructing and maintaining typical sites was given technical order procedures
which included identifying site location on appropriate maps and posting and fencing to
prevent unauthorized entry. The Air Force switched to disposal at licensed commercial sites
in the 1958-1959 time frame and the technical order requirements for burial, and site
maintenance requirements was rescinded. Unfortunately, no alternate instructions were
provided on maintaining existing sites and a gradual loss of site records ensued. In 1971, the
Air Force initiated an effort to find and consolidate existing site records and reestablish site
maintenance."
A review of the facts regarding these sites is as follows:
1. Materials were buried under authorized procedures (Air Force and AEC).
2. The materials were buried on active duty military reservations that themselves could be
considered to be under active control. However, the disposal sites were under passive
control.
3. The loss of knowledge occurred because of a lapse in institutional reporting and
maintenance procedures.
4. The lapse was not longer than 12 years (1958-1971).
5. The 12-year lapse resulted in the loss of many radioactive waste burial sites. Many
are still unaccounted for in 1994.
The following three scenarios could account for the reported losses:
1. The facilities at the time of burial did not comply with the technical directive,
therefore no location records exist.
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2. Interviews with base personnel resulted in an assertion of a burial site but there is no
location information. These sites are then reported as lost. The sites may or may not
exist.
3. The facilities did comply, but when active maintenance was lost the site fence and
placards were destroyed and the historical records, if any, were not sufficient to
establish a location.
12.3.4.3 Spring Valley Munitions Dump
The Spring Valley site in Washington DC is a highly visible example of the burial of
potentially hazardous substances (i.e., World War I era chemical munitions) and the
subsequent loss of knowledge of these activities until accidental discovery during construction
related excavation many years later (BAK94).
The site history begins in 1916-1917 when the U.S. Bureau of Mines established Camp Leach
to study chemical warfare agents. American University donated land for this effort during the
war emergency. In 1917-1918, the U.S. Bureau of Mines activity was consolidated under
military command with the establishment of the American University Experimental Station,
under the Chemical Warfare Service. By 1918, there were 1800 staffers at the station.
The burial of chemical munitions at the site was not documented by the Experimental Station.
Burial activities were considered routine and not exceptional in terms of present or future
hazards. The standing order for burial was that the material must be buried three feet under
the surface and not in contact with groundwater.
In 1921, the Secretary of American University, Albert Osbome wrote an article describing the
burial of chemical munitions at the site. He did not provide a location for the burial site.
After this article, knowledge of the site and any dumping conducted there was apparently lost.
Records did exist at American University and at the National Archives hi Suitland, Maryland
but there was no general awareness of the records.
In 1986, a backhoe operator dug up a cache of chemical munitions. He notified the
authorities and the knowledge of the site was reestablished in the subsequent investigation.
The investigation revealed the prior discovery of a "bomb" in the 1950's on the site of the
Experimental Station.
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Personal memories of the site also existed, but were not part of the institutional memory.
One citizen, Eric Olsen, was in possession of and had knowledge of photographs showing
burial activities at the Station, taken by his grandfather (BAK94).
The following key points can be derived from the Spring Valley incident:
• There was no real effort or intent of the authorities at the time of burial to retain
some sort of institutional memory of the activities.
• One case of early warning of the activities made by A. Osborne, was apparently
dismissed as a matter of little interest.
• While records of the activities existed in certain archives, there was no person or
institution who retained a knowledge of them.
• The activities were institutionally rediscovered through excavation activities.
• It is not known if other chemical munitions were discovered at the site during the
years of residential development, but not reported.
12.3.5 Format for Records
In order to ensure that messages on site markers are understood, it has been suggested that the
message be recorded in several languages. For example, the WIPP Markers Panel
recommended that the marker message be recorded in the six official languages of the United
Nations (English, Chinese, Arabic, French, Spanish, and Russian) as well as Navaho and/or
Apache (KAP86, GIV82, ADA86, TRA93). For perspective, it is helpful to recognize that
Chinese is the most widely used language spoken on Earth today (Durant). Numbers of
people speaking various languages are as follows (TRA94):
• Chinese (Mandarin) - 907 million
• English - 456 million
• Hindi - 383 million
• Spanish - 362 million
• French 123 million
While comments on the use of multiple languages have primarily involved markers, the same
logic can be applied to records. Records should be written in multiple languages as well. As
previously suggested in this chapter, consideration of recording information in the major
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languages used by religious scholars (i.e., Hebrew, Latin and Arabic) also warrants
consideration. The ONWI Human Interference Task Force envisioned that detailed
information (e.g., 500 to 1,500 pages) in English and a more condensed version (e.g., 200
pages) in multiple languages would be archived (HUM84). There may be a question of
institutional will about the extent to which detailed records will be translated into various
languages for archival purposes.
Again, drawing a parallel with marker considerations, records must be stored on durable
materials. The ONWI Task Force reported that some types of acid-free paper may survive a
millennium under reasonable conditions (HUM84). Paper made from cotton or linen fibers
has lasted for 1,000 years. Papyrus had survived for considerably longer in Egypt. The Task
Force recommended that conventional paper would be a suitable storage medium for records
which are periodically updated or maintained, but it proposed that more permanent records be
prepared using special papers and stored in a protected environment. The importance of
incorporating disposal site information on large numbers of maps widely distributed in the
United States and throughout the world was stressed.
The NEA Working Group noted that the principal media currently used for conserving
information include paper, microfilm, and magnetic and optical disks (NEA93). They quoted
a lifetime of 1,000 years for paper, 200-400 years for microfilm (with one regeneration
cycle), and less than 10 years for magnetic and optical media. On this basis, NEA felt that
paper and microfilm were the preferred media for long term storage. The same position was
adopted by the Nordic Committee for Nuclear Safety Research - NKS (JEN93).
12.4 GOVERNMENT OWNERSHIP AND REGULATIONS
12.4.1 General Comments
While government ownership and regulations regarding land and resource use are embraced in
the definition of passive institutional controls, substantive questions have been raised as to the
persistence of such governmental controls over the millennia. Historical continuity of
governments has ranged from days to centuries.
In studying inadvertent human intrusion, SNL convened four expert panels to estimate modes
and likelihoods of future intrusion (HOR91). The teams felt that the likelihood of continued
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U.S. political control over the WIPP was small or non-existent. Changes in government
control can lead to loss of information about a repository. Although physical destruction of
information can be predicted for some scenarios (e.g. war, insurrection, and changes in
record-keeping practices), one possible result of change in governmental control would be a
change in policies regarding the importance of protecting the WIPP site and its records. Even
with continuity of government, there is no guarantee that such continuity can be equated to
continuity of government policies or continuity of government control over the WIPP site.
Governmental policies vary significantly from one administration to the next. A future
government might decide it was no longer necessary to maintain ownership of the site or
update records for changes in language or technology. However, if some future government
makes a conscious decision to take action detrimental to the repository, it is appropriate that
the burden of that decision lie with that future society.
History provides numerous examples of instances where government has changed dramatically
over the centuries, but institutional knowledge has not been lost. This is particularly true
where symbolic or written language is associated with an artifact. The Egyptian pyramids are
a classic example. Built nearly 5,000 years ago by an ancient society that is not reflected in
the Egyptian society of today, located in an area subjugated by many conquerors since,
documented in a symbolic language vastly different than today's languages, and desecrated by
thousands of years of vandalism, the funerary monuments still convey a message understood
by modern society about their design, construction, and function. The NEA Working Group
cited French experience with institutional controls. Louis XVI created the Office of Quarries
in Paris to prevent disturbances to buildings constructed over quarry sites (NEA93). The
Office is still in existence and maintains control over portions of the Paris infrastructure, in
spite of historical trauma including passing from an absolute monarchy to a series of
republican governments with interspersed revolution, civil war, invasion, and riot. Institutions
frequently outlive the governments which inaugurate them.
12.4.2 WIPP Land Withdrawal Act
Under Sec. 3(a) of the LWA, a sixteen square mile area was "withdrawn from all forms of
entry, appropriation and disposal under the public land laws, including without limitation the
mineral leasing laws, the geothermal leasing laws, the material sales laws (except as provided
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in section 4(b)(4) of this Act23), and the mining laws." Jurisdiction over the withdrawn
lands is assigned to the Secretary of Energy. The LWA does not give to the U.S.
Government a right to any water which it did not already possess at the time the LWA was
passed. If the U.S. Government wishes to obtain water rights for purposes associated with the
LWA, it must do so in accordance with the laws of the State of New Mexico.
Sec. 4(b)(5) of the LWA prohibits, in perpetuity, surface and subsurface mining or oil and gas
production, including slant drilling from outside the withdrawal area on the withdrawn lands.
There is one exception to the drilling prohibition. Rights under two existing oil and gas
leases are not affected, although, if dictated by regulatory requirements, the Secretary of
Energy can acquire these leases. As discussed in Section 3.4.1, slant drilling from outside the
withdrawal area into these two leases, which lie in the southwest corner of the WIPP site, is
currently permitted at depths below 6,000 feet (EEG92).
Sec. 13(b) of the LWA requires that DOE prepare, within five years of the date of enactment
(i.e. October 1997), a plan for the management and use of the withdrawn area after
decommissioning. This has yet to be done. However, DOE has prepared a plan for the
management and use of the withdrawn area prior to decommissioning as required by Sec.
4(b)(l) of the LWA (DOE93a).
EPA examined actions planned by DOE in the current Land Management Plan as possible
precursors to future actions associated with the post-decommissioning plan. DOE has
specified that drilling and mining activity within one mile of the WIPP land withdrawal
boundary be monitored by DOE in coordination and cooperation with the BLM and/or the
State of New Mexico. These agencies have agreed to forward to DOE for review and
comment all Applications for Permit to Drill in this boundary zone, together with mining and
reclamation plans. According to the Land Management Plan, "this review will afford DOE
the opportunity to verify that the proposed oil and gas or mining activities surrounding the
withdrawal area will not encroach upon the withdrawn lands." It is possible that if DOE
judges a hole is to be drilled too close to the WIPP site boundary, they could request that the
permit granted to the operator include a condition that DOE be provided with downhole
vertical deviation surveys contemporaneous with the drilling activity. If the surveys detect
subsurface deviation which could encroach upon the WIPP site, the driller could be required
23 Section 4(b)(4) permits the disposal of salt tailings which were produced during mining of the repository but
are not needed for backfill.
