APPENDIX C Savannah River Site Spent Nuclear Fuel Management Program
Department of Energy Programmatic
Spent Nuclear Fuel Management
and
Idaho National Engineering Laboratory
Environmental Restoration and
Waste Management Programs
Final Environmental Impact Statement
Volume 1
Appendix C
Savannah River Site
Spent Nuclear Fuel Management Program
April 1995
U.S. Department of Energy
Office of Environmental Management
Idaho Operations Office
CONTENTS
1. INTRODUCTION 1-1
2. BACKGROUND 2-1
2.1 SRS Overview 2-1
2.1.1 Site Description 2-1
2.1.2 Site History 2-6
2.1.3 Mission 2-6
2.1.4 Management 2-7
2.2 Regulatory Framework 2-7
2.2.1 Federal 2-7
2.2.2 State 2-8
2.2.3 Local 2-9
2.3 Spent Nuclear Fuel Management Program at the Savannah River Site 2-9
2.4 Vulnerabilities Associated with SRS Spent Nuclear Fuel 2-11
2.5 Representative Host Sites 2-14
2.5.1 F- and H-Areas 2-16
2.5.2 Undeveloped Representative Host Site 2-16
3. SPENT NUCLEAR FUEL ALTERNATIVES 3-1
3.1 SRS Management Approach 3-2
3.1.1 Management Options 3-2
3.1.1.1 Wet Storage 3-3
3.1.1.2 Dry Storage 3-3
3.1.1.3 Processing and Dry Storage 3-3
3.1.2 Management Plan 3-4
3.1.2.1 Aluminum-clad Fuels 3-4
3.1.2.2 Nonaluminum-clad Fuels 3-4
3.2 Description of Alternatives 3-6
3.2.1 Overview 3-6
3.2.2 Alternative 1 - No Action 3-10
3.2.2.1 Overview 3-10
3.2.2.2 SRS Alternative 1 - Wet Storage 3-10
3.2.3 Alternative 2 - Decentralization 3-11
3.2.3.1 Overview 3-11
3.2.3.2 SRS Options 2a, 2b, and 2c 3-12
3.2.3.2.1 Option 2a - Dry Storage 3-12
3.2.3.2.2 Option 2b - Wet Storage 3-12
3.2.3.2.3 Option 2c - Processing and Storage 3-13
3.2.4 Alternative 3 - 1992/1993 Planning Basis 3-13
3.2.4.1 Overview 3-13
3.2.4.2 SRS Options 3a, 3b, and 3c 3-13
3.2.4.2.1 Option 3a - Dry Storage 3-14
3.2.4.2.2 Option 3b - Wet Storage 3-14
3.2.4.2.3 Option 3c - Processing and Storage 3-14
3.2.5 Alternative 4 - Regionalization 3-14
3.2.5.1 Overview 3-14
3.2.5.2 SRS Options 4a, 4b, and 4c (Regionalization A) 3-15
3.2.5.2.1 Option 4a - Dry Storage 3-16
3.2.5.2.2 Option 4b - Wet Storage 3-16
3.2.5.2.3 Option 4c - Processing and Storage 3-16
3.2.5.3 SRS Options 4d, 4e, 4f, and 4g (Regionalization B) 3-16
3.2.5.3.1 Option 4d - Dry Storage 3-17
3.2.5.3.2 Option 4e - Wet Storage 3-17
3.2.5.3.3 Option 4f - Processing and Storage 3-17
3.2.5.3.4 Option 4g - Shipment Off the Site 3-18
3.2.6 Alternative 5 - Centralization 3-18
3.2.6.1 Overview 3-18
3.2.6.2 SRS Options 5a, 5b, 5c, and 5d 3-18
3.2.6.2.1 Option 5a - Dry Storage 3-19
3.2.6.2.2 Option 5b - Wet Storage 3-19
3.2.6.2.3 Option 5c - Processing and Storage 3-19
3.2.6.2.4 Option 5d - Shipment Off the Site 3-20
3.3 Comparison of Alternatives 3-20
4. AFFECTED ENVIRONMENT 4-1
4.1 Overview 4-1
4.2 Land Use 4-1
4.3 Socioeconomics 4-5
4.3.1 Employment and Labor Force 4-5
4.3.2 Personal Income 4-6
4.3.3 Population 4-6
4.3.4 Housing 4-7
4.3.5 Community Infrastructure and Services 4-7
4.3.6 Government Fiscal Structure 4-8
4.4 Cultural Resources 4-9
4.4.1 Archeological Sites and Historic Structures 4-9
4.4.2 Native American Cultural Resources 4-10
4.4.3 Paleontological Resources 4-11
4.5 Aesthetics and Scenic Resources 4-11
4.6 Geology 4-11
4.6.1 General Geology 4-12
4.6.2 Geologic Resources 4-16
4.6.3 Seismic and Volcanic Hazards 4-16
4.7 Air Resources 4-23
4.7.1 Meteorology and Climatology 4-23
4.7.1.1 Occurrence of Violent Weather 4-26
4.7.1.2 Atmospheric Stability 4-27
4.7.2 Nonradiological Air Quality 4-27
4.7.2.1 Background Air Quality 4-27
4.7.2.2 Air Pollutant Source Emissions 4-27
4.7.2.3 Ambient Air Monitoring 4-28
4.7.2.4 Atmospheric Dispersion Modeling 4-28
4.7.2.5 Summary of Nonradiological Air Quality 4-28
4.7.3 Radiological Air Quality 4-28
4.7.3.1 Background and Baseline Radiological Conditions 4-28
4.7.3.2 Sources of Radiological Emissions 4-31
4.8 Water Resources 4-32
4.8.1 Surface Water 4-32
4.8.1.1 SRS Streams 4-35
4.8.1.2 Surface Water Quality 4-36
4.8.2 Groundwater Resources 4-36
4.8.2.1 Hydrostratigraphic Units 4-36
4.8.2.2 Groundwater Flow 4-40
4.8.2.3 Groundwater Quality 4-42
4.8.2.4 Groundwater Use 4-46
4.9 Ecological Resources 4-46
4.9.1 Terrestrial Ecology 4-48
4.9.2 Wetlands 4-49
4.9.3 Aquatic Ecology 4-50
4.9.4 Threatened and Endangered Species 4-50
4.10 Noise 4-52
4.11 Traffic and Transportation 4-53
4.11.1 Regional Infrastructure 4-53
4.11.1.1 Regional Roads 4-54
4.11.1.2 Regional Railroads 4-54
4.11.2 SRS Infrastructure 4-54
4.11.2.1 SRS Roads 4-54
4.11.2.2 SRS Railroads 4-58
4.12 Occupational and Public Radiological Health and Safety 4-60
4.12.1 Occupational Health and Safety 4-60
4.12.2 Public Health and Safety 4-62
4.13 Utilities and Energy 4-65
4.13.1 Electricity 4-65
4.13.2 Water Consumption 4-66
4.13.3 Fuel Consumption 4-66
4.13.4 Wastewater Treatment 4-66
CONTENTS (continued)
4.14 Materials and Waste Management 4-67
4.14.1 High-Level Waste 4-70
4.14.2 Transuranic Waste 4-73
4.14.3 Mixed Low-Level Waste 4-74
4.14.4 Low-Level Waste 4-74
4.14.5 Hazardous Waste 4-75
4.14.6 Sanitary Waste 4-75
4.14.7 Hazardous Materials 4-75
5.0 ENVIRONMENTAL CONSEQUENCES 5-1
5.1 Overview 5-1
5.2 Land Use 5-1
5.3 Socioeconomics 5-2
5.3.1 Potential Impacts 5-2
5.4 Cultural Resources 5-5
5.5 Aesthetic and Scenic Resources 5-6
5.6 Geologic Resources 5-6
5.7 Air Quality Consequences 5-7
5.7.1 Alternative 1 - No Action 5-20
5.7.2 Alternative 2 - Decentralization 5-21
5.7.3 Alternative 3 - 1992/1993 Planning Basis 5-21
5.7.4 Alternative 4 - Regionalization 5-21
5.7.5 Alternative 5 - Centralization 5-21
5.8 Water Quality and Related Consequences 5-22
5.8.1 Alternative 1 - No Action 5-27
5.8.1.1 Option 1 - Wet Storage 5-27
5.8.2 Alternative 2 - Decentralization 5-27
5.8.3 Alternative 3 - 1992/1993 Planning Basis 5-28
5.8.4 Alternative 4 - Regionalization 5-28
5.8.5 Alternative 5 - Centralization 5-28
5.9 Ecology 5-29
5.9.1 Alternative 1 - No Action 5-29
5.9.2 Alternative 2 - Decentralization 5-29
5.9.2.1 Option 2a - Dry Storage 5-29
5.9.2.2 Option 2b - Wet Storage 5-30
5.9.2.3 Option 2c - Processing and Storage 5-30
5.9.3 Alternative 3 - 1992/1993 Planning Basis 5-30
5.9.4 Alternative 4 - Regionalization 5-30
5.9.5 Alternative 5 - Centralization 5-30
5.9.5.1 Option 5a - Dry Storage 5-30
5.9.5.2 Option 5b - Wet Storage 5-32
5.9.5.3 Option 5c - Processing and Storage 5-32
5.9.5.4 Option 5d - Shipment off the Site 5-32
5.10 Noise 5-33
5.11 Traffic and Transportation 5-34
5.11.1 Traffic. . . 5-34
5.11.2 Transportation 5-34
5.11.2.1 Onsite Spent Nuclear Fuel Shipments 5-35
5.11.2.2 Incident-Free Transportation Analysis 5-35
5.11.2.3 Transportation Accident Analysis. . . 5-36
5.11.3 Onsite Mitigation and Preventative Measures 5-37
5.12 Occupational and Public Health and Safety 5-38
5.12.1 Radiological Health 5-38
5.12.2 Nonradiological Health 5-40
5.12.3 Industrial Safety 5-44
5.13 Utilities and Energy 5-46
5.14 Materials and Waste Management 5-48
5.14.1 Alternative Comparison 5-50
5.14.2 Impact on the SRS Waste Management Capacity 5-51
5.15 Accident Analysis 5-51
5.15.1 Historic Accidents at the Savannah River Site 5-52
5.15.2 Potential Facility Accidents 5-53
5.15.2.1 Alternative 1 - No Action 5-55
5.15.2.2 Alternative 2 - Decentralization 5-57
5.15.2.2.1 Option 2a - Dry Storage 5-57
5.15.2.2.2 Option 2b - Wet Storage 5-61
5.15.2.2.3 Option 2c - Processing and Storage 5-62
5.15.2.3 Alternative 3 - 1992/1993 Planning Basis 5-63
5.15.2.3.1 Option 3a - Dry Storage 5-63
5.15.2.3.2 Option 3b - Wet Storage 5-64
5.15.2.3.3 Option 3c - Processing and Storage 5-64
5.15.2.4 Alternative 4 - Regionalization 5-64
5.15.2.4.1 Option 4a - Dry Storage 5-65
5.15.2.4.2 Option 4b - Wet Storage 5-65
5.15.2.4.3 Option 4c - Processing and Storage 5-65
5.15.2.4.4 Option 4d - Dry Storage 5-65
5.15.2.4.5 Option 4e - Wet Storage 5-65
5.15.2.4.6 Option 4f - Processing and Storage 5-65
5.15.2.4.7 Option 4g - Shipping Off Site. . . 5-66
5.15.2.5 Alternative 5 - Centralization 5-66
5.15.2.5.1 Option 5a - Dry Storage 5-66
5.15.2.5.2 Option 5b - Wet Storage 5-67
5.15.2.5.3 Option 5c - Processing and Storage 5-67
5.15.2.5.4 Option 5d - Shipping Off Site 5-67
5.15.3 Chemical Hazard Evaluation 5-67
5.15.3.1 Receiving Basin for Offsite Fuel 5-67
5.15.3.2 Reactor Basins 5-69
5.15.3.3 H-Area 5-69
5.15.3.4 F-Area 5-69
5.15.4 Secondary Impacts 5-70
5.15.4.1 Biotic Resources 5-70
5.15.4.2 Water Resources 5-70
5.15.4.3 Economic Impacts 5-72
5.15.4.4 National Defense 5-72
5.15.4.5 Environmental Contamination 5-72
5.15.4.6 Endangered Species 5-72
5.15.4.7 Land Use 5-72
5.15.4.8 Treaty Rights 5-72
5.15.5 Adjusted Point Estimate of Risk Summary 5-73
5.16 Cumulative Impacts . . . . . . 5-73
5.16.1 Land Use . . . . . . . 5-86
5.16.2 Socioeconomics. . . . 5-91
5.16.3 Air Quality. . . . . . . . 5-91
5.16.4 Water Resources . . . 5-97
5.16.5 Occupational and Public Health and Safety 5-98
5.16.6 Waste Management. . . 5-99
5.17 Unavoidable Adverse Environmental Impacts 5-99
5.18 Relationship Between Short-Term Use of the Environment and the
Maintenance and Enhancement of Long-Term Productivity 5-100
5.19 Irreversible and Irretrievable Commitments of Resources 5-101
5.20 Potential Mitigation Measures 5-102
5.20.1 Pollution Prevention 5-102
5.20.2 Socioeconomics 5-102
5.20.3 Cultural Resources 5-103
5.20.4 Geology 5-103
5.20.5 Air Resources 5-104
5.20.6 Water Resources 5-104
5.20.7 Ecological Resources 5-104
5.20.8 Noise 5-105
5.20.9 Traffic and Transportation 5-105
5.20.10 Occupational and Public Health and Safety 5-105
5.20.11 Utilities and Support Services 5-105
5.20.12 Accidents 5-105
Attachment A - Accident Analysis A-1
TABLES
2-1. Description of functions and principal facilities at SRS areas 2-4
2-2. SRS Fuel Inventory by Facility 2-10
2-3. SRS vulnerabilities by facility, vulnerability, tracking number,
priority categorization, and Action Plan status 2-13
3-1. Quantities of spent nuclear fuel that would be received, shipped,
and managed at the SRS under the five alternatives 3-2
3-2. Actions required under each of the five alternatives at the SRS 3-7
3-3. Comparison of impacts for the five alternatives 3-21
4-1. Forecast employment and population data for the Savannah River
Site and the region of influence 4-6
4-2. Earthquakes in the SRS region with a Modified Mercalli Intensity
greater than V 4-20
4-3. Earthquakes in the SRS region with a Modified Mercalli Intensity
greater than IV or a magnitude greater than 2.0 4-21
4-4. Estimated ambient concentration contributions of criteria air
pollutants from existing SRS sources and sources planned for
construction or operation through 1995 4-29
4-5. Baseline 24-hour average modeled concentrations at the SRS
boundary - toxic air pollutants regulated by South Carolina from
existing SRS sources and sources planned for construction or
operation through 1995 4-30
4-6. Radioactivity in air at SRS perimeter at 160-kilometer (100-mile)
radius 4-31
4-7. Average atmospheric tritium concentrations on and around the
Savannah River Site 4-31
4-8. Operational groupings and function of radionuclide sources 4-32
4-9. Annual quantity of radionuclide emissions from the Savannah River
Site 4-33
4-10. Water quality in the Savannah River above the confluence with Upper
Three Runs near the Savannah River Site in 1990 4-37
4-11. Water quality in the Savannah River below the confluence with
Lower Three Runs near the Savannah River Site in 1990 4-38
4-12. Representative groundwater quality data for nonradioactive
constituents from the Savannah River Site 4-43
4-13. Representative groundwater data for radioactive constituents from
the Savannah River Site 4-44
4-14. Land cover of undeveloped areas on the Savannah River Site 4-47
4-15. Threatened, endangered, and candidate plant and animal species
of the SRS 4-51
4-16. SRS traffic counts - major roads 4-59
4-17. Radioactivity in air at the Savannah River Site and vicinity 4-61
4-18. Tritium measured in air at the Savannah River Site 4-61
4-19. Maximum radioactivity concentrations in soil at the Savannah River
Site 4-62
4-20. Annual involved worker doses, 1983-1987 4-63
4-21. Annual involved worker doses, 1993 4-63
4-22. Major sources of radiation exposure to the public in the vicinity
of the Savannah River Site 4-63
4-23. Average atmospheric tritium concentrations in the vicinity of the
Savannah River Site 4-64
4-24. Current capacities and usage of utilities and energy at SRS 4-65
4-25. Average annual waste generation forecast for the Savannah River Site 4-73
5-1. Direct construction employment and total population changes by
alternative, 1995-2004 5-3
5-2. Estimated increases in employment and population related to
construction activities for Option 5b, from 1995 to 2004 5-4
5-3. Estimated incremental air quality impacts at the Savannah River
Site boundary from operations of SNF alternatives - criteria
pollutants. 5-8
5-4. Estimated incremental air quality impacts at the Savannah River
Site boundary from operations of SNF alternatives - toxic
pollutants. 5-11
5-5. Incremental air quality pollutant emission rates related to spent
nuclear fuel alternatives - criteria pollutants 5-14
5-6. Incremental air quality pollutant emission rates related to spent
nuclear fuel alternatives - toxic pollutants 5-16
5-7. Estimated maximum annual emissions (in curies) of radionuclides
to the atmosphere from spent nuclear fuel management activities 5-20
5-8. Annual groundwater and surface water usage requirements for each
alternative 5-23
5-9. Estimated maximum liquid radiological releases (in curies) to
the Savannah River from spent nuclear fuel management activities 5-24
5-10. Collective doses and health effects for onsite, incident-free
spent nuclear fuel shipments by alternative 5-36
5-11. Impacts on maximally exposed individual from spent nuclear fuel
transportation accident on the Savannah River Site 5-38
5-12. Impacts on offsite population from spent nuclear fuel transportation
accident on the Savannah River Site 5-38
5-13. Incremental radioactive contaminant annual exposure summary 5-41
5-14. Incremental fatal cancer incidence and maximum probability for
workers 5-42
5-15. Incremental fatal cancer incidence and maximum probability for the
maximally exposed individual and offsite population (air and water
pathways) 5-43
5-16. Nonradiological annual incremental health effects summary 5-45
5-17. Incremental industrial hazard maximum annual incidence summary 5-46
5-18. Estimates of annual electricity, steam, and domestic wastewater
treatment requirements for each alternative 5-47
5-19. Annual average and total volume of radioactive wastes produced
under each alternative during the 40-year interim management period 5-49
5-20. Highest point estimates of risk among receptor groups (option 1) 5-57
5-21. Radioactive release accidents and health effects for spent nuclear
fuel alternatives 5-58
5-22. Highest point estimates of risk among receptor groups (option 2a) 5-61
5-23. Highest point estimates of risk among receptor groups (option 2b) 5-62
5-24. Highest point estimates of risk among receptor groups (option 2c) 5-63
5-25. Results of analyzed chemical accident 5-68
5-26. Qualitative summary of expected secondary impacts 5-71
5-27. Adjusted point estimates of risk for the maximally exposed offsite
individual (radiological accidents) 5-74
5-28. Adjusted point estimates of risk for the colocated worker
(radiological accidents) 5-78
5-29. Adjusted point estimates of risk for the general population -
80 kilometers (radiological accidents) 5-82
5-30. Cumulative impacts associated with construction and operation of
spent fuel alternatives at Savannah River Site 5-87
5-31. Total maximum ground-level concentrations of criteria and toxic
air pollutants at SRS boundary resulting from normal operations and
spent nuclear fuel management alternatives 5-92
5-32. Annual cumulative health effects to workers and offsite population
due to SRS radioactive releases during incident-free operations 5-95
FIGURES
2-1. National location of SRS. 2-2
2-2. Location of principal SRS facilities. 2-5
2-3. Representative host sites on Savannah River Site 2-15
3-1. Diagram of how SRS would manage aluminum-clad and nonaluminum-clad
fuels 3-5
3-2. Types of facilities required for each alternative 3-9
4-1. Generalized land use at the Savannah River Site and vicinity. 4-3
4-2. Federal and state forests and parks within a 2-hour drive from
Savannah River Site 4-4
4-3. Location of the Savannah River Site in the southern United States. 4-13
4-4. Generalized subsurface cross-section across the Savannah River Site 4-14
4-5. Stratigraphy of the SRS region 4-15
4-6. Geologic structures within 150 km of Savannah River Site 4-17
4-7. Geologic faults of the Savannah River Site 4-19
4-8. Seismic hazard curve for SRS 4-24
4-9. Wind rose for the Savannah River Site (1987-1991) 4-25
4-10. Savannah River Site, showing 100-year floodplain, major stream
systems and facilities 4-34
4-11. Comparison of lithostratigraphy and hydrostratigraphy for the SRS
region 4-39
4-12. Groundwater contamination at the Savannah River Site 4-45
4-13. Regional transportation infrastructure 4-55
4-14. Major SRS roads and access points 4-56
4-15. SRS railroad lines 4-57
4-16. Waste management facilities at the Savannah River Site 4-69
4-17. Flow diagram for high-level radioactive waste handling at the
Savannah River Site 4-71
4-18. Flow diagram for waste handling at the Savannah River Site 4-72
5-1. Accident analysis process 5-56
1. INTRODUCTION
The U.S. Department of Energy (DOE) is engaged in two related decisionmaking processes
concerning: (1) the transportation, receipt, processing, and storage of spent nuclear fuel (SNF) at the
DOE Idaho National Engineering Laboratory (INEL) which will focus on the next 10 years; and
(2) programmatic decisions on future spent nuclear fuel management which will emphasize the next 40
years.