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to take corrective measures or cease drilling. Enforcement of such a provision will probably
require continuous monitoring of the drill bit location.
12.5 OTHER METHODS OF PRESERVING DISPOSAL SYSTEM KNOWLEDGE
12.5.1 Subsurface Markers
To provide redundancy in the event that surface markers are destroyed by vandalism, erosion,
other aging processes, or natural disaster, it has been suggested that subsurface markers be
employed to augment surface markers (ADA86, BEN91). If the surface markers are
destroyed or removed, buried markers may still provide a warning to intrusion.
12.5.1.1 Passive Markers
Fired clay subsurface markers were proposed to DOE by DOE contractors specifically for
shallow burial grounds at Hanford based on archeological evidence of the longevity of such
materials (ADA86). In the Hanford study, it was recommended that three layers of
subsurface markers be emplaced at depths of 2, 4 and 16 feet to deter activities such as
farming or building construction on the site. While such subsurface markers might deter
surface excavation, they might not be discovered in the process of exploratory drilling into an
underlying geologic repository. It is highly unlikely that a drilling crew would detect the
presence of the markers. The buried markers would create no impediment to the drilling
process and would probably be destroyed when contacted by the drill bit or missed entirely.
Emplacement of such markers would involve extensive surface excavation.
12.5.1.2 Buried Sensors
Another approach to subsurface markers is to employ buried sensors of various types
(BEN91). These could include buried objects or materials which would create an acoustic,
magnetic, or radioactive anomaly. To create an acoustic anomaly, large granite shapes whose
acoustic signal would define the center of the repository could be buried. Magnetic markers
might include buried iron ore or special high field permanent magnets. Radioactive markers
could be located outside the boundaries to signal the presence of the repository to intruders
entering from various directions. All of these were suggestions made by the Southwest Team
of the Futures Panel.
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Team B of the WIPP Markers Panel suggested that earthworks at the WIPP site could be
spiked with relatively inexpensive, high dielectric constant materials such as metal sulfides or
magnetite which provide a strong radar signal to anyone exploring the site by remote sensing
(TRA93). The ONWI Task Force felt that the site markers themselves would be readily
visible to remote sensors carried by satellite (HUM84).
12.5.2 Protective Barriers
It has been suggested in a report by Ptyalin that the final defensive measure in a defense-in-
depth strategy for reducing the likelihood of human intrusion could be a protective barrier
system (TOL93). In theory, the protective barrier could reduce the likelihood of drilling in
the event that markers and records were lost. The Ptyalin study uncovered no published
research on concepts or actual designs of a protective barrier system. Ptyalin outlined the key
features which a protective barrier system would need to provide. They are as follows:
• "be capable of disabling a drill bit, be impervious to a drill bit or, at a minimum,
be capable of deflecting the drill bit safely away from the disposal system;
• be potentially capable of withstanding multiple encounters with a drill bit without
the loss of function;
• be composed on materials of little economic value and will not degrade over the
10,000 year, post closure period; and
• not attract unwanted attention to the site or encourage exploration activities."
Possible locations for the conceptual protective barriers included at the surface of the
repository site, just beneath the surface, just above the emplaced waste, and within the waste
panels. Combinations of these options are also possible. Location of barriers at or just below
the surface would require significant amounts of materials to cover the 0.5 square kilometer
footprint of the repository. Ptyalin noted that operators have encountered problems in drilling
into old landfills containing layers of rubber tires (at least 10 meters thick), layers of steel
fencing, and baling wire. Encounters with these materials resulted in loss of circulation of the
drilling mud, inability to cut through the materials, and difficulty in removing the drill string
from the borehole. These problems occurred when using small truck-mounted rigs. There is
no information to suggest that similar problems would be encountered when using a large
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stationary rig capable of drilling to depths of 3,000 to 5,000 meters. The long term stability
of such artificial layers is also open to question.
No suitable materials for protective barriers located just above the wastes have been identified.
To reduce effectively, the likelihood of drilling intrusion, the material would need to be
resistant to attack by the drill bit, have corrosion resistance in the repository environment, and
be emplaced with a sufficient areal density to effectively blanket the waste. Advanced
materials such as tungsten carbide composites might meet these specifications. However, such
an approach would be extremely costly. The buried material could be of sufficient economic
value to represent a recoverable resource to some future generation. The impact of further
disruption of the natural geologic barriers to accommodate emplacement of a protective
barrier would need to be addressed.
The same conceptual material problems exist in considering encasing the waste drums with
some sort of armor. While there are materials which might hinder the encroachment of a drill
bit, such materials are costly and they may be an attractive resource to future generations.
Also, their longevity in the repository environment has not been demonstrated.
EPA's review of the many options for PICs considered by DOE and DOE expert panels
supports the approach taken in the rule by clearly demonstrating that while there may be no
definite proof of the effectiveness or survivability of a particular PIC system, there are many
approaches with positive characteristics and many proposals that seem to hold promise for
improving the WIPP disposal system and thus protecting the public health and the
environment.
12.6 REFERENCES
ADA86 Adams, M.R. and M.F Kaplan, "Marker Development for Hanford Waste Site
Disposal," Paper presented at Waste Management 86, Volume 1, pp. 425-431,
March 2-6, 1986.
AST92 Ast, D.G., M. Brill, W. Goodenough, M. Kaplan, F. Newmeyer, and W.
Sullivan, "Marking the Waste Isolation Pilot Plant for 10,000 Years" (Team A
report under TRA93), April 1992.
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BAK92 Baker, V.R., F.D. Drake, B.R. Finney, D.B. Givens, J. Lomberg, L.E. Narens,
and W.S. Williams, "The Development of Markers to Deter Inadvertent Human
Intrusion Into the Waste Isolation Pilot Plant (WIPP)" (Team B report under
TRA93), April 22, 1992.
BAK94 "Private Communication from Mark Baker," U.S. Army Corps of Engineers,
Baltimore Division, April 5, 1994.
BEN91 Benford, G., et al., "Ten Thousand Years of Solitude? On Inadvertent Intrusion
into the Waste Isolation Pilot Project Repository," LA-12048-MS, Los Alamos
National Laboratory, March 1991.
BRA79 Brausch, L.M. and D.M. Winters, "Study Report - Assessment of Waste
Isolation Periods Prior to Access by Exploratory Drilling in Geologic Waste
Repositories," DOE/AL/05346--T5, WIPP Technical Support Contractor, July
31, 1979.
CHA84 Chalapka, George, "From Palaeoart to Casual Paintings: The Chronological
Sequence of Arnhem Land Plateau Rock Art," Monograph Series 1, Northern
Territory Museum of Art and Sciences, Darwin 1984.
DAV80 Davis, Emma Lou, Kathryn H. Brown, and Jacqueline Nichols, "Evaluation of
Early Human Activities and Remains in the California Desert," U.S.
Department of the Interior, Bureau of Land Management, California Desert
District: Cultural Resources Publication, Anthropology-History, 1980.
DQE81 U.S. Department of Energy, "Gnome Site Decontamination and
Decommissioning Project: Radiation Contamination Clearance Report March
28, 1979 - September 23, 1979," DOE/NV/00410-59, Reynolds Electrical &
Engineering Co., Inc., August 1981.
DOE93a U.S. Department of Energy, "Waste Isolation Pilot Plant Land Management
Plan," DOE/WIPP 93-004, 1993.
DOE93b U.S. Department of Energy-Nevada Operations Office, "Remedial Investigation
and Feasibility Study of the Tatum Dome Test Site, Lamar City, Mississippi",
vol.1, Submitted to Mississippi Office of Pollution Control, Hazardous Waste
Division, September 1993.
DOE94a Communication with Doug Duncan, O/AMEM, DOE/NV, Las Vegas, NV,
January 14, 1994.
DOE94b Communication with R. Navarro, Nevada Operations Office, Las Vegas, NV,
January 14, 1994.
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DUR54 Durant, Will, "Our Oriental Heritage," Simon and Schuster, 1954.
EEG92 Environmental Evaluation Group, "Implications of Oil and Gas Leases at the
WIPP Site on Compliance with EPA TRU Waste Disposal Standards," EEG-50,
M.K. Silva and J.K. Channell, June 1992.
GIL85 Gillis, D., "Preventing Human Intrusion into High-Level Nuclear Waste
Repositories," Underground Space, Vol. 9, pp. 51-59, 1985.
GIV82 Givens, David B., "From Here to Eternity: Communicating with the Distant
Future," Et Cetera, pp. 159-179, Summer 1982.
GOG87 "Stockage des dechets radioactifs en formations geologique Criteres
techniques de choix de site," Rapport du groupe de travail preside par le
Professeur Goguel, Ministere de 1'Industrie, June 1985-May 1987.
GRA93 Gray, Denis D., "Champion of Aboriginal Art," Archaeology 46:4:45-47,
Archaeological Institute of America.
HED83 Hedges, Ken, Editor, "Rock Art Papers Vol. 1," San Diego Museum Papers No.
16, San Diego Museum of Man, San Diego, 1983.
HOR91 Hora, S.C. et al., "Expert Judgement on Inadvertent Human Intrusion into the
Waste Isolation Pilot Plant," SAND90-3063, Sandia National Laboratories,
December 1991.
HUD78 Hudson, Travis and Earnest Underhay, "Crystals in the Sky: An Intellectual
Odyssey Involving Chumash Astronomy, Cosmology, and Rock Art," Ballena
Press Anthropological Papers No. 10, Socorro, NM, 1978.
HUD79 Hudson, Travis, Georgia Lee, and Ken Hedges, "Solstice Observers and
Observatories in Native California," Journal of California and Great Basin
Anthropology 1:39-63.
HUM84 Human Interference Task Force, "Reducing the Likelihood of Future Human
Activities That Could Affect Geologic High Level Waste Repositories,"
BMI/ONWI-537, May 1984.
JEN93 Jensen, Mikael, "Conservation and Retrieval of Information - Elements of a
Strategy to Inform Future Societies about Nuclear Waste Repositories,"
Nordiske Seminar og Arbejdrapporter 1993:596, August 1993.