DOE is analyzing the environmental consequences of these spent nuclear fuel management
actions in this two-volume Environmental Impact Statement (EIS). Volume 1 supports broad
programmatic decisions that will have applicability across the DOE complex and describes in detail the
purpose and need for this DOE action. Volume 2 is specific to actions at the INEL. This document,
which limits its discussion to the Savannah River Site (SRS) spent nuclear fuel management program,
supports Volume 1 of the EIS. Other documents supporting Volume 1 focus on spent nuclear fuel
management programs for the Hanford Site, INEL, Naval Nuclear Propulsion Program, and other sites.
As part of its planning process for this two-volume EIS, DOE issued an Implementation Plan on
October 29, 1993. The organization of this document is consistent with the provisions established in
the Implementation Plan and are outlined below:
- Chapter 2 contains background information related to the SRS and the framework of
environmental regulations pertinent to spent nuclear fuel management.
- Chapter 3 identifies spent nuclear fuel management alternatives that DOE could implement
at the SRS, and summarizes their potential environmental consequences.
- Chapter 4 describes the existing environmental resources of the SRS that spent nuclear fuel
activities could affect.
- Chapter 5 analyzes in detail the environmental consequences of each spent nuclear fuel
management alternative and describes cumulative impacts. The chapter also contains
information on unavoidable adverse impacts, commitment of resources, short-term use of the
environment and mitigation measures.
2. BACKGROUND
The chapter contains an overview of the Savannah River Site (SRS) and a description of the
regulatory framework related to the actions that this document evaluates. In addition, it discusses the
U.S. Department of Energy (DOE) Spent Nuclear Fuel (SNF) Management Program as it relates to the
SRS. Finally, it describes the representative sites located on the SRS that could serve as locations for
spent nuclear fuel facilities.
2.1 SRS Overview
The SRS is a key DOE facility for research on and processing of special nuclear materials. The
U.S. Government built the Site in the early 1950s to produce the basic materials - primarily
plutonium-239 and tritium - used in the fabrication of nuclear weapons. The DOE Savannah River
Operations Office manages the SRS, and Westinghouse Savannah River Company (WSRC) operates
the Site under contract to DOE.
2.1.1 Site Description
The SRS occupies an area of approximately 310 square miles (800 square kilometers) in western
South Carolina, in a generally rural area about 25 miles (40 kilometers) southeast of Augusta, Georgia,
and 12 miles (19 kilometers) south of Aiken, South Carolina (Figure 2-1). The Savannah River forms
the southwestern border of the SRS, which includes portions of Aiken, Barnwell, and Allendale
Counties. The average population density (1990 census data) in the six-county region of influence
around the Site is 140 people per square mile (54 per square kilometer); the largest concentration is
2,595 people per square mile (1,002 per square kilometer) in the City of Augusta (HNUS 1992). Four
other population centers - Aiken, Allendale, Barnwell, and North Augusta, South Carolina - are
within 22 miles (40 kilometers) of the Site. Three small towns - Jackson, New Ellenton, and
Snelling, South Carolina - are adjacent to the SRS boundary to the northwest, north, and east,
respectively. Based on 1990 U.S. Census Bureau data, the population within a 50-mile (80-kilometer)
radius of the SRS is approximately 620,100 (Arnett et al. 1993).
The Site consists primarily of managed upland forest with some wetland areas. Facilities and
roadways occupy approximately 5 percent of the SRS land area. Access to the Site is controlled, with
Figure 2-1. National location of SRS. public transportation limited to through traffic on South Carolina Highway 125 (SRS Road A),
U.S. Highway 278, SRS Road 1, and the CSX Railroad corridor.
The SRS contains 15 major production, service, and research and development (R&D) areas that
previously supported nuclear materials production and can support processing operations and waste
management activities. Major SRS facilities include five nuclear reactors, two chemical separations
plants, a fuel and target fabrication facility, the Defense Waste Processing Facility (DWPF), the
Replacement Tritium Facility, a heavy-water rework plant, and the Savannah River Technology Center
(SRTC), formerly called the Savannah River Laboratory. In addition, the University of Georgia
Research Foundation operates the Savannah River Ecology Laboratory (SREL) on the Site under
contract to DOE. Under an interagency agreement, the U.S. Forest Service operates the Savannah
River Forest Station, which manages the natural resources and secondary roads on the Site. These
facilities are in defined areas scattered across the Site. Each area is identified by a letter designation,
as summarized in Table 2-1. Figure 2-2 shows the locations of the principal SRS facilities. The
reactor, waste storage, and separations areas are at least 4 miles (6 kilometers) inside the nearest SRS
boundary.
The primary SRS facilities were related to the production of nuclear materials. M-Area
manufactured fuel and target components for shipment to the SRS reactors. Originally, the Site
operated five reactors; at present, all are in shutdown status. Shielded railroad cars transported
irradiated fuel to the F- or H-Area Canyon for the recovery of nuclear materials. The F- and H-Area
separations processes dissolve irradiated components in acid, and extract and separate the desired
nuclear materials. In H-Area, additional processes extract other products from irradiated components.
DOE neutralizes and stores the high-level liquid radioactive waste generated by the separations
facilities in underground tanks. DOE plans to process this waste into a borosilicate glass waste form
in the Defense Waste Processing Facility when that facility becomes operational, and to store this glass
waste form at the SRS until an offsite geological repository is available. [DOE has prepared a
Supplemental EIS related to Defense Waste Processing Facility operations (DOE 1994a).] In addition
to the underground waste storage tanks, DOE has established a centrally located 196-acre
(0.8-square-kilometer) site between F- and H-Areas, called E-Area, for the disposal of solid low-level
radioactive waste and the storage of transuranic (TRU) radioactive waste and mixed (hazardous and
radioactive) waste. The Site also has a central sanitary landfill and buildings in the Central Shops
Table 2-1. Description of functions and principal facilities at SRS areas.
Area Function Principal facilities
A Main DOE administration area, Main administration building, Savannah River
research laboratories Technology Center, Savannah River Ecology
Laboratory, powerhouse
B Wackenhut Services, Inc., Administration building, WSRC Engineering
administration area (security) building, WSRC training buildings
C One of five SRS reactors C-Reactor, training facilities, cooling basin
D Central powerhouse and heavy-water Powerhouse, heavy-water rework facility
rework
E Waste disposal and storage Solid Waste Disposal Facility
F Process plutonium F-Area Canyon, FB-Line, tank farm
G Various support functions Spread throughout the Site: railroad yard,
U.S. Forest Service installations
H Process uranium and tritium H-Area Canyon, HB-Line, Effluent Treatment
Facility, tank farm, Receiving Basin for Offsite
Fuels, Consolidated Incineration Facility
K One of five SRS reactors K-Reactor, cooling basins, cooling tower
L One of five SRS reactors L-Reactor, cooling basins
M Production of fuel and target Slug and target production facilities, effluent
assemblies treatment facility
N Receiving Central Shops
P One of five SRS reactors P-Reactor, cooling basins
R One of five SRS reactors R-Reactor, cooling basins
S Process high-level radioactive waste Defense Waste Processing Facility
TNX Applied research and development Analytical laboratory, Defense Waste Processing
Technology facilities, various mockups, effluent
treatment facilities
Z Waste treatment and handling Saltstone facility
(N-Area) for the storage of nonradioactive hazardous wastes and mixed waste. DOE is preparing an
EIS on waste management activities at the SRS (DOE 1995a).
The Site contains facilities for processing support and for research and development. These
include operational coal-fired powerhouses in A-, D-, and H-Areas that generate electricity and steam.
Figure 2-2. Location of principal SRS facilities (see Table 2-1). The largest powerhouse, which is in D-Area, produces electricity and sends process steam to C-, F-,
H-, and S-Areas through a 7-mile (11-kilometer) steam line. D-Area also contains the heavy-water
rework facility at which DOE purified the deuterium oxide (heavy water) used as the moderator and
coolant in SRS reactors. TNX-Area facilities study chemical and waste processing problems and test
production-scale equipment. Finally, A-Area facilities include the Savannah River Technology Center,
the Savannah River Ecology Laboratory, and the DOE and Westinghouse Savannah River Company
administrative offices.
The SRS employs approximately 20,000 people. Most of these employees work for
Westinghouse Savannah River Company and its subcontractors. The remainder work for DOE, the
Savannah River Ecology Laboratory, Wackenhut Services, Inc., the U.S. Forest Service, and other
contractors.
2.1.2 Site History
The U.S. Atomic Energy Commission (AEC), a DOE predecessor agency, selected the location
for the SRS in November 1950 after a study of more than 100 prospective sites. The government
selected E. I. du Pont de Nemours and Company, Inc., to build and operate the facility. Construction
began in February 1951; the basic plant was completed in 1956 at a cost of $1.1 billion, including the
land. On October 3, 1952, operations began with the startup of a unit of the heavy-water extraction
plant. Criticality occurred in the first production reactor on December 28, 1953.
In 1972, the AEC designated the SRS as the nation's first National Environmental Research Park.
Through the years, scientists have performed a wide range of investigations on the diverse habitats,
flora, and fauna of the Site.
2.1.3 Mission
The historic mission of the SRS was to serve the national security interests of the United States
by safely processing nuclear materials while protecting the health and safety of employees and the
public and protecting the environment. The SRS was responsible for producing tritium and special
nuclear materials for national defense. At present, it supports the viability of the weapons stockpile by
recycling limited-life components. The SRS also produces isotopes for nonweapons applications in the
nation's space program and for medical applications.
The SRS spent nuclear fuel mission is to manage DOE-owned spent fuel in a cost-effective way
that protects the safety of SRS workers, the public, and the environment. The goals of near-term
activities are the accurate quantification and characterization of DOE-owned spent nuclear fuel,
assessment of spent nuclear fuel storage facilities, elimination of current spent nuclear fuel storage
vulnerabilities, and identification of technologies and requirements for interim management and
ultimate disposition of spent nuclear fuel.
2.1.4 Management
The DOE Savannah River Operations Office manages the SRS; the Westinghouse Savannah
River Company operates the Site under contract to DOE. Westinghouse assumed operational
responsibility in April 1989 from E. I. du Pont de Nemours and Company, Inc., which had operated
the Site since 1951.
2.2 Regulatory Framework
This section summarizes the framework of environmental protection regulations applicable to
spent nuclear fuel management at the SRS. The framework is based on Federal and South Carolina
laws and one local ordinance, as discussed below. Volume 1 (Section 7.0) of this Environmental
Impact Statement (EIS) provides additional information on the major Federal environmental laws and
regulations, Executive Orders, and DOE Orders that apply to spent nuclear fuel management
alternatives.
2.2.1 Federal
The U.S. Environmental Protection Agency (EPA) has authorized South Carolina to implement
most provisions of the Clean Air Act, Resource Conservation and Recovery Act, and Clean Water Act
that apply to SRS spent nuclear fuel management. EPA Region IV has the lead responsibility for
Clean Air Act standards for radionuclide emissions from DOE facilities, imposing monitoring and
approval requirements on SRS spent nuclear fuel management activities that could result in
radionuclide emissions.
In addition, EPA Region IV has Resource Conservation and Recovery Act authority over
radioactive hazardous (mixed) waste management, affecting wastes from spent nuclear fuel processing.
EPA Region IV and the DOE Savannah River Operations Office have entered into a Federal Facility
Compliance Agreement on SRS mixed waste management.
The U.S. Army Corps of Engineers District Engineer for the Charleston District implements the
Clean Water Act Section 404 and the Rivers and Harbors Act permitting program for SRS spent
nuclear fuel construction activities that would affect U.S. waters.
In accordance with the Endangered Species Act, the SRS would consult with the U.S. Fish and
Wildlife Service, Charleston Field Office on impacts that spent nuclear fuel construction activities
could have on threatened and endangered species.
2.2.2 State
The South Carolina Department of Health and Environmental Control implements the following
State laws that would affect SRS spent nuclear fuel management activities:
- Pollution Control Act (nonradioactive emissions and discharges, and nonhazardous waste
management)
- Hazardous Waste Management Act (nonradioactive hazardous waste management)
- Safe Drinking Water Act
- Groundwater Use Act
- Stormwater Management and Sediment Reduction Act
The U.S. Army Corps of Engineers District Engineer for the Charleston District has an
agreement with the South Carolina Department of Health and Environmental Control whereby that
department issues Clean Water Act Section 401 water quality certifications. The South Carolina
Department of Health and Environmental Control also receives SRS reports in accordance with the
Emergency Planning and Community Right-To-Know Act.
The South Carolina State Department of Archives and History includes the State Historic
Preservation Office. In accordance with the National Historic Preservation Act, the SRS would consult
with the State Historic Preservation Officer on impacts that construction activities could have on
cultural resources.
2.2.3 Local
The only local requirement applicable to SRS spent nuclear fuel management is the Aiken
County Sediment Control Ordinance, which would affect construction activities.
2.3 Spent Nuclear Fuel Management Program at the Savannah River Site
This EIS addresses the management of approximately 2,742 metric tons of heavy metal (MTHM;
3,023 tons) of spent nuclear fuel that would be stored at various locations within the DOE Complex
over the next 40 years (1995-2035). At present, DOE has stored approximately 206.3 MTHM
(227.4 tons), or about 8 percent of this material, at the SRS. The spent nuclear fuel currently stored at
the SRS that DOE has included in the analyses in this document includes:
- 184.4 MTHM (203.3 tons) of Savannah River Defense Production [highly enriched uranium
(HEU) aluminum-clad fuels], including plutonium target material, and other aluminum-clad
fuels
- 4.6 MTHM (5.1 tons) of commercial spent fuel (primarily zirconium-clad)
- 11.9 MTHM (13.1 tons) of test and experimental reactor Zircaloy-clad fuel
- 5.4 MTHM (6.0 tons) of test and experimental reactor stainless steel-clad fuel
Spent nuclear fuel is currently stored in the Receiving Basin for Offsite Fuels (RBOF), in three
reactor disassembly basins, and in basins in F- and H-Canyons. Table 2-2 shows the quantity of spent
fuel stored at these facilities.
Table 2-2. SRS Fuel Inventory by Facility.
Facility Quantity (MTHM)
Receiving Basin for Offsite Fuel 60.73
L-Reactor Disassembly Basin 118.11
K-Reactor Disassembly Basin 3.32
P-Reactor Disassembly Basin 1.41
F-Canyon 22.63
H-Canyon 0.07
Total 206.27
Source: Wichmann (1995).
The F- and H-Area Canyons at the SRS are among the only remaining operable chemical
separations facilities of their kind in the DOE Complex. Each canyon has an associated storage basin
that serves as an interim staging area where reactor fuel bundles and targets await the Chemical
Separations Process. The basins currently contain 13 reactor fuel assemblies (H-Area) and aluminum-
clad targets (F-Area).
DOE has stored most of the remaining aluminum-clad spent nuclear fuel from SRS reactor
operations under water in concrete reactor storage basins. Three reactor disassembly basins (K-, P-,
and L-Reactors) contain reactor fuel and target material. These structures were built in the 1950s and
were not intended for the prolonged storage of radioactive materials. Wet (underwater) storage, while
potentially viable for stainless steel-clad fuel elements, is not satisfactory for aluminum-clad elements,
which are subject to corrosion and pitting.
In March 1992, chemical processing operations were suspended in the canyons to address a
potential safety concern. The concern was subsequently addressed but prior to resumption of
processing, the Secretary of Energy directed that defense related chemical separations activities (i.e.,
reprocessing) be phased out at the SRS. Since the decision, DOE has determined that further action
related to the disposition of nuclear material, including spent nuclear fuel, is subject to the National
Environmental Policy Act (NEPA) process. Non-safety related facility operations have remained shut
down with the exception of Pu-238 processing associated with the support of NASA missions.
As a result of these shut-downs, the canyons and the basins used for storage of spent nuclear fuel
and irradiated targets have a large inventory of in-process solutions and fuel and targets (respectively).
Some materials stored in the L- and K-Reactor disassembly basins have corroded, releasing fissile
materials to the pool water. DOE is preparing an environmental impact statement that will evaluate
risks that these and other SRS materials represent to the public and workers and will assess the
near-term need for the actions to stabilize these materials to ensure continued safe management
(DOE 1995b). These actions would take place over the short-term (about 10 years), until DOE can
make programmatic decisions on disposition.
DOE stores other spent fuel in the Receiving Basin for Offsite Fuels (RBOF) on the SRS. This
basin, which is in H-Area near the center of the Site, has been operating and receiving fuels of U.S.
origin since 1964. This 15,000-square-foot (1,393-square-meter) facility consists of an unloading
basin, two storage basins, a repackaging basin, a disassembly basin, and an inspection basin. The
basins and their interconnecting transfer canals hold about 500,000 gallons (1,893,000 liters) of water.
Spent fuel elements arrive in lead-lined casks weighing from 24 to 70 tons (about 22 to 64 metric
tons), which a crane lifts from a railroad car or truck trailer and places in the unloading basin. About
30 percent of the fuels in the Receiving Basin for Offsite Fuels consist of uranium clad in stainless
steel or Zircaloy, which SRS facilities cannot process without modifications.
2.4 Vulnerabilities Associated with SRS Spent Nuclear Fuel
In August 1993, the Secretary of Energy commissioned a comprehensive baseline assessment of
the environmental, safety, and health vulnerabilities associated with the storage of spent nuclear fuel in
the DOE complex. The purpose of this assessment was to determine the inventory and condition of
the Department's Reactor Irradiated Nuclear Material, which includes spent nuclear fuel and reactor
irradiated target material. The assessment also evaluated the condition of the facilities that store spent
fuel and identified the vulnerabilities and problems currently associated with these facilities.
Vulnerabilities in nuclear facilities are conditions or weaknesses that could lead to radiation exposure
to the public, unnecessary or increased exposure to workers, or release of radioactive materials to the
environment. Loss of institutional controls, such as a cessation of facility funding or reductions in
facility maintenance and control, could cause some vulnerabilities.