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JUD50 Papers of Neil M. Judd, Box 3: Manuscripts of Writings: "American
Petroglyphs," Paper No. 61, submitted 5/9/50 for a volume honoring Dr. Gustat
Hailstorm on his 70th anniversary. National Anthropological Archive, Natural
History Museum, Smithsonian Institution, Washington DC.
KAP82 Kaplan, M.F., "Archeological Data as a Basis for Repository Marker Design,"
BMI/ONWI-354, The Analytic Sciences Corporation, October 1982.
KAP86 Kaplan, M.F and M. Adams, "Using the Past to Protect the Future,"
Archaeology, pp. 51-54, September-October 1986.
LEW93 Lewin, Roger,"Paleolithic Paint Job," Discover 14:7: 64-70, July 1993.
MAL65 Mallowan, M.E.L., "Early Mesopotamia and Iran," McGraw-Hill, 1965.
MUR87 Murphey, Kellya, "Rock Art at the Kanaka-Briggs Creek Locality (10 GG 307),
Gooding County, Idaho," Journal of California and Great Basin Anthropology
9:1:74-99, 1987.
NAT95 National Research Council, Technical Basis for Yucca Mountain Standards,
National Academy Press, Washington D.C., 1995.
NEA93 Nuclear Energy Agency, "Assessment of Future Human Actions at Radioactive
Waste Disposal Sites," Draft report of an NBA Working Group, 92303b-2,
compiled by Daniel A. Galson, October 4, 1993.
NEA95 Nuclear Energy Agency, "Future Human Actions at Disposal Sites, A report of
the NEA Working Group on Assessment of Future Human Actions at
Radioactive Waste Disposal Sites," Paris, 1995.
NGS75 "Mystery Surrounds the Biggest Planet," K.F. Weaver, in National Geographic,
Vol. 147, No. 2, February 1975.
NGS90 "Neptune - Voyager's Last Picture Show," Rick Gore, in National Geographic,
Vol. 178, No. 2, August 1990.
RAI93 Raimbault, P. and C. Valentin-Ranc, "How to Mark Repositories in Geological
Formation?," Agence Nationale pour la Gestion des Dechets Radioactifs
(ANDRA), in Proceedings of SAFEWASTE '93 (Avignon, France, June 13-18,
1993), Vol. 3, pp. 212-221, SFEN, Paris, 1993.
RAV85 Raven, Christopher, "A Stone Alignment in the Northern Great Basin with a
(Probably) Coincidental Astronomical Orientation," Journal of California and
Great Basin Anthropology 7:1:89-98, 1985.
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ROC77 Rochlin, G.I., "Nuclear Waste Disposal — Two Social Criteria," in Science,
Vol. 195, No. 4273, pp. 23-31, 1977.
SAN92 Sandia National Laboratories, "Preliminary Performance Assessment for the
Waste Isolation Pilot Plant," SAND92-0700/1, December 1992.
SEB84 Sebeok, T.A., "Communication Measures to Bridge Ten Millennia,"
BMI/ONWI-532, Indiana University, April 1984.
TAN84 Tannenbaum, P.H., "Communication Across 300 Generations: Deterring
Human Interference with Waste Deposit Sites," BMI/ONWI-535, University of
California, Berkeley April 1984.
TOL93 Tolan, T.L., "The Use of Protective Barriers to Deter Inadvertent Human
Intrusion into a Mined Geologic Facility for the Disposal of Radioactive Waste:
A Review of Previous Investigations and Potential Concepts," SAND91-7097,
Tolan, Beeson, and Associates, June 1993.
TRA93 Trauth, K., et al., "Expert Judgment on Markers to Deter Inadvertent Human
Intrusion into the Waste Isolation Pilot Plant," SAND92-1382, Sandia National
Laboratories, November 1993.
TRA94 Conde'Nast Traveler. March 1994.
TUR94 Turpin, Solveig A., "Beneath the Sands of Time: Archaic Petroglyphs Mark a
Sacred Spot on the Ritual Landscape of Western Texas," Archaeology 47:2:50-
53, 1994.
USA72 USAF Radioisotope Committee, "Burial of Radioactive Waste in the USAF,"
Wright-Patterson AFB, 15 March 1972.
WOO63 Wooley, Leonard, "Digging Up the Past," Penguin 1963.
12-72
-------
Appendix 12A:
Federal Register Notice Identifying WIPP Land Withdrawal Area
-------
federal Rogktar / VcL 57. No. 327 / Tuesday. November 24. 1392 / frJodoes SS277
(NM-fl20-4210-06: NMNM 5S234]
L«ga1 Description for ttw Wa«t«
Isolation Pttot Plant (WPP)
Withdrawal; KM
AGENCY: Bureau
(BLM). interior.
ACnow: Notice of Legal Description for
thc WIPP WithdravnL
SUMMARY: On October 30, 19S2. Public
Law (PL) 102-573. Sha Wasie Isolation
Pilol PLani Land Wtthdr-awaJ.Act WAS
signed into iaW. Section 3 of Public LJ»VV
1QI^S79 requires that u'ltbra 30 days
afterth«d«te€>fth£ eivactoientof the
Act, the Secretary of interior sfaaU
publish dn the federaJ Register a ivoiioc
containrcg a iega! riescxiprioa of the
Withdrawal XhisNaticE ooatatrxs tbat
legal descriptjon.
f OR FURTHER TNFOflMATTDH CONTACT:
Clarence F.Houglawd BLM. New
Mexico Staie Office, 50S-?3e-7o93.
; Purstiant
to Section 3 of PL 102-579 the legal
description for tfa« WiPP WUhrira wai is
as follows:
New Mexico Mndpal Meridian
T.22S-.K-.MEL..
Sect, 15 to .17 Inclusive;
Sec. 18. lots 1 to 4. inclusive. EV&. acd
Sec. 19, lots 1 to 4. inclusive. En/i,
Sect 20 to 22 inclusive;
Sees. ZT lo 29 inclusive;
Sec. '30. lots 1 to 4. inclusive. E'A. and
Sec. 31. hrts 1 to 4. incfesivc. EV4. and
EV4WV4:
Sacs. 32 to 34 inclusive.
The area described contain* MX244-60
acres, jnora or J«»g. in Eddy County, New
Mexico.
Dated: November 13. 1992.
Monte G. Jordan.
Associate State Director-
[FR Doc. 92-2S458 Filed 11-2&-92: 8;45 am)
BJLUWJCOOt OtO-TD-d
-------
Appendix 12B:
Letter to U.S. Archivist Transmitting WIPP Land Withdrawal Information
-------
United States Department of the Interior
BUREAU OF LAND MANAGEMENT
NEW MEXICO STATE OFFICE
1474 RODEO RD-
P.O. BOX 27115
SANTA FE. NEW MEXICO 87502-7 US
IN B£PLY REFER TCK
NMNM 55234
2310 (923)
NOV ! 6 1992
Dr. Don Wilson
Archivist of the United States
National Archives
Rm. Ill
Washington, B.C. 20408
Dear Dr. Wilson:
In accordance with Section ? ot Public taw IBK-U'ra, the Waste ieetat'iaiH
Plant Withdrawal Act, enclosed are the map and the legal description for the
Waste Isolation Pilot Plant Withdrawal.
Section 3 requires that ccpies of the map and the legal description for the
Withdrawal be filed with the Congress, the Secretary of Energy, the Governor
of the State, and the Archivist of the United States.
Please contact Clarence Hougland at (505) 438-7593, if you have questions or
need further assistance.
Siiv&erely,
Larry L. Wdodard
State Director
2 Enclosures:
1 Withdrawal Map
2 - Legal Description (1 p)
-------
13. Engineered Barriers
13.1 REGULATORY BACKGROUND
13.1.1 Environmental Protection Agency Regulations
The Assurance Requirements contained in Subpart B of 40 CFR part 191 require, under
§191.14(d), that
Disposal systems shall use different types of barriers to isolate wastes from the
accessible environment. Both engineered and natural barriers shall be included.
The disposal standards (§191.12) define a "barrier" as
any material or structure that prevents or substantially delays movement of water or
radionuclides toward the accessible environment. For example, a barrier may be a
geologic structure, a canister, a waste form with physical and chemical characteristics
that significantly decrease the mobility of .radionuclides, or a material placed over and
around the waste, provided that the material or structure substantially delays
movement of water or radionuclides.
Use of barriers was also specified in the WIPP LWA. Section 9 of the Act — Environmental
Protection Agency Disposal Regulations — specifies hi §9(g) — Engineered and Natural
Barriers, Etc, ~ that DOE shall use both engineered and natural barriers and waste form
modifications to isolate the waste after disposal to the extent necessary to comply with
Subpart B of 40 CFR part 191. The Act defines "engineered barriers" .to mean backfill,
room seals, panel seals, and any other manmade barrier components of the disposal system.
As is the case for all Assurance Requirements, the intent of the requirement for engineered
barriers is to enhance the WIPP's long-term compliance with the Containment Requirements
(§191.13). While Engineered Barriers are required pursuant to the Assurance Requirements
of §194.14(d), Engineered Barriers are not necessarily required to meet the Containment
Requirements of §191.13. However, unlike most other Assurance Requirements, which
provide some qualitative measure of increased confidence in the ability of the disposal system
to do its job for the 10,000 year regulatory period, natural and engineered barriers are an
inherent part of the disposal system. Thus, quantitative performance of the disposal system
is evaluated by examining the disposal system as a whole. The effects of Engineered
13-1
-------
Barriers employed at the WIPP must be considered as performance assessments; excluding
such barriers would result in inaccurate modeling of the disposal system as defined in
§191.12(a).
In 40 CFR part 194, EPA reiterates that engineered barriers are required as originally
specified in §191.14(d). To ensure that a defensible position on the assurance aspects of
engineered barriers is developed, EPA requires that DOE evaluate the benefits and
detriments of various engineered barrier alternatives, such as cementation, shredding,
supercompaction, incineration, vitrification, improved waste canisters, grout and bentonite
backfill, melting of metals, alternative configurations of waste placement in the disposal
system, and alternative disposal system dimensions. The potential benefit of the engineered
barrier alternatives would be the prevention or substantial delay of movement of water or
radionuclides toward the accessible environment. Potential detriments might include
increased worker exposure involved in barrier implementation, increased total system costs,
and significant program delays. The DOE application for certification of compliance must
include justification for the selection or rejection of each type of engineered barrier
evaluated.