Based on this evaluation process DOE released a report to the Secretary of Energy, entitled Spent
Fuel Working Group Report on Inventory and Storage of the Department's Spent Nuclear Fuel and
other Reactor Irradiated Nuclear Materials and Their Environmental, Safety and Health
Vulnerabilities (i.e., "The Working Group Report," Volumes I, II, and III), to the public on
December 7, 1993 (DOE 1993). This report identified over 100 vulnerabilities associated with spent
fuel storage in the DOE complex, including 19 at the Savannah River Site. The report also determined
that five facilities and three burial grounds warranted priority attention from management to avoid
unnecessary increases in worker radiation exposure and cost during cleanup. The Savannah River Site
L- and K-Reactor Disasssembly Basins were among these facilities. The report grouped vulnerabilities
associated with each facility into three categories for management attention based on when corrective
action should be initiated: less than 1 year, 1 to 5 years, and more than 5 years.
After issuing the Working Group Report, DOE developed a Plan of Action to address all
vulnerabilities, taking into consideration currently available resources for implementation. The Plan of
Action is a consolidation of individual action plans designed to address each spent nuclear fuel
vulnerability in a manner that reflects the DOE (1) sense of urgency, (2) concern for worker
protection, (3) commitment to avoid or otherwise mitigate environmental impacts, and (4) need for
compatible long-term solutions.
The interim goal for the Savannah River Site reactor disassembly basins, pending completion of
the removal of the stored material, is the stabilization of basin conditions to reduce corrosion and to
address known vulnerabilites. The long-term goal of the action plan is a safe start of the removal of
reactor-irradiated nuclear material within a 5-year period, consistent with safe and environmentally
sound operations, including completion of appropriate NEPA review. These actions will lead to
mitigating the identified vulnerabilities while DOE pursues other courses of action.
The 19 vulnerabilities identified for the Savannah River Site now have complete Action Plans
(DOE 1994b, 1994c, 1994d). Table 2-3 lists SRS vulnerabilities by facility, tracking number, priority
categorization, and Action Plan status.
DOE is currently implementing a number of the 19 Action Plans. These actions have been
evaluated under the NEPA review process. The remaining corrective actions, those that will be carried
out through FY99, would also undergo NEPA review prior to implementation. Only one of these
outstanding actions, the construction of a dry storage facility, would likely require detailed NEPA
documentation (e.g., an EIS). The construction of such a facility is addressed programmatically in this
EIS as part of the Decentralization, 1992/1993 Planning Basis, Regionalization, and Centralization
alternatives. Construction of new facilities would require site-specific NEPA documentation, however.
Table 2-3. SRS vulnerabilities by facility, vulnerability, tracking number, priority categorization, and
Action Plan status.
Priority
Site/Facility Eight major Less than Greater than Action Plan
Vulnerability Number facilities wit1 year 1 year status
Description vulnerabilities
SRS/L-Reactor Disassembly Basin y Complete
SRS-01
Potential unmonitored buildup of radionuclide or fissile
materials in sand filters.
SRS/L-Reactor Disassembly Basin y Complete
SRS-04
Lack of authorization basis in operating the sand filter
cleanup system for L-Area Disassembly Basin.
SRS/Reactor Disassembly Basins y Complete
SRS-05
Corrosion of aluminum clad fuel, targets, and
components.
SRS/L-Reactor Disassembly Basins y Complete
SRS-06
Cesium-137 activity level in L-Basin.
SRS/L-Reactor Disassembly Basins y Complete
SRS-07
Determine whether gas bubbles release is a potential
hazard above the bucket storage area at L-Reactor.
SRS/K-, L-, P-Reactors y Complete
SRS-08
Lack of Reactor Authorization Basis.
SRS/K-Reactor Disassembly Basins y Complete
SRS-09
Corrosion of Mark 31 A and B target slugs in K and L
disassembly basins.
SRS/P-Reactor Disassembly Basins y Complete
SRS-10
Hoist Rod Corrosion
SRS/K-, L-Reactor Disassembly Basins y Complete
SRS-11
Reactor Disassembly Basin Safety Analysis Envelope.
SRS/L-Reactor Disassembly Basin y Complete
SRS-12
Inadvertent flooding of L-Reactor Disassembly Basin.
SRS/K-Reactor Disassembly Basin y Complete
SRS-13
Inadvertent flooding of K-Reactor Disassembly Basin.
SRS/P-Reactor Disassembly Basin y Complete
SRS-14
Inadvertent flooding of P-Reactor Disassembly Basin.
Table 2-3. (continued).
Priority
Site/Facility Eight major
Vulnerability Number facilities witLess than Greater than Action Plan
Description vulnerabilitie1 year 1 year status
SRS/RBOF; P-, R-, L-, C-, R-Reactors y Complete
SRS-15 (NOTE: RBOF is a less than 1 year
vulnerability)
Conduct of operations at reactor facilities and RBOF.
SRS/Receiving Basin for Offsite Fuel (RBOF) y Complete
SRS-16
Inadequate tornado protection at RBOF.
SRS/Receiving Basin for Offsite Fuel (RBOF) y Complete
SRS-17
Seismic vulnerability of RBOF.
SRS/H-Area Canyon y Complete
SRS-18
Seismic vulnerability of H-Area Canyon.
SRS/F-Area Canyon y Complete
SRS-19
Seismic vulnerability of F-Area Canyon.
SRS/K-, L-, P-Reactor Disassembly Basins and RBOF y Complete
SRS-20
Inadequate leak detection system in the underground
water-filled RINM storage basin.
SRS/L-, K-, P-Reactor Disassembly Basins y Complete
SRS-21
Inadequate seismic evaluation and potential inadequacies
of structures, systems, and components to withstand a
design basis event.
2.5 Representative Host Sites
DOE has identified two SRS areas as representative host sites for potential facilities related to the
implementation of programmatic decisions on spent nuclear fuel management (Figure 2-3):
- F- and H-Areas (considered together) for the modification or expansion of existing facilities,
new wet storage, and support facilities
- An undeveloped site for the construction of major new facilities, primarily an Expended
Core Facility or dry storage vault.
Figure 2-3. Representative host sites on Savannah River Site. 2.5.1 F- and H-Areas
These two areas contain most of the current spent nuclear fuel facilities and operations at the
SRS, including the Receiving Basin for Offsite Fuels. Therefore, DOE would focus future actions
under any of the alternatives in these areas as well, for cost-effectiveness and because construction
would occur in areas that had been previously disturbed.
F- and H-Areas are about 2 miles (3.2 kilometers) apart near the center of the SRS. The nearest
Site boundary is approximately 7.5 miles (12 kilometers) to the west. DOE uses the land within a
5-mile (8-kilometer) radius of the two areas either for industrial purposes associated with SRS
operations or as managed forest land. The closest facility to F- and H-Areas is the E-Area Solid
Waste Disposal Facility, which lies between the two areas (Figure 2-3). DOE uses this facility to
dispose of SRS solid low-level radioactive waste and to store TRU radioactive waste and mixed waste.
The F-Area separations facilities occupy about 420 acres (1.7 square kilometers). These facilities
were designed primarily for the recovery of plutonium-239 from irradiated and unirradiated feed
materials. DOE used the F-Area Canyon to dissolve target materials and produce solutions that
contained the various products extracted from fission products. Further processing converted the
products from solution to solid form for shipment off the Site. Large tanks in F-Area store high-level
liquid radioactive waste for future stabilization and disposal through the Defense Waste Processing
Facility.
H-Area facilities occupy about 395 acres (1.6 square kilometers). The H-Area Canyon processed
irradiated fuel elements or target assemblies from reactors. Primary operations included the dissolution
of irradiated targets and fuel tubes, chemical and physical separation, and purification of materials.
DOE stores high-level liquid waste in large tanks in H-Area, as in F-Area, for future processing and
disposal through the Defense Waste Processing Facility.
2.5.2 Undeveloped Representative Host Site
DOE has selected an undeveloped representative host site for the construction of new facilities
that F- or H-Area could not accommodate. This site is to the south and east of H-Area, adjacent to
SRS Road E and close to an existing railroad line, as shown in Figure 2-3. The SRS could make
connections to existing electricity, water, and steam networks with minimal additional construction.
The use of this site would have the advantage of consolidating spent nuclear fuel-related activities near
F- and H-Areas and close to the center of the SRS.
This site is representative of many available areas on the SRS that could support spent nuclear
fuel management activities. For example, DOE has identified a different representative site for the
possible construction of the Expended Core Facility for the management of naval spent nuclear fuel
(see Appendix D of Volume 1 of this Environmental Impact Statement). DOE would conduct a
detailed siting analysis before implementing any programmatic decision at the SRS. DOE would
assess, as necessary, the environmental consequences of the siting of any facilities as part of the site-
specific NEPA documentation.
3. SPENT NUCLEAR FUEL ALTERNATIVES
This chapter describes the five management alternatives for spent nuclear fuel that the
U.S. Department of Energy (DOE) has evaluated for the Savannah River Site (SRS) as part of
Volume 1 of this Environmental Impact Statement. These alternatives are:
1. No Action
2. Decentralization
3. 1992/1993 Planning Basis
4. Regionalization (with 2 subalternatives for the SRS)
5. Centralization (with 2 subalternatives for the SRS)
The activities covered by the alternatives range from maintaining the current inventory of spent
fuel at the SRS without receiving any more shipments (Alternative 1), through keeping the existing
inventory and accepting or sending off some limited shipments (Alternatives 2 through 4), to receiving
at the Site all DOE spent nuclear fuel and some from other sources (Alternative 5). DOE also
examined an option for shipping all spent nuclear fuel at the SRS to another location
(a variation of Alternatives 4 and 5). Table 3-1 summarizes the quantities of material that would be
received, shipped out, and ultimately managed at the SRS under the various alternatives. DOE has
assessed the aluminum-clad spent nuclear fuel separately from nonaluminum-clad fuel (i.e., stainless
steel and Zircaloy) because the options for managing them at the Site could be different as explained
in Section 3.1.
The analytical approach used in this document produces estimates of consequences that would be
as large as or larger than any that could occur or be expected under the alternatives and provides a
comparison of the impacts of the principal technologies for managing spent nuclear fuel at the SRS.