Waste inventory scheduled for disposal at the WIPP is hi a state of flux. Some waste is in
packages which presumably meet the waste acceptance criteria (WAC) for the repository;
some of the existing waste must be repackaged to meet the WAC; some of the existing waste
must be treated to meet the WAC; and a significant portion of the waste has not been
generated (BIR95). Consequently, DOE is required in its benefit/detriment study of
engineered barrier alternatives to separately consider wastes in various packaging states.
13.1.2 Nuclear Regulatory Commission Regulations
The NRC has promulgated regulations for "Disposal of High-Level Radioactive Wastes in
Geologic Repositories" as 10 CFR part 60. The NRC definition of "barrier", i.e.. "any
material or structure that prevents or substantially delays movement of water or
radionuclides", is very similar to the EPA definition. In addition, the NRC rule also defines
"engineered barrier system" as the waste packages and the underground facility.
"Underground facility" is a defined term meaning underground structure, including openings
and backfill materials, but excluding shafts, boreholes, and their seals. The NRC specifically
excludes shaft seals as an element of the engineered barrier system, while EPA regulations
13-2
-------
and the WIPP LWA do not specifically mention shaft seals. Shaft seals could, by inference,
be included as "any other manmade barrier components of the disposal system" in the
engineered barriers definition hi the WIPP LWA.
The NRC license application for such a repository requires a Safety Analysis Report (SAR),
which includes an analysis of the effectiveness of natural and engineered barriers against the
release of radioactive material to the environment (§60.21(c)(l)(ii)(D)). This analysis must
incorporate a comparison of the effectiveness of alternatives to the major design features
affecting waste isolation, with emphasis on those features which provide longer radionuclide
containment and isolation. The NRC rule further requires that the waste packages provide
"substantially complete" containment of the high-level waste for a period of 300 to 1,000
years. The release rate from the engineered barrier system after this containment period for
any radionuclide is limited to one part per 100,000 per year, based on the amount of the
nuclide present 1,000 years after the repository is closed (§60.113(a)(l)(ii)(B)). The NRC
regulations are designed to address the containment of high-level wastes, which pose
somewhat different containment issues than transuranic wastes.
13.2 CONSIDERATION OF ENGINEERED ALTERNATIVES
13.2.1 Engineered Alternatives Task Force
In 1989, DOE formed an Engineered Alternatives Task Force (EATF) whose objectives were
to identify plausible engineering modifications to the existing WIPP design, and to evaluate
the feasibility and effectiveness of these modifications in facilitating compliance of the WIPP
with EPA disposal standards contained in 40 CFR part 191 (DOE91a). Potential repository
problems addressed by the EATF included gas generation by the waste and consequences of
future, inadvertent human intrusion. DOE was concerned that these problems might interfere
with the WIPP's compliance with the containment requirements of 40 CFR part 191. The
EATF activities were not designed to address the assurance aspects of engineered barriers.
The first step in the EATF methodology was to identify and screen potential engineered
alternatives. To accomplish this, an Engineered Alternatives Multidisciplinary Panel
(EAMP) was formed by assembling a group of experts with relevant backgrounds. The
EAMP met in late 1989 and early 1990 to conduct identification and screening activities. A
13-3
-------
total of 64 alternatives were identified for initial consideration as summarized in Table 13-1
taken from the EATF Final Report (DOE91a).
The EAMP's list of potentially useful alternatives was distilled down to 14 alternative
scenarios which considered various combinations of waste treatments, backfill options, waste
container changes, waste emplacement options, and facility design changes. Recognizing that
all waste is not amenable to the same treatment option, EAMP categorized wastes as sludges,
solid organics, and solid inorganics for the study. Each alternative was compared with a
baseline which assumed no waste form modifications, no container modifications, no load
management of the wastes, no facility design changes, and use of a salt backfill. The
alternatives are summarized in Table 13-2.
Because the processes which can affect the WIPP are often coupled and non-linear, it is
difficult to assess the impact of various engineered alternatives by inspection and logic.
Consequently, DOE developed the Design Analysis Model to assist in the quantitative
assessment of alternatives. The Design Analysis Model, which is deterministic rather than
probabilistic, modeled the following processes:
• creep closure of the surrounding rock
• gas generation, consumption, and dispersion
• brine inflow, consumption, and dispersion
• panel seal leakage
• consolidation of the shaft seals and advection of gas and brine through the
shaft seals
• diffusion and advection of gases into the host rock and adjacent anhydrite beds
• gas compressibility
• waste compaction
• development of a disturbed rock zone around the storage rooms, and
• radionuclide releases caused by three types of inadvertent human intrusion
scenarios into the repository.
The peak gas pressure reached in the repository was the figure of merit used to assess the
effect of various engineered alternatives on gas generation potential. Gas pressures
calculated with the Design Analysis Model exceeded lithostatic pressure (i.e., the pressure of
the surrounding host rock) in the base case and for Alternatives 1, 2, 3, 10, 11, 12, and 14.
Lithostatic pressures were not exceeded for the other seven alternatives. For each of those
13-4
-------
Table 13-1
Potentially Useful Engineered Alternatives Considered By the Engineered Alternatives
Multidisciplinary Panel (EAMP) (From DOE91a)
Waste Form Modification Alternatives
Compact Waste
Incinerate and Cement
Incinerate and Vitrify
Wet Oxidation
Shred and Bituminize
Shred and Compact
Shred and Cement
Shred and Polymer Encapsulation
Shred, Add Salt, and Compact
Plasma Processing
Melt Metals
Add Salt Backfill
Add Other Sorbents
Add Gas Suppressants
Shred and Add Bentonite
Acid Digestion
Sterilize
Add Copper Sulfate
Add Gas Getters
Add Fillers
Segregate Waste Forms
Decontaminate Metals
Change Waste Generating Process
Add Anti-Bacterial Material
Accelerate Waste Digestion Process
Alter Corrosion Environment in WIPP
Alter Bacterial Environment hi WIPP
Transmutation of Radionuclides
Vitrify Sludges
Backfill Alternatives
Salt Only
Salt Plus Gas Getters
Compact Backfill
Salt Plus Brine Sorbents
Preformed Compacted Backfill
Grout Backfill
Bitumen Backfill
Add Gas Suppressants
Waste Management Alternatives
Minimize Space Around Waste Stack
Segregate Waste In WIPP
Decrease Amount of Waste Per Room
Emplace Waste and Backfill
Simultaneously
Selective Vegetative Uptake
Facility Design Alternatives
Brine Isolate Dikes
Raise Waste Above the Floor
Brine Sumps and Drains
Gas Expansion Volumes
Seal Disposal Room Walls
Vent Facility
Ventilate Facility
Add Floor of Brine Sorbents
Change Mined Extraction Ratio
Change Room Configuration
Seal Individual Rooms
Two Level Repository
Passive Marker Alternatives
Monument Forest Over Repository
Monument Covering the Entire Repository
Buried Steel Plate Over Repository
Artificial Surface Layer Over Repository
Add Marker Dye to Strata
Miscellaneous Alternatives
Drain Castile Reservoir
Grout Culebra Formation
Increase Land Withdrawal Area to
Regulatory Boundary
Waste Container Alternatives
Change Waste Container Shape
Change Waste Container Material
13-5
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Table 13-2
Engineered Alternatives Evaluated by the EATF Relative to the Baseline Case (From DOE91a)
Alternative #
Baseline
Alternative 1
Alternative 2
Alternative 3
Alternative 4
Alternative 5
Alternative 6
Alternative 7
Alternative 8
Alternative 9
Alternative 10
Alternative 11
Alternative 12
Alternative 13
Alternative 14
Sludges
As received
As received
Cement
Cement
Cement
Cement
Vitrify
Vitrify
Vitrify
Vitrify
As received
As received
As received
Vitrify
As received
Solid Organics
As received
Shred/Cement
Shred/Cement
Shred/Cement
Incin/Cement
Incin/Cement
Incin/Vitrify
Incin/Vitrify
Incin/Vitrify
Incin/Vitrify
As received:
Less Metals
Supercompact
Supercompact
Incin/Vitrify
Supercompact
Solid Inorganics
As received
Shred/Cement
Shred/Cement
Shred/Cement
Shred/Cement
Shred/Cement
Melt metals*
Melt metals*
Melt metals**
Melt metals**
Decontaminate
Metals***
Supercompact
Supercompact
Melt metals**
Supercompact
Backfill
Salt
Salt
Salt
Cement grout
Salt
Cement grout
Salt
Cement grout
Salt
Cement grout
None
Salt
Cement grout
None
Salt aggregate
Grout
Waste Container
As received
As received
As received
As received
As received
As received
As received
As received
Non-ferrous
Non-ferrous
Non-ferrous/
Rectangular
As received
As received
Non-ferrous/
Rectangular
As received
Waste Management
As designed
As designed
As designed
As designed
As designed
As designed
As designed
As designed
As designed
As designed
Minimize space around
waste
Single layer: 2000
drums
Single layer: 2000
drums
Minimize space around
waste
Compartmentalize
waste, 2000 drums per
room
Facility Design
As designed
As designed
As designed
As designed
As designed
As designed
As designed
As designed
As designed
As designed
New dimensions:
10'x31'xl88'
New dimensions:
6'x33'x300'
New dimensions:
6'x33'x300'
New dimensions:
10'x31'xl88'
Salt dikes:
Waste Separation
Metals are melted into TRU waste ingots.
Metals are melted with glass/glass frit; radionuclides partition into the slag, and metals are eliminated from the WIPP inventory.
Metals are decontaminated by vibratory finishing and eliminated from the WIPP inventory.
-------
alternatives where lithostatic pressure was not exceeded, rigorous thermal processing
techniques (incineration, metal melting, and/or vitrification) were assumed to be used to
modify the waste form.
To assess the impact of the engineered alternatives on human intrusion, the figure of merit
used was the "Measure of Relative Effectiveness" (MRE) which compares the cumulative
releases of selected radionuclides for each alternative to the releases from the baseline
design. Lower values of the MRE are indicative of improved performance as compared to
the baseline. For an El scenario, where a borehole penetrates the repository and a brine
pocket in the underlying Castile Formation, Alternatives 7, 8, 9, and 13 in Table 13-2 were
effective hi reducing the consequences of inadvertent human intrusion. As noted above,
these same alternatives were effective in reducing gas pressures as well, but involved
rigorous thermal processing.
For the E2 scenario, where a borehole penetrates the repository but not an underlying brine
reservoir, Alternatives 3, 5, 6, 7, 12, 13, and 14 were effective in reducing the consequences
of human intrusion. Of these attractive alternatives, Alternative 3 is the probably the
simplest since it only requires cementing or shredding and cementing of the waste and a
cement grout backfill.
For the E1E2 scenario, where two boreholes penetrate the same panel in the repository and
one also penetrates an underlying Castile brine reservoir, all of the alternatives except No. 11
were efficacious. Alternative No. 11 was unattractive for all conditions examined.
If some type of waste form modification involving thermal processing is to be considered for
TRU wastes, there are major cost and schedule implications. Metal melting and incineration
have been practiced on low-level waste, and incineration to a limited extent, on TRU waste.
Vitrification has not been fully developed and reduced to practice for routine waste
processing. Substantial periods of time (probably at least a decade) are required to design
thermal treatment facilities for TRU wastes, obtain budgetary approvals, obtain the required
environmental permits, construct the facilities, and conduct extensive startup tests before
waste processing can begin.
If such rigorous alternatives are not necessary to demonstrate regulatory compliance, then
there may be other, easier to implement, alternatives which may satisfy the assurance
13-7
-------
requirements of 40 CFR part 191. For example, with alternative backfills it may be possible
to control the brine pH, thereby minimizing radionuclide solubility or narrowing the range of
expected solubilities. Backfill may also serve as a vehicle for carbon dioxide removal from
the disposal rooms, thus reducing gas pressure buildup in the repository.
13.2.2 Engineered Barriers Study for §194.44
The 40 CFR part 194 compliance criteria specify that an evaluation of engineered barrier
alternatives be included hi the application for certification of compliance. The rule also
specifies a minimum number of alternatives which must be considered. These are basically
the same alternatives as DOE examined under the aegis of the EATF study in 1991.
However, it should be noted that the EATF study combined the various individual
alternatives into the 14 summary alternatives listed hi Table 13-2. Consequently, it
was not possible to distill from that study the effects of individual alternatives. For example,
any changes associated with alternative container materials are probably masked by changes
hi other alternatives which were simultaneously considered.
Whereas the EATF study focused on reduction of gas generation and consequences of
inadvertent human intrusion, §194.44 requires a broader look at engineered barrier
alternatives, including:
• ability of the barrier to prevent or substantially delay movement of water or
waste
• altered worker exposure
• ability to remove waste from the repository
• transportation risks
• uncertainty in performance assessments
• public input
• impact on other waste disposal programs
• system costs, and
• mitigation of human intrusion consequences.
It should also be noted that the EATF study did not include remote-handled (RH) TRU
wastes. Only contact-handled (CH) TRU was considered. While RH TRU wastes account
for only about 4% of the total repository design volume, it is currently estimated that RH
TRU accounts for about 37% of the total radioactivity in the repository (BIR95).
Additionally, RH TRU underlies about 11% of the surface area of the repository that might
13-8
-------
be intercepted by inadvertent human intrusion. Clearly, RH TRU needs to be considered in
an engineered barriers evaluation.
13.2.3 Waste Inventory
Selection of an appropriate waste form modification is dependent on the nature of the waste
to be treated/modified. Use of the same treatment technology for modification of all wastes
is probably not possible. For example, organic waste streams may be amenable to
incineration, but this technology would be inappropriate for heterogeneous and metal wastes.
DOE sites which generate or store transuranic (TRU) waste have identified about 360
different TRU waste streams. Based on their physical/chemical matrix, these waste streams
are assigned waste matrix codes (WMCs), and WMCs with similar physical and chemical
properties are grouped into the 11 waste matrix code groups (WMCGs) listed below:
• Solidified Inorganics
• Salt Waste
• Solidified Organics
• Soils
• Uncategorized Metal
• Lead/Cadmium Metal Waste
• Inorganic Non-Metal Waste
• Combustibles
• Graphite
• Heterogeneous
• Filters
Quantities of waste in each waste matrix code group have been estimated in the W1PP
Baseline Inventory Report (BIR95). These inventory quantities are based on retrievably
stored TRU waste currently located at each site, and projections of future volumes of waste
which have not yet been generated. If the volumetric sum of the stored waste plus projected
waste volumes is less than the repository capacity, the projected volumes are scaled upward
to obtain the anticipated volume (i.e., the additional volume which would fill the repository
to capacity). Estimates of the WIPP inventory for both contact-handled and remote-handled
TRU waste by waste matrix code group are included in Table 13-3. The anticipated volumes
are scaled based on the assumption that the CH TRU capacity of the WIPP is 6.2 million
13-9
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Table 13-3
Transuranic Waste Disposal Inventory for WIPP
(Volumes in Cubic Meters) (From BIR95)
Waste Matrix Code Group
Contact Handled Waste
Combustible
Filter
Graphite
Heterogeneous
Inorganic Non-metal
Lead/Cadmium Metal Waste
Salt Waste
Soils
Solidified Inorganics
Solidified Organics
Uncategorized Metal
Unknown
Total CH Volumes
Remote Handled Waste
Combustible
Filter
Heterogeneous
Lead/Cadmium Metal Waste
Salt Waste
Solidified Inorganics
Uncategorized Metal
Unknown
Total RH Volumes
Total TRU Waste Volumes
Stored
Volumes
7.1E+03
4.3E+02
6.7E+02
3.0E+04
1.2E+03
5.6E+01
3.3E+01
3.7E+02
1.7E+04
1.5E+03
•1.2E+04
1.7E+03
7.3E+04
1.5E+01
8.9E-01
4.4E+02
O.OE+00
O.OE+00
6.1E+02
8.8E+01
1.1E+01
1.2E+03
7.4E+04
Projected
Volumes
2.7E+04
1.1E+03
4.3E+01
4.6E+03
3.2E+02
1.3E+02
6.0E+01
4.5E+02
8.0E+03
3.0E+02
8.6E+03
O.OE+00
5.1E+04
3.2E+00
2.1E+00
3.3E+03
6.0E+00
2.8E+00
1.7E+02
8.6E+01
2.4E+01
3.6E+03
5.4E+04
Anticipated
Volumes
3.4E+04
1.5E+03
7.1E+02
3.5E+04
1.5E+03
1.8E+02
9.2E+01
8.3E+02
2.5E+04
1.8E+03
2.1E+04
1.7E+03
1.2E+05
1.8E+01
3.0E+00
3.8E+03
6.0E+00
2.8E+00
7.9E+02
1.7E+02
3.5E+01
4.8E+03
1.3E+05 ,
WIPP Disposal
Volumes
6.2E+04
2.6E+03
7.6E+02
3.9E+04
1.8E+03
3.1E+02
1.5E+02
1.3E+03
3.4E+04
2.1E+03
3.0E+04
1.7E+03
1.8E+05
2.0E+01
4.3E+00
5.9E+03
9.8E+00
4.6E+00
9.0E+02
2.3E+02
3.5E+01
7.1E+03
1.8E+05
13-10
-------
cubic feet (176,000 m3) and the RH TRU capacity is 250,000 cubic feet (7,080 m3).1
§194.44 requires that the benefit and detriment of engineered barriers be examined separately
for:
• existing waste already packaged
• existing waste requiring repackaging
• existing waste not yet packaged, and
• to-be-generated waste.
Table 13-3 shows that about two-thirds of the anticipated volumes of both the CH TRU and
RH TRU waste is yet to-be-generated. With regard to existing wastes, it should be noted
that for wastes to be acceptable for shipment to the WIPP, they must be packaged either in
55-gallon drums or standard waste boxes (SWBs) that meet DOT Type A packaging
requirements (DOE91b). In addition, some wastes in 55-gallon drums or SWBs may not
meet the current WIPP WAC. For example, a number of waste matrix codes involve liquid
waste streams which must be solidified to ensure that the wastes contain no more than 1 %
free liquid. While these wastes are listed as solidified organics or inorganics in the BIR,
they must be treated and repackaged to comply with the WIPP WAC (DOE91b).2 As
specified hi §194.44(d), these types of existing wastes must be examined separately from
existing wastes which are currently certifiable for shipment to WIPP.
A significant fraction of the waste planned for disposal at the WIPP is mixed waste
containing both hazardous and radioactive components. DOE has estimated that about
60;000 m3 of mixed TRU waste is currently in inventory or will be generated over the next
five years (DOE94). Of this volume, about 20,000 m3 is expected to meet the WIPP WAC
without further treatment, while the balance will require additional treatment before shipment
to the WIPP.
'This exceeds the WIPP LWA limit of 6.2 million ft3 on the total of RH and CH TRU for disposal at the
WIPP.
2The most recent version of the WIPP WAC is 1991. It is not clear at this time what changes DOE will
make to the existing, and possibly dated, waste acceptance criteria to ensure compliance.
13-11
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13.3 REFERENCES
BIR95 "Waste Isolation Pilot Plant Transuranic Waste Baseline Inventory Report,"
CAO-94-1005, Revision 1, prepared by WIPP Technical Assistance Contractor
for U.S. Department of Energy, February 1995.
DOE91a U.S. Department of Energy, "Evaluation of the Effectiveness and Feasibility
of the Waste isolation Pilot Plant Engineered Alternatives: Final Report of the
Engineered Alternatives Task Force," DOE/WIPP 91-007, Revision 0, July
1991
DOE91b U.S. Department of Energy, "Waste Acceptance Criteria for the Waste
Isolation Pilot Plant," DOE/WIPP-069, Revision 4, 1991.
DOE94 U.S. Department of Energy, "National Summary Report of Draft Site
Treatment Plans," vol. 1, final draft, November 1994.