This chapter also provides an overview of the SRS management approach and describes the five
alternatives as they relate to the SRS (Sections 3.1 and 3.2). In addition, the chapter summarizes and
compares the potential environmental consequences of each alternative (Section 3.3).
Table 3-1. Quantities (MTHM)a of spent nuclear fuel that would be received, shipped, and managed
at the SRS under the five alternatives.b,c
Alternative Fuel Type Currently at Receive Ship Out Totals managed at
SRS SRS under this
alternative
1. No Action Aluminum 184.40 0.00 0.00 184.40
Nonaluminum 21.87 0.00 0.00 21.87
Totals 206.27 0.00 0.00 206.27
2. Decentralization Aluminum 184.40 11.02 0.00 195.42
Nonaluminum 21.87 2.60 0.00 24.47
Totals 206.27 13.62 0.00 219.89
3. 1992/1993 Planning Basis Aluminum 184.40 13.69 0.00 198.09
Nonaluminum 21.87 2.80 0.00 24.67
Totals 206.27 16.49 0.00 222.76
4. Regionalization - A Aluminum 184.40 28.69 0.00 213.09
(by fuel type) Nonaluminum 21.87 0.00 (21.87) 0.00
Totals 206.27 28.69 (21.87) 213.09
4. Regionalization - B Aluminum 184.40 19.93 0.00 204.33
(by location at SRS) Nonaluminum 21.87 30.42 0.00 52.29
Totals 206.27 50.35 0.00 256.62
4. Regionalization - B Aluminum 184.40 0.00 (184.40) 0.00
(by location, elsewhere) Nonaluminum 21.87 0.00 (21.87) 0.00
Totals 206.27 0.00 (206.27) 0.00
5. Centralization Aluminum 184.40 28.69 0.00 213.09
(at SRS) Nonaluminum 21.87 2,506.84 0.00 2,528.71
Totals 206.27 2,535.53 0.00 2,741.80
5. Centralization Aluminum 184.40 0.00 (184.40) 0.00
(elsewhere) Nonaluminum 21.87 0.00 (21.87) 0.00
Totals 206.27 0.00 (206.27) 0.00
a. To convert metric tons of heavy metal to tons, multiply by 1.1023.
b. Numbers may not sum due to rounding.
c. Source: Wichmann (1995).
3.1 SRS Management Approach
3.1.1 Management Options
DOE has evaluated three options for the management of spent nuclear fuel at the SRS under the
five alternatives considered for this EIS. These technical management options are wet storage or dry
storage of all fuels and the processing of aluminum-clad fuels. DOE could implement these options
individually or in combination under any of the five alternatives. DOE would base its selection of one
or more of these technical management options on additional analysis, including a separate SRS-
specific National Environmental Policy Act (NEPA) review based on this programmatic EIS.
3.1.1.1 Wet Storage. As described above in Section 2.3, the SRS currently maintains its
spent nuclear fuel in wet storage in the Receiving Basin for Offsite Fuels and several reactor basins.
Wet storage under the 40-year interim management plan (except under the No Action alternative)
would require that DOE construct a new wet storage pool at the SRS and move all fuel to this facility.
Prior to this transfer, DOE could place all the aluminum-clad fuel in stainless steel canisters to prevent
further corrosion and breakdown of the fuel cladding. The stainless steel- and Zircaloy-clad fuels
could also require canning. The SRS would monitor and maintain the water quality and the condition
of the fuel in the storage pool throughout the interim management period.
Under this wet storage option, the spent nuclear fuel would be in an interim storage form, which
could require further treatment depending on the DOE decision on its ultimate disposition.
3.1.1.2 Dry Storage. DOE currently has no dry storage facilities for spent nuclear fuel at the
Site. Dry storage of SRS aluminum-clad fuels under this management plan would require technology
development prior to the construction of a dry storage facility. Although such facilities exist at other
DOE sites and at commercial locations, DOE believes that the characteristics of SRS spent fuel are
sufficiently different to require some research and development before the design and construction of a
facility for this fuel. DOE would can all fuel before placing it into the dry storage vaults. It would
also have to maintain and monitor the facility for the remainder of the 40-year management period.
As with wet storage, the dry storage option would place the spent fuel into an interim storage
form that could require further treatment later depending upon DOE's decision on ultimate disposition.
3.1.1.3 Processing and Dry Storage. One method under this option would be for the SRS
to process existing aluminum-clad spent nuclear fuel through the existing separations facilities in the
F- and H-Area Canyons, and place the nonaluminum-clad fuels and any future receipts in dry storage.
The process using existing capability would result in the generation of both separated actinides
(e.g., uranium oxide), which would be stored on the site in existing facilities, and solutions of fission
products that would be placed in existing waste storage facilities for later conversion to a glassified
form through the Defense Waste Processing Facility (DWPF). DOE would maintain and monitor the
dry storage facility containing the nonaluminum-clad spent fuel. Variations of this processing option
are also possible, such as processing all the aluminum-clad fuel currently on the Site plus all that is
received from elsewhere, or developing the capability at the SRS for processing for vitrification
without chemical separations.
The process option selected for evaluation in this document is representative of possible
processing options that might be employed, but is not necessarily the one that DOE would select.
Detailed NEPA evaluations would be required to implement any spent nuclear fuel management plan
at the SRS.
3.1.2 Management Plan
Figure 3-1 summarizes DOE's overall plan for the interim management of aluminum-clad and
nonaluminum-clad fuels at the SRS. This flowchart shows actions for all alternatives except No
Action, as explained in Section 3.2.1.
3.1.2.1 Aluminum-clad Fuels. Depending on the alternative and option selected, DOE could
(within constraints of mission commitments) consolidate some aluminum-clad fuel in the Receiving
Basin for Offsite Fuels to take advantage of this facility's superior water quality and then move all
aluminum-clad fuel into dry storage, wet storage, or initiate processing (Figure 3-1). DOE could also
process aluminum-clad fuel without any consolidation work. Before moving the fuel into dry or wet
storage, DOE would place it in cans. DOE would hold the canned fuel or the stabilized products from
processing in storage for the 40-year interim management period until it decided their final disposition.
DOE would place aluminum-clad fuels received by the SRS from other locations in wet or dry
storage. DOE could not implement any of the options for aluminum-clad fuels, with the exception of
processing using existing SRS capabilities, without a technology development effort.
3.1.2.2 Nonaluminum-clad Fuels. DOE options for the management of nonaluminum-clad
fuels at the SRS are somewhat different, in that only dry or wet storage is considered (Figure 3-1).
The processing of these fuels at the Site is not an option because the SRS does not currently have
operational facilities capable of separating these materials. To improve aluminum-clad fuel storage,
DOE could consolidate the nonaluminum-clad fuel inventory in a reactor basin where the more
resistant stainless steel or Zircaloy cladding would be less susceptible to corrosion. The fuel would
remain there until DOE built new dry or wet storage facilities. DOE would then can the fuel and
move it into the new storage. DOE would place any nonaluminum-clad fuel received at the SRS after
completion of the new facilities directly into storage. The fuel would remain in this interim storage
until DOE decided its ultimate disposition.
Figure 3-1. Diagram of how SRS would manage aluminum-clad and nonaluminum-clad fuels. "Near- term Receipts" refers to the fuel that would be received before new wet or dry storage facilities are
available.
3.2 Description of Alternatives
3.2.1 Overview
Table 3-2 compares actions under each of the five alternatives. These actions relate to the
requirements for transportation, stabilization, facilities, and research and development that DOE would
address for each alternative. Transportation would include onsite movements as well as the receipt or
shipment of spent fuel. The consideration of facilities addresses not only new ones that could be
required, but also the use of existing structures and capabilities such as the F- and H-Area Canyons at
SRS. Finally, each alternative would involve some level of research and development on matters
related to spent nuclear fuel interim management (e.g., stabilization, transportation casks) and its
ultimate disposition.
Alternative 1 (No Action) addresses only the interim wet storage option, while the analysis of
Alternatives 2 through 5 considers three options: dry storage, wet storage, and processing of existing
aluminum-clad fuels and placing the other fuels into storage. In addition, Alternatives 4 and 5 include
an option for the shipment of spent nuclear fuel off the SRS. This analytical approach shows the
relative impact of viable interim storage technologies for the range of alternatives this EIS is
considering for the SRS. However, this information is not sufficient to support the selection of a
specific interim storage technology at the SRS because DOE has not completed site-specific research
and development for dry storage and wet storage methods or an evaluation of other processing options.
In addition, the specific quantities of offsite fuel that DOE would manage are subject to change. The
selection of an interim storage technology will be the subject of separate NEPA documentation specific
to the SRS.
Figure 3-2 is a matrix showing the types of facilities that would be required for each alternative
and option. The list includes those facilities already operating at the SRS (e.g., Receiving Basin for
Offsite Fuels) as well as potential facilities (e.g., fuel characterization facility). DOE considered these
facilities in its evaluation of the consequences of each alternative, as described in Chapter 5.
The alternatives described below address interim storage to 2035; further treatment of the spent
nuclear fuel would be necessary before DOE obtained a final disposable waste form. This EIS does
not address this additional treatment. However, DOE would carry out a full NEPA documentation for
any decision on final disposition of spent nuclear fuel.
Table 3-2. Actions required under each of the five alternatives at the SRS.
Alternative Transportation Stabilization Facilities Research and Development
1. No Action No shipments to or from the Site. Place aluminum-clad fuels that Store fuels in Receiving Basin for Continue existing spent nuclear
Limit onsite transfers to those are badly corroded and in Offsite Fuels and in an upgraded fuel-related research and
required for safe storage. danger of cladding failure in reactor basin. Requires no new development.
containers and return them to facilities.
wet storage.
2. Decentralization Receive about 13.6 MTHM (15.0 Can aluminum-clad fuels and Store fuels in Receiving Basin for Develop technology (canning
tons) of aluminum-clad and place them in wet or dry Offsite Fuels or upgraded reactor and storage design) to store SRS
nonaluminum-clad fuels. Limit storage or process existing fuel basin until new wet or dry storage aluminum-clad fuels in dry
onsite transfers to those required through F- and H-Canyons. facility is built. Requires new storage vault. Conduct research
for safe storage, consolidation, Can stainless-steel and characterization facility, new wet and pilot-scale operations to
and research and development. Zircaloy fuels and place in wet or dry canning facility, and new determine best technology for
Later relocate fuels to new wet or or dry storage. wet or dry storage facility. ultimate disposition of
dry storage facility or move aluminum-clad fuels.
aluminum-clad fuels to F- and
H-Canyons for processing.
3. 1992/1993 Planning Receive about 16.5 MTHM (18.2 Can aluminum-clad fuels and Store fuels in Receiving Basin for Develop technology (canning
Basis tons) of aluminum-clad and place them in wet or dry Offsite Fuels or upgraded reactor and storage design) to store SRS
nonaluminum-clad fuels. Limit storage or process existing fuel basin until new wet or dry storage aluminum-clad fuels in dry
onsite transfers to those required through F- and H-Canyons. facility is built. Requires new storage vault. Conduct research
for safe storage, consolidation, Can stainless steel and characterization facility, new wet and pilot-scale operations to
and research and development. Zircaloy fuels and place in wet or dry canning facility and new determine best technology for
Later relocate fuels to new wet or or dry storage. wet or dry storage facility. ultimate disposition of
dry storage facility, or move aluminum-clad fuels.
aluminum-clad fuels to F- and H-
Canyon for processing.
4. Regionalization - A Receive about 28.7 MTHM (31.6 Can aluminum-clad fuels and Store fuel in existing Receiving Develop technology (canning
(by fuel type at the tons) of aluminum-clad fuel. place them in wet or dry Basin for Offsite Fuels or and storage design) to store
SRS) Ship to Idaho National storage; or process existing upgraded reactor basin until new aluminum-clad fuels in dry
Engineering Laboratory about fuel through F- and wet or dry storage facility is storage vault. Conduct research
21.9 MTHM (24.1 tons) of H-Canyons. available, or until fuel is and pilot-scale operations to
stainless steel and Zircaloy fuel. processed. Requires new receiving determine best technology for
Relocate aluminum-clad fuels to and characterization facilities, new ultimate disposition of
Receiving Basin for Offsite wet or dry canning facilities, and aluminum-clad fuels.
Fuels, as necessary; then to new new wet or dry storage facilities.
wet or dry storage facilities, or
move aluminum-clad fuels to F-
and H-Canyon for processing.
4. Regionalization - B Receive approximately 50.4 Can aluminum-clad fuels and Store fuels in Receiving Basin for Develop technology (canning
(by location at the MTHM (55.6 tons) of spent fuel place them in wet or dry Offsite Fuels or upgraded reactor and storage design) to store SRS
SRS) from other locations. Limit storage; or process existing basin until new storage facility is aluminum-clad fuels in dry
onsite transfers to those required aluminum-clad fuels through available. Store new fuel storage vault. Conduct research
for safe storage, consolidation, F- and H-Canyons and store shipments in new wet or dry and pilot-scale operations to
and research and development. remaining fuel. Characterize storage facility. Requires new determine best technology for
Relocate fuels to new dry or wet and can fuel received from receiving, characterization and ultimate disposition of
storage facility or move offsite that is not in a form canning facilities, new wet or dry aluminum-clad fuels.
aluminum-clad fuel to F- and suitable for direct placement storage facility, and possibly a new
H-Canyons for processing. into storage. Expended Core Facility.
4. Regionalization - B Move all fuels to new Characterize and can all spent Store existing fuels in Receiving Develop technology for
(by location characterization facility prior to fuel prior to shipment. Basin for Offsite fuel and in a stabilization, canning, and
at another site) shipment offsite. Ship out about reactor basin until characterization shipment of degraded aluminum-
206.3 MTHM (227.4 tons) of and shipment offsite. Requires clad fuel.
spent fuel. new characterization facility.
5. Centralization (at Receive about 2,535.5 MTHM Can aluminum-clad fuels and Store fuel in Receiving Basin for Develop technology (canning
the SRS) (2,794.9 tons) of spent fuel from place them in wet or dry Offsite Fuels or in an upgraded and storage design) to store SRS
offsite. Limit onsite transfers to storage; or process existing reactor basin until new storage aluminum-clad fuels in dry
those required for safe storage, aluminum-clad fuels through facilities are available. Store new storage vault. Conduct research
consolidation, and research and F- and H-Canyons and store fuel shipments in new wet or dry and pilot-scale operations to
development. Relocate fuels to remaining fuels. Characterize storage facility. Requires new determine best technology for
new dry or wet storage facility or and can fuel received from receiving, characterization and ultimate disposition of spent
move aluminum-clad fuel to F- offsite that is not in a form canning facilities, new wet or dry nuclear fuels.
and H-Canyons for processing. suitable for direct placement in storage facility, and new Expended
storage. Core Facility.
5. Centralization (at Move all fuels to new Characterize and can all spent Store existing fuel in Receiving Develop technology for
another site) characterization facility prior to fuel prior to shipment. Basin for Offsite Fuel or in an stabilization, canning, and
shipment offsite. Ship out about upgraded reactor basin until shipment of degraded aluminum-
206.3 MTHM (227.4 tons) of characterization and shipment clad fuel.
spent fuel. offsite. Requires new
characterization facility.
Figure 3-2. Types of facilities required for each alternative. 3.2.2 Alternative 1 - No Action
3.2.2.1 Overview. This alternative deals only with the minimum actions that DOE would
deem necessary for the continued safe and secure management of spent nuclear fuel. It is not a status
quo condition. Rather, across its complex of facilities, DOE would maintain spent nuclear fuel close
to generation or current storage locations with no shipment between sites. Facility upgrades or
replacements and onsite fuel transfers would occur only to support safe and secure interim storage.
DOE would continue existing and new research and development activities for spent fuel interim
management. Stabilization activities would be limited only to those minimum actions required to store
spent nuclear fuel safely.
3.2.2.2 SRS Alternative 1 - Wet Storage. DOE would initiate the various SRS programs
and activities necessary to obtain optimum use of existing spent nuclear fuel facilities for the extended
storage of existing Site inventories totalling 206.3 metric tons (227.4 tons) of heavy metal (MTHM) in
the following quantities:
- 184.4 MTHM (203.3 tons) of Savannah River Defense Production [highly enriched uranium
(HEU) aluminum-clad fuels], including plutonium target material, and other aluminum-clad
fuels
- 4.6 MTHM (5.1 tons) of commercial spent nuclear fuel (primarily zirconium-clad)
- 5.4 MTHM (6.0 tons) of test and experimental reactor stainless steel-clad fuel
- 11.9 MTHM (13.1 tons) of test and experimental reactor Zircaloy-clad fuel
The goal of this program would be to relocate some aluminum-clad fuels to the Receiving Basin
for Offsite Fuels where precisely maintained water quality would prolong the storage life of these fuel
types. In addition, DOE would relocate a portion of the stainless steel- and Zircaloy-clad fuels to a
reactor basin, where their more resistant cladding would maintain fuel containment for an extended
period. These actions would be accomplished within the constraints of mission requirements.
The following describes one method that could be employed to improve the storage of
aluminum-clad fuel. Variations of this plan that would involve only the use of existing storage basins
are also possible.
- Select a reactor basin for upgrading and for the interim storage of SNF.
- Relocate aluminum-clad fuels from the selected reactor basin to other onsite basins to enable
cleaning and repair of the basin chosen for upgrade to improve water quality.
- Consolidate fuels in the Receiving Basin for Offsite Fuels to the extent possible.
- After cleaning and renovating the selected reactor basin, move a portion of the stainless steel
and Zircaloy-clad fuel assemblies now at the Receiving Basin for Offsite Fuels to the
renovated reactor basin.
- Move the aluminum-clad fuels temporarily stored at other locations to the Receiving Basin
for Offsite Fuels or the renovated reactor basin.
DOE will continue to place heavily corroded aluminum-clad fuel elements that could be in
danger of cladding failure into containers in the wet pool as required to minimize any spread of
materials throughout the pool. This action would be much simpler than canning the elements, which
would occur under the other alternatives.
This alternative would require no new facilities. DOE would continue existing spent nuclear
fuel-related research and development.
3.2.3 Alternative 2 - Decentralization
3.2.3.1 Overview. Under this alternative, DOE would maintain existing spent nuclear fuel in
storage at the current locations, and the SRS would receive some shipments of university fuel and
foreign fuel. This alternative differs from the No Action alternative by allowing significant facility
development and upgrades. DOE could transport fuel on the Site for safety, fuel consideration, or
research and development activities. In addition, DOE could undertake actions it deemed desirable,
though not essential, for safety and could perform spent nuclear fuel processing, treatment, research,
and development.
3.2.3.2 SRS Options 2a, 2b, and 2c. DOE analyzed three options specific to the SRS for
this alternative: Option 2a deals with dry storage, Option 2b deals with wet storage, and Option 2c
involves processing existing SRS aluminum-clad spent nuclear fuel and storing the remaining fuel.
The amount of spent fuel that the SRS would manage includes its current inventory, as described
above for Alternative 1, plus:
- 11.0 MTHM (12.0 tons) of aluminum-clad fuel
- 1.1 MTHM (1.2 tons) of stainless steel-clad fuel
- 0.7 MTHM (0.8 ton) of Zircaloy-clad fuel
- 0.8 MTHM (0.9 ton) of other experimental fuel
Under this alternative, SRS would manage a total of about 219.9 MTHM (242.4 tons) of spent
nuclear fuel. The SRS would receive spent fuel from research reactors as existing storage allowed and
as new storage was constructed.
3.2.3.2.1 Option 2a - Dry Storage - Under this option, DOE would store existing SRS
inventories in wet pools while developing the technology and constructing the necessary facilities to
examine, characterize, and can the fuels and transfer them to a new dry storage vault to await
treatment for final disposition.