13-12
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14. Individual and Groundwater Protection Requirements
14.1 INTRODUCTION
Both the individual and groundwater protection requirements of the disposal regulations — 40
CFR part 191 subparts B and C — apply to doses received from the wastes in the disposal
system assuming that it is not disrupted by the occurrence of human intrusion or unlikely
natural events. Specifically, the individual protection requirements at §191.15 limit the
annual committed effective dose of radiation to any member of the public to no more than 15
millirem. The ground-water protection requirements §191.24 of subpart C, which limit
releases to ground water to no more than the limits set by the MCL for radionuclides
established in 40 CFR part 141. Both are concerned with human exposure to radionuclides
from disposal systems and both limit such exposure for 10,000 years. Based on the similar
forms of the two numerical requirements, EPA decided to adopt an approach that combines
compliance criteria for these requirements into one section addressing the following issues:
• the definition of a protected individual,
• consideration of exposure pathways,
• consideration of underground sources of drinking water,
• the scope of compliance assessments, and
• the basis for a determination of compliance with these requirements (results of
compliance assessments).
14.2 GROUNDWATER PROTECTION
The groundwater protection requirements of 40 CFR part 191 apply to USDWs in the
accessible environment. Those USDWs that lie within the controlled area are not considered
to be protected groundwater and the requirements of subpart C of 40 CFR part 191 do not
apply. In 40 CFR part 194, the Agency implemented the requirements of subpart C of 40
CFR part 191 with the expectation that USDWs which lie closer to the disposal system will
have a greater chance of being affected by releases of waste. In view of this, the analysis of
the doses received from USDWs located large distances from the disposal system would not
be likely to reveal information about the disposal system's performance not already disclosed
14-1
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by the analysis of those USDWs proximal to the disposal system. As a result, the
groundwater protection requirements as implemented for the WIPP in §194.52 apply to those
USDWs in the accessible environment that are expected to be affected by the disposal system
over the regulatory time frame. The determination of which USDWs are expected to be
affected shall be based upon the underground interconnections among bodies of surface
water, ground water, and underground sources of drinking water.
Additionally, since the MCLs are applied equally to all USDWs, the "maximally exposed"
aquifer will be determinative of compliance with the groundwater protection requirements.
In other words, if the maximally exposed USDW is in compliance, then those lesser exposed
USDWs, perhaps lying further from the disposal system, will likely be in compliance as
well.
The Agency established the definition of the underground source of drinking water (USDW)
in the promulgation of 40 CFR part 191 in 1993. The definition of USDW is taken directly
from the Agency's underground injection control regulations found in 40 CFR parts 144
through 146. The complete description of the definition of USDW and the rationale which
underpins it may be found in the Federal Register noticed which promulgated 40 CFR part
191, found at 58 Fed. Reg. 66398-66416.
In addition to considering interconnected USDWs, 40 CFR part 194 requires that the
calculations of doses received from USDWs should assume that drinking water is withdrawn
directly from the contaminated USDW and consumed at a rate of two liters per day. This
requirement re-states the requirements of §141.16 of 40 CFR part 141. This latter
regulation, which established the MCLs applicable to community water systems, stipulated
that "dose equivalents shall be calculated on the basis of a 2 liter per day drinking water
intake." This consistency between the two regulatory regimes reflects the Agency's desire to
apply the underlying substantive requirements of the Safe Drinking Water Act to its program
that regulates the disposal of spent nuclear fuel, high-level and transuranic radioactive wastes
such as the WIPP.
In furtherance of this goal, the groundwater protection requirements of subpart C directly
incorporate the maximum contaminant levels (MCLs) established under the Safe Drinking
Water Act at 40 CFR part 141. Disposal systems shall therefore be designed to provide a
reasonable expectation that 10,000 years of undisturbed performance after disposal shall not
14-2
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cause the levels of radioactivity in any USDW in the accessible environment to exceed the
limits specified in 40 CFR part 141 as they exist on January, 19, 1994. Current SDWA
MCLs for radionuclides were promulgated on July 9, 1976 (41 FR 28402) and became
effective on June 24, 1977.
14.3 INDIVIDUAL PROTECTION
The requirements of §194.51 apply to the maximally exposed individual located in the
accessible environment. The Agency designated the maximally exposed individual as the
protected individual consistent with the stated objective of the disposal regulations. As noted
hi the promulgation of the disposal regulations in 1993:
The EPA has chosen a 15-millirem CED [committed effective dose] per year
limit because it finds the lifetime risk represented by this level of exposure to
present an acceptable risk for the purposes of this rulemaking since it involves
only a small number of potential sites and would result in only a small number
of people potentially exposed to the maximum allowed individual risk.
Thus, to ensure that only a small number of persons will be potentially exposed to waste at
the WIPP, the Agency required that, hi compliance assessments of undisturbed performance,
the protected individual must be the maximally exposed individual. Additionally, §191.15 of
the disposal regulations specifies that "the disposal system shall not cause the committed
effective dose, received through all potential pathways to the disposal system, to any member
of the accessible environment to exceed 15 millirems." In the final rule for the WIPP,
§194.52, therefore requires that the dose to individuals be calculated via all potential
exposure pathways. In developing criteria for individual protection, the Agency reviewed the
technical bases for release, transport, exposure, and dose and risk analyses supporting the
promulgation of the disposal regulations. These analyses strongly suggested that the release
and transport of radionuclides from a disposal system would most likely occur via ground
water, and that the maximum radiation dose delivered to any individual beyond the site
boundary would be the sum of the doses delivered through the water-dependent exposure
pathways, e.g., consumption of contaminated drinking water, ingestion of products (i.e.,
meat and milk) from animals fed contaminated water, ingestion of crops irrigated with
contaminated water, and direct radiation exposure due to radionuclides deposited on ground
surfaces due to irrigation, among others. EPA also recognized that different dose and risks
estimates were possible depending on the land use exposure scenario selected.
14-3
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14.3.1 Consideration of Exposure Pathways (§194.52)
Given a plausible release scenario involving migration of contaminants from a disposal
system via ground water to the accessible environment, several water-dependent exposure
pathways are theoretically possible, including:
• ingestion of contaminated drinking water;
• ingestion of contaminated home-grown produce (fruits and vegetables) irrigated
with contaminated well water;
• ingestion of meat (beef) from livestock fed contaminated well water or
contaminated crops irrigated with contaminated well water;
• ingestion of contaminated milk from livestock fed contaminated well water or
contaminated crops irrigated with contaminated well water;
• ingestion of contaminated soil irrigated with contaminated well water;
• dermal contact with contaminated soil irrigated with contaminated well water;
• inhalation of airborne suspended or resuspended contaminated soil irrigated
with contaminated well water; and
• direct radiation exposure to photon-emitting radionuclides in soil irrigated with
contaminated well water.
The maximum radiation dose delivered to the protected individual depends on several factors,
including (but not limited to):
• all radionuclides and their range of concentrations in ground water;
• surface media that may become contaminated as a result of the potential
ground water uses (e.g., drinking water, irrigation, dust suppression, etc.);
• exposure scenarios and pathways based on potential land uses (e.g.,
residential, commercial, industrial, agricultural, etc.); and
• exposure factor assumptions (e.g., intake rates, exposure times, etc.) and dose-
to-risk conversion factors.
14-4
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In order to implement an all-pathways analysis of the radiation dose that a member of the
public receives it is necessary to decide: (1) where the person is to be located; (2) the human
intake, radiation risks, and dose calculations to be used; and (3) the environmental pathways
to be considered, the appropriate scenarios, and pathway parameter values to be used.
14.3.2 Location of Protected Individual
§191.15 limits the annual dose from the waste in the disposal system to any member of the
public in the accessible environment. The definition of the accessible environment includes
the ground surface on the WIPP Site. However, the preamble to the Final 40 CFR part 191
Rule says that "Groundwater withdrawn for consumption directly from within the controlled
area need not be included hi the analyses because geologic media within the controlled area
are an integral part of the disposal system's capability to provide long-term isolation" (58 FR
66403).
14.3.3 Calculation of Radiation Dose
Appendix B of 40 CFR part 191 describes how the Annual Committed Effective Dose is
calculated once individual organ doses hi rads are obtained. Consistent with the future states
assumptions, the radiation weighting factors and the tissue weighting factors are assumed to
remain unchanged in the future. The dose calculations also require concentrations of
individual radionuclides present in the air, water, and foods taken into the body and the
applicable rates of air, water, and food intakes. Then, dose conversion factors are necessary
to relate the intake of a radionuclide to the Committed Effective Dose.
14.3.4 EPA's Standardized Exposure Scenarios and Default Exposure Parameter
Values for Human Health Risk Assignment
One example of the treatment of exposure scenarios is that used by EPA's Superfund
program and Office of Radiation and Indoor Air (ORIA) to assess human health risks to
individuals due to exposure to hazardous chemical substances and radionuclides from cleanup
sites. These exposure scenarios provide an example of an approach that uses all-pathways of
exposure to a maximally exposed individual. The specific application of the final rule, 40
CFR part 194, will depend on the specific considerations of the WIPP site and the
surrounding region. These EPA guidelines assume reasonable maximum exposure (RME)
14-5
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conditions under different post-cleanup cases including the following four land-use
classifications: 1) residential, 2) commercial/industrial, 3) agricultural, and 4) recreational.
This section defines, for each scenario, the principal exposure pathways, key exposure
parameters, and standardized default parameter values. Several EPA documents may be
consulted for additional information (see references: EPA89a, EPA89b, EPA91a, EPA92,
and EPA94). For additional illustration, Table 14-1 compares EPA, DOE, and NRC intake
rates and exposure assumptions that could be used as default values in the scenarios.
EPA's Superfund program currently defines exposure scenarios within the context of the four
land-use classifications listed above (EPA89a and EPA91a). EPA defines RME as "the
maximum exposure that [any individual] is reasonably expected to [receive] at a site"
(EPA89a) or as the "high-end individual exposure" (EPA91a). In both cases, EPA describes
the RME concept as an approach which uses standardized exposure pathways and default
exposure factor values to calculate maximum reasonable estimates of contaminant intake and
risk for individuals in an exposed population.
The RME approach provides estimates of individual intake and risk that are protective and
reasonable, but not the worst possible case. EPA developed the RME concept and
standardized exposure scenarios and assumptions to: (1) reduce unwarranted variability in
assumptions used in baseline risk assessments to characterize potentially exposed populations,
and (2) achieve consistency in evaluating site risks and setting cleanup goals at CERCLA
sites.