The SRS would proceed with the fuel rearrangement plan described
above for Alternative 1 to provide acceptable storage conditions to minimize failures of the
aluminum-clad material before its placement in a dry-storage container.
Placement in a dry-storage facility would require a technology development program into DOE
capabilities to examine, characterize, and can aluminum-clad fuel elements before placing them in a
vault. In addition, the SRS would investigate technologies for the ultimate disposition of spent nuclear
fuel. In addition to a dry storage facility, the SRS would build new fuel receiving, characterization,
and dry canning facilities.
3.2.3.2.2 Option 2b - Wet Storage - Under this option, DOE could rearrange existing
spent nuclear fuel as described above for Alternative 1 to provide interim wet storage capacity while
constructing new facilities.
SRS could also modify this rearrangement plan to accept shipments of
spent fuel from offsite and place them directly into the Receiving Basin for Offsite Fuels, as
circumstances warrant. The new wet storage facilities required under this option would include the
capability to examine and characterize fuels and to can deteriorating fuels in a stainless steel package
for placement in the new pool. DOE would move all fuel to the new storage pool once it was
complete. SRS would build new fuel receiving, characterization, and wet-canning facilities as well as
a new wet storage pool. SRS would investigate technologies for the ultimate disposition of spent
nuclear fuel.
3.2.3.2.3 Option 2c - Processing and Storage - Under this option, SRS would
process existing aluminum-clad spent nuclear fuel to consolidate and stabilize the nuclear material for
storage in vaults, and would place the stainless steel- and Zircaloy-clad fuel and new receipts of
aluminum-clad fuel in dry storage.
The fuel would remain in the current wet pools while awaiting
processing or the construction of new dry storage facilities. DOE would use existing F- and H-Area
facilities to process the aluminum-clad fuel to safe, stable, consolidated forms.
The new facilities that the SRS would require under this option would be similar to those
described for dry storage (Option 2a), except they would be much smaller because the amount of fuel
to be stored would be small: only about 11.0 MTHM (12.0 tons) of aluminum-clad and about 24.5
MTHM (27.0 tons) of nonaluminum-clad fuel.
The SRS would investigate technologies required for the ultimate disposition of spent fuel.
3.2.4 Alternative 3 - 1992/1993 Planning Basis
3.2.4.1 Overview. This alternative assumes the continued transportation, receipt, processing,
and storage of spent nuclear fuel. Foreign and university research reactor spent nuclear fuel would be
sent to the INEL and the SRS. DOE would assess the construction of new facilities required to
accommodate current and projected spent nuclear fuel storage requirements. This alternative would
include activities related to the treatment of spent nuclear fuel, including research and development
and pilot programs to support future decisions on its ultimate disposition.
3.2.4.2 SRS Options 3a, 3b, and 3c. DOE analyzed the same three options for this
alternative as for Alternative 2: dry storage (Option 3a), wet storage (Option 3b), and the processing
of existing SRS aluminum-clad fuel and storing the remaining fuel (Option 3c). The quantities of fuel
would be somewhat greater than those for Alternative 2 because the options assume that the SRS
would manage its present inventory (see Alternative 1) plus approximately:
- 13.7 MTHM (15.1 tons) of aluminum-clad fuel
- 1.3 MTHM (1.4 tons) of stainless steel-clad fuel
- 0.7 MTHM (0.8 ton) of Zircaloy-clad fuel
- 0.8 MTHM (0.9 ton) of other experimental fuel
- a small amount (<0.1 ton) of commercial nonaluminum-clad fuel
The total spent nuclear fuel managed would equal about 222.8 MTHM (245.6 tons). The Site
would receive shipments of fuel from other locations as existing space allowed and as new facilities
were completed.
3.2.4.2.1 Option 3a - Dry Storage - The Site would store current inventories in
existing wet pools while developing technology and constructing facilities necessary to examine,
characterize, and can the fuels and transfer them to a new dry storage vault to await treatment for final
disposition.
The actions that SRS would undertake under this option and the new facilities to be constructed
would be the same as those described for Option 2a - Dry Storage under Alternative 2
(Decentralization) in Section 3.2.3.2.1.
3.2.4.2.2 Option 3b - Wet Storage - DOE could rearrange existing spent nuclear fuel
as described in Alternative 1 above to provide interim wet storage capacity while building new
facilities.
The Site could also accept new shipments directly into the Receiving Basin for Offsite
Fuels, as required. The actions that SRS would undertake under this option, and the new facilities to
be constructed, would be the same as those described for Option 2b - Wet Storage under Alternative 2
(Decentralization) in Section 3.2.3.2.2.
3.2.4.2.3 Option 3c - Processing and Storage - Under this option, the SRS would
process existing aluminum-clad spent nuclear fuel and would place the stainless steel- and Zircaloy-
clad fuel and new receipts of aluminum-clad fuel in storage as described for Option 2c - Processing
under Alternative 2 (Decentralization) in Section 3.
2.3.2.3. The requirements for new facilities and for
technology development would also be the same.
3.2.5 Alternative 4 - Regionalization
3.2.5.1 Overview. This alternative has two subalternatives. The first (Regionalization A)
would involve the distribution of existing and new spent nuclear fuel among DOE sites based
primarily on the similarity of fuel type, although DOE would also consider transport distances,
available processing capabilities, available storage capabilities, or a combination of these factors.
Under this subalternative, SRS would receive all aluminum-clad fuel and would transfer its existing
inventory of stainless steel- and Zircaloy-clad fuel to another DOE site. The SRS would manage a
total of about 213.1 MTHM (234.9 tons) of spent fuel under the Regionalization A subalternative.
The second subalternative (Regionalization B) would require DOE to consolidate all existing and
new spent fuel at two sites - one to the east of the Mississippi River and one to the west -
depending on the location or generation site of the fuel. Under this alternative, the SRS would either
receive all spent nuclear fuel in the east [approximately 256.6 MTHM (282.9 tons)] or ship its current
inventory offsite to the Oak Ridge Reservation in Tennessee. An additional option if SRS becomes
the Eastern Regional Site is for DOE to construct an Expended Core Facility at the SRS to manage
some Naval fuel. This option is described in Appendix D of Volume 1 of this EIS.
Under either subalternative, DOE would undertake facility upgrades, replacements, and additions
as appropriate. This alternative would include research and development and pilot programs to support
current management and future decisions on spent fuel disposition.
3.2.5.2 SRS Options 4a, 4b, and 4c (Regionalization A). DOE analyzed three options
for the regionalization of fuels by fuel type: dry storage (Option 4a), wet storage (Option 4b) and
processing of existing SRS aluminum-clad fuels and storing the remaining fuel (Option 4c). This
subalternative assumes that the SRS would manage:
- Its current inventory of 184.4 MTHM (203.3 tons) of aluminum-clad fuels, plus
- Approximately 28.7 MTHM (31.6 tons) of research reactor aluminum-clad fuel from other
sites
The SRS would ship to the Idaho National Engineering Laboratory approximately:
- 5.4 MTHM (6.0 tons) of stainless steel-clad fuel
- 4.6 MTHM (5.1 tons) of commercial nonaluminum-clad fuel
- 11.9 MTHM (13.1 tons) of Zircaloy-clad spent fuel
DOE would manage a total of about 213.1 MTHM (234.9 tons) of spent nuclear fuel at the SRS
under this subalternative. The site would receive shipments from other locations as existing space
became available and as it shipped the nonaluminum-clad fuel.
3.2.5.2.1 Option 4a - Dry Storage - The actions that the SRS would undertake under
this option, and the new facilities to be constructed, would be the same as for those described for
Option 2a - Dry Storage under Alternative 2 (Decentralization) in Section 3.
2.3.2.1.
This option would require an extensive research and development program into capabilities to
examine, characterize, and can the SRS aluminum-clad fuel for dry storage.
3.2.5.2.2 Option 4b - Wet Storage - The SRS would carry out the same actions and
construct the same types of facilities under this option as it would for Option 2b - Wet Storage under
Alternative 2 (Decentralization) as described in Section 3.
2.3.2.2. Research and development activities
would also be similar to those conducted under this Decentralization alternative, except the SRS would
not perform studies on nonaluminum-clad fuels.
3.2.5.2.3 Option 4c - Processing and Storage - Under this option, the SRS would
process the existing aluminum-clad fuel as described for Option 2c - under Alternative 2
(Decentralization) and place the aluminum-clad fuel received from offsite into wet storage.
The
requirements for new construction would be different than in Option 2c, in that dry storage facilities
would not be required because the nonaluminum-clad fuels would be shipped off the site. The small
amount of aluminum-clad fuel to be received could be more readily stored in pools rather than
developing new dry storage. Therefore, Option 4c would require DOE to construct a new fuel
receiving, wet canning and wet storage facility to manage the fuel received after the major processing
operations are completed. These facilities would be much smaller than those required for other
alternatives.
3.2.5.3 SRS Options 4d, 4e, 4f, and 4g (Regionalization B). DOE analyzed the same
three options for the regionalization of spent fuel on the basis of geographic location as for the other
alternatives: dry storage (Option 4d), wet storage (Option 4e), and processing of existing
aluminum-clad fuel and storing the remaining fuel (Option 4f). In addition, it assessed the option of
shipping all SRS inventory offsite (Option 4g).
The amount of material that the SRS would manage if all the spent fuel in the East were shipped
to the Site would total about 256.6 MTHM (282.9 tons). This would include the current SRS
inventory of about 206.3 MTHM (227.4 tons) as detailed in Section 3.2.2 plus:
- 19.9 MTHM (21.9 tons) of aluminum-clad fuel
- 26.7 MTHM (29.4 tons) of commercial nonaluminum-clad fuel
- 1.0 MTHM (1.1 ton) of stainless steel-clad fuel
- 1.3 MTHM (1.4 tons) of experimental Zircaloy-clad fuel
- 1.4 MTHM (1.5 tons) of other experimental fuel
The activities that DOE would have to undertake