The Agency recognizes that exposure conditions at specific sites can and often do differ from
the generic case described above. For this reason, in the Superfund program EPA has
encouraged the use of site-specific scenarios and exposure factors to estimate intakes and
risks at Superfund sites, provided these assumptions can be justified and documented
(EPA89a).
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Table 14-1. Comparison of EPA, DOE, and NRC Intake Rates and Exposure Assumptions
Factor
Category
Intake Rates
Parameter (Units)
Drinking Water
Ingestion Rate
(1/d)
Inhalation Rate
(nWd)
Soil Ingestion
Rate (mg/d)
Leafy Vegetables
(g/d)
Non-Leafy
Vegetables (g/d)
Fruits (g/d)
Residential
Commercial/
Industrial
Residential
Commercial/
Industrial
Child Resident
Adult Resident
Commercial/
Industrial
Total
Contaminated
Fraction
Actual Intake
(total x fract.)
Total
Contaminated
Fraction
Actual Intake
(total x fract.)
Total
Contaminated
Fraction
Actual Intake
(total x fract.)
Agency Default Values
EPA
2
1
20
20
200
100
50
NS
NS
NS
200
0.4
80
140
0.3
42
DOB
1.4
NS*
23
NS
NS
100
NS
38
0.5
19
236
0.4
118
96
0.5
48
NRC
2
NS
29
NS
NS
50
NS
30
0.25
7.5
140
0.25
35
126
0.25
32
Distribution of Values Reported in the List References
Mean±SD
1.4±0.4
14±4
105 ±82
71 ±77
40 ±15
0.25
10
200 ±83
0.25
50
140±58
0.2
28
90th%
1.9
75
314
268
95th%
95
422
327
RME
2
30
175
0.4
70
770
0.4
308
313
0.3
94
Range
0.3-3
0-800
0-800
0.5-480
3-200
26-510
30-487
Comments
For worker exposures, EPA
assumes half the residential daily
intake consumed during an 8-hour
work day.
DOE uses ICRP reference man
data: 16 hours resting. NRC also
uses ICRP data but assumes 24
hours of light activity.
EPA assumes exposure durations of
6 years for children and 24 years
adults. The weighted intake rate is
120mg/day for 30 years.
EPA does not distinguish "leafy
vegetables" from all vegetables.
DOE values assume that non-leafy
vegetables are 54% of total intake
of fruits, vegetables, and grains.
DOE values assume that fruits are
22% of total intake of fruits,
vegetables, and grains.
14-7
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Table 14-1. Comparison of EPA, DOE, and NRC Intake Rates and Exposure Assumptions (Continued)
Factor
Category
Intake Rates
Parameter (Units)
Grains (g/d)
Milk (Id)
Beef and Poultry
(g/d)
Fish (g/d)
Other Seafood
(g/d)
Total
Contaminated
Fraction
Contaminated
Intake
Total
Contaminated
Fraction
Contaminated
Intake
Total
Contaminated
Fraction
Actual Intake
(total x fract.)
Total
Contaminated
Fraction
Actual Intake
(total x fract.)
Total
Contaminated
Fraction
Actual Intake
(total x fract.)
Agency Default Values
EPA
NS
0
NS
0.4
0.4
0.16
170
0.44
75
54
1
54
NS
NS
NS
DOE
105
0.5
53
0.25
1
0.25
173
1
173
15
0.5
7.5
2.5
0.5
1.2
NRC
189
0.25
47
0.27
NS
NS
214
NS
NS
27
NS
NS
NS
NS
NS
Distribution of Values Reported in the List References
MeaniSD
125
0.4±0.01
0.4
0.16
100±2
0.44
44
12±12
2.1±2.0
90th %
0.75
95th%
42
RME
0.85
0.3
300
75
58
54
14
Range
0.25-1.0
67-124
0-140
Comments
EPA does not use grains in risk
calculations (0% of homegrown
grains are consumed). DOE values
are calculated assuming grains are
24% of total intake of fruits,
vegetables and grains.
DOE assumes 100% of the milk is
contaminated for areas greater than
20,000m2, and applies a correction
factor for smaller areas. NRC lists
0.3 I/day as an average daily
intake.
DOE assumes 100% of the meat is
contaminated for areas greater than
20,000m2, and applies a correction
factor for smaller areas. EPA
numbers are for beef only: data
was available for poultry and eggs.
NRC lists 260 g/day as an average
daily intake.
Mean and 95th % values listed are
for all consumers. A median
90th% for fishermen are 30 g/day
and 140g/day. NRC assumes 19
g/day as an average daily intake.
EPA and NRC do not specify
separate values for "other seafood."
NRC assumes 2.7g/day as an
average daily intake.
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Table 14-1. Comparison of EPA, DOE, and NRC Intake Rates and Exposure Assumptions (Continued)
1 Factor
Category
Exposure
Assumptions
Other
Factors
Parameter (Unite)
Exposure Time
(h/d)
Exposure
Frequency (d/y)
Exposure Duration
(y)
Indoors-
Residential
Indoors-
Commercial/
Industrial
Outdoors-
Residential
Outdoors-
Commercial/
Industrial
Residential
Commercial/
Industrial
Residential
Commercial/
Industrial
Gamma Shielding Factor
Soil Concentration in Air (/*g/m3)
Ratio of Indoor Dust to Outdoor Dust
Dilution Factor for Drinking Water
Livestock Soil Intake Rate (kg/d)
Fodder Intake Rate for Beef (kg/d)
Fodder Intake Rate for Milk (kg/d)
Agency Default Values
EPA
NS
NS
NS
NS
350
250
30
25
0.8
0.2
NS
1-100
NS
NS
NS
DOB
12.5
NS
6.3
NS
350
NS
30
NS
0.7
200
0.4
C*
0.5
68
55
NRC
13
NS
4.7
NS
365
NS
NS
NS
0.33
100
0.5
C
0.6
44
67
Distribution of Values Reported in the List References
Meaa±SD
14.2±2.5
7.5±6.9
0.72±0.89
0.5±0.6
9±9
113±95
0.6±0.7
90th %
24
11.2
2.4
1.8
30
25
?5(h%
RME
200
Range
2-24
0-16
0-24
0-7.7
0-365
0-365
0->33
0-1
9-1800
0-1
1-100
0.2-2.9
Comments .,.,;:;
Calculated from EPA89 assuming 8
hours as an average work day.
Calculated from EPA89 assuming 8
hours as an average work day.
EPA assumes 2 weeks vacation per
year away from home.
EPA assumes 5 d/wk, 50wk/yr.
EPA assumes dilution factors of 1 ,
10, and 100.
DOE based on IAEA92, NRC
based IAEA82.
DOE based on NCRP91, NRC
based on IAEA82.
14-9
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Table 14-1. Comparison of EPA, DOE, and NRC Intake Rates and Exposure Assumptions (Continued)
Factor
Category
Other
Factors
Parameter (Units)
Volatilization
Factor for Rn-222
(pCi/m3 per
pCi/g)
Volatilization
Factor for Rn-220
(pCi/m3 per
pCi/g)
Indoors
Outdoors
Indoors
Outdoors
Agency Default Values
;fEPA-.::
NS
120
NS
5
:^QB
c*
c
C
C
NRC
NS
NS
NS
NS
Distribution of Values Reported in the List References
Mean±SD
1250±3110
120±110
100±96
90lh%
3400
95th% ::
5000
RME
4000
Range
400
30,000
20-500
25-500
Comments
Data from EPA's National
Residential Radon Survey assuming
IpCi/g Ra-226.
Assumes IpCi/g Ra-224 in soil.
* NS = Not Specified by Agency; NC = Not considered in soil model calculations; C = Calculated by RESRAD.
EPA EPA89 Exposure Factors Handbook. Office of Health and Environmental Assessment, EPA 600/8-89 043, 1989.
References: EPA91 Risk Assessment Guidance for Superfund. Volume 1: Hurnan Health Evaluation Manual. Supplemental Guidance. "Standard
Default Exposure Factors", OSWER Directive 9285.6-03, 1991.
DOE
References: DOE92 Data Collection Handbook for Establishing Residual Radioactive Material Guidelines with RESRAD. 1992.
NRC NRC92 Residual Radioactive Contamination from Decommissioning. NEUREG/CR-5512, PNL-7994, 1992.
References: NRC77 Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance
with 10 CFR 50. Appendix 2. Reg Guide 1.109, 1977.
Other IAEA82 Generic Models and Parameters for Assessing the Environmental Transfer of Radionuclides from Routine
References: Releases: Exposure of Critical Groups. Safety Series No. 57, 1982.
IAEA92 Handbook of Parameter Values for the Prediction of Radionuclide Transfer in Temperate Environments. 9th Draft, 1992.
NCRP91 Screening Models for Releases of Radionuclides to Air. Surface Water, and Ground Water, draft document, 1991
14-10
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Residential Exposure Scenario. The Superfund guidelines employ residential exposure
scenarios whenever there are homes on or near a contaminated site, or whenever future
residential development is a reasonable expectation based on consideration of local zoning
laws, land-use trends, and site suitability. Five exposure pathways are evaluated routinely
under these scenarios to assess risks from radionuclides in soil (EPA91a): 1) direct external
radiation from photon-emitting radionuclides in the soil, 2) inhalation of resuspended
contaminated dust, 3) inhalation of radon and radon decay products (only when radium is
present in soil), 4) ingestion of contaminated drinking water, and 5) ingestion of
contaminated soil. Two additional pathways-consumption of contaminated home-grown
produce and fish are also considered at some residential sites, but only when site-specific
circumstances warrant inclusion.
Commercial/Industrial Exposure Scenario. The Superfund guidelines utilize occupational
exposure scenarios whenever the land use is, or is expected to be, commercial or industrial.
These scenarios typically assess adult worker exposures that assume exposure occurs at the
workplace during an 8-hour work day, five days per week, 50 weeks per year, for 25 years.
Exposure pathways considered under these scenarios are identical to those evaluated for
residential exposures, with the omission of pathways for consumption of home-grown
produce and fish. Values for exposure factors and intake rates assumed for
commercial/industrial exposures are generally less than those assumed for residential
exposures.
Agricultural Exposure Scenario. The Superfund guidelines utilize agricultural exposure
scenarios whenever individuals live or work in contaminated areas zoned for farming
activities, such as growing crops or raising livestock. Under these scenarios, EPA assumes
farm family members are exposed through the same five principal pathways evaluated for
individuals under the residential setting, plus the mandatory inclusion of the plant pathway
(i.e., consumption of home-grown produce). EPA also considers additional pathways for the
ingestion of contaminated beef and dairy products, but only when such pathways are valid for
the site conditions and lifestyles of the onsite populations.
Additional soil exposure pathways considered under the agricultural exposure scenario
include: 1) ingestion of home-grown produce (fruits and vegetables) contaminated with
radionuclides taken up from soil, 2) ingestion of meat (beef) containing radionuclides taken
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up by cows grazing on contaminated plants (fodder), and 3) ingestion of milk containing
radionuclides taken up by cows grazing on contaminated plants (fodder).
Recreational Exposure Scenario. Under the recreational exposure scenario, the Superfund
guidelines include pathways for consumption of locally caught fish - both for subsistence and
recreation - and for dermal exposures that might occur during swimming and wading. Fish
pathways are evaluated only when there is access to a contaminated water body large enough
to produce a consistent supply of edible-sized fish over the anticipated exposure period.
Pathways for assessing exposures during swimming and wading are currently being re-
evaluated by EPA, along with other potential recreational exposure pathways, such as
hunting and dirt-biking.
14.3.5 Exposure Scenarios Considered by DOE and the NRC
In general, the Department of Energy and the Nuclear Regulatory Commission consider
similar land-use scenarios hi the remediation of actual sites (DOE93 and NRC92). However,
in some cases, DOE or NRC may evaluate additional exposure scenarios and pathways that
are not based on any specific land-use consideration - such as the intruder exposure scenario
- or may apply different default values for exposure factors and intake rates than those
currently recommended by EPA. Table 14-1 compares EPA, DOE, and NRC default
exposure factor values. It should be noted that all three agencies strongly recommend the
use of site-specific data for modeling doses and risks, but only when the data are available
and meet appropriate data quality objectives and data usability requirements.
14.4 SCOPE OF COMPLIANCE ASSESSMENTS (§194.54)
In accordance with §191.15(a) and §191.24(b), calculations of compliance with the individual
and ground-water protection requirements must consider the undisturbed performance of the
disposal system. "Undisturbed performance" is defined at §191.12(p) as "the predicted
behavior of a disposal system, including consideration of the uncertainties hi predicted
behavior, if the disposal system is not disrupted by human-intrusion or the occurrence of
unlikely natural events."
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To clarify the Agency's intent for this requirement, §194.54 specifies that any application for
certification of compliance shall include information which:
(1) identifies the potential processes, events, or sequences of processes and events
that may occur over the regulatory tune frame;
(2) identifies the processes, events, or sequences of processes and events that may
be included in compliance assessment results provided in any compliance
application; and
(3) documents why any processes, events or sequences of processes and events
identified under paragraph (a)(l) of this section were not included in
compliance assessment results provided in any compliance application.
Unlike the containment requirements, the individual and groundwater protection requirements
do not apply to cumulative releases nor do they contain probabilistic requirements, such as
the requirement that certain releases be less than 1 in 1,000 likely to be exceeded (191.13).
Instead, the individual and groundwater requirements apply to the doses received during one
individual's lifetime, versus 10,000 years for the containment requirements. Further, the
expected value of the dose received — the mean value — must be less than the applicable dose
limit, for example, 15 mrem in the case of the individual protection requirements. There is
no regulatory significance to the probability with which the dose limit will be exceeded, and
hence these requirements cannot be treated analogously to the probabilistic containment
requirements. Therefore, providing a numerical cut-off for probability, such as the 1 in
10,000 threshold test applicable to performance assessments, would not be applicable.
However, some screening of processes and events was contemplated in 40 CFR part 191,
which in the definition of "undisturbed performance" in 40 CFR part 191 state that
compliance assessments may exclude from consideration any unlikely natural processes and
events.
Several differences emerge upon examination of the performance assessments needed for the
containment requirements and the compliance assessments needed for the individual and
groundwater protection requirements. For example, the individual protection requirements
apply only to the accumulation of dose over an individual's lifetime versus 10,000 years in
the containment requirements. Second, as just explicated, the individual and groundwater
protection requirements are not probabilistic, unlike the containment requirements. Third,
whereas the focus of the individual and groundwater protection requirements is on the
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contribution of natural processes and events to doses to individuals, the containment
requirements focus on the contribution such processes and events make toward releases of
radionuclides to the accessible environment. In view of these considerations, the Agency
recognized that the significantly different form of the containment requirements versus the
individual and groundwater protection requirements necessitated a different treatment of the
screening of processes and events.
In compliance assessments, therefore, the Agency requires that a qualitative judgment be
made regarding the likelihood with which groundwater and individual exposure pathways will
be affected, over the time scale of an individual's lifetime (not 10,000 years as in the
containment requirements) by the occurrence of different natural events. Although the
universe of processes and events considered in the performance assessments (for .the
containment requirements) will closely resemble that of compliance assessments, the different
regulatory requirements attending each analysis, as noted above, might allow for subtle
differences regarding whether the individual events should be included in the analysis. As
with performance assessments, the final rule at §194.54(a) requires compliance applications
to document why any processes and events or sequences of processes and events that may
occur over the regulatory time frame were not included in compliance applications.
14.5 RESULTS OF COMPLIANCE ASSESSMENTS (§194.55)
As discussed above, the part 191 disposal standards require that compliance assessments
include consideration of the uncertainties associated with the undisturbed performance of the
disposal system. To accomplish this assessment, it is necessary to identify all disposal
system parameters that can affect the performance of the WIPP, as well as to identify the
uncertainty associated with each parameter. This approach is identical to the one used to
certify and demonstrate compliance with the containment requirements of 40 CFR part 191.
As part of this approach, EPA requires a three-step process, whereby:
1. all uncertain disposal system parameters are identified;
2. probability distribution functions are developed for these parameters (a
probability distribution function assigns a probability of occurrence to each
value for a given parameter); and
14-14
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3. following steps 1 and 2, statistical sampling techniques are used to draw
random samples from across the full range of probability distributions for
parameter values used in compliance assessments.
The Agency believes that this process will help ensure that all possible values of a parameter
have been considered in compiling compliance assessment results.
Two types of statistical sampling techniques are used frequently, namely the Monte Carlo
and Latin Hypercube techniques. The Monte Carlo technique uses a random sampling
scheme, which, as the name implies, involves the selection of values for a particular
parameter at random within the predefined probability function for the parameter. The major
disadvantage of this technique is that a large number of iterations is necessary to ensure that
the selected values are sampled adequately. In comparison, the Latin Hypercube technique
uses a special case of stratified sampling that involves the systematic partitioning of the range
of values for a particular parameter into some number of strata. The principal advantage of
this technique is that it requires less sampling iterations to ensure that the entire range of
values is represented, because it draws samples from each stratum.
Also under §194.55, EPA requires that the range of estimated radiation doses to individuals
(as generated through use of the computational techniques referred to above), and the range
of estimated radionuclide concentrations in ground water must be large enough such that the
maximum estimate generated exceeds the 99th percentile of the population of estimates with
at least a 95% probability. The "population of estimates" refers to the set of all possible
estimates that can be generated from all disposal system parameter values used in
compliance assessments. A single estimate, in effect, samples this population.
The Agency is including this provision for the purpose of ensuring that there is a 95 %
probability that 99% of all possible values have been exceeded by the maximum estimate
generated. This is similar to the requirement for the number of CCDFs (complementary
cumulative distribution functions) which must be generated for purposes of compliance with
the containment requirements.
In order to assure that all pertinent information is provided to the Agency, EPA is also
requiring that compliance applications display the full range of estimated, radiation doses and
the full range of estimated radionuclide concentrations. The Agency believes that this
requirement will help ensure that a full range of values is considered in compliance
assessments.
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Finally, the Agency requires that any compliance certification application provide information
which demonstrates that there is at least a 95 % level of statistical confidence that the mean
and the median of the full range of estimated radiation doses and of the full range of
estimated radionuclide concentrations meet the requirements set forth hi sections 15 and
subpart C of 40 CFR part 191. The mean estimate provides a measure of compliance that
expresses the average impacts of the disposal system on individuals and ground water. The
median estimate provides a measure of compliance that expresses the central tendency of a
population of estimates. Specifically, the median represents the point that a calculated
estimate would be equally likely to fall above or below. Insofar as both statistics contain
useful information, the Agency's approach assures that both meet the limits of the individual
and ground-water protection requirements.
It is important to note that a reasonable expectation of compliance with the individual and
ground-water protection requirements will not be based solely on a final statistical estimate of
doses to individuals or radionuclide concentrations in ground water. Whether a reasonable
expectation of compliance will be achieved or not will be evaluated on the basis of the full
record before the Agency and a thorough consideration of the methods and assumptions that
produced compliance assessment results. For instance, in certifying and determining
compliance, the Agency will consider such factors as the reasonableness of the processes and
events considered, the appropriateness of any expert judgment elicitation used to provide
inputs to the assessments, the adequacy of peer review, the quality of the models, and the
quality of data inputs to those models.
14.6 REFERENCES
DOE93 Department of Energy, "Manual for Implementing Residual Radioactive
Material Guidelines Using RESRAD, Version 5.0," Final Draft, 1993.
EPA89a Environmental Protection Agency, "Risk Assessment Guidance for Superfund,
Volume 1, Human Health Evaluation Manual, Part A," EPA/540/1-89-002,
1989.
EPA89b Environmental Protection Agency, "Exposure Factors Handbook," EPA/600/8-
89/043, 1989.
EPA91a Environmental Protection Agency, "Risk Assessment Guidance for Superfund,
Volume 1, Human Health Evaluation Manual. Part B, Development of Risk
Based Preliminary Remediation Goals," PB92-963333, 1991.
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EPA92 Environmental Protection Agency, "Guidance on Risk Assessment for Risk
Managers and Risk Assessors," Memo from F. Henry Habicht II to Assistant
and Regional Administrators, February 26, 1992.
EPA94 Environmental Protection Agency, "Draft Guidance for Soil Screening Level
Framework, Quick Reference Fact Sheet, Review Draft," 1994.
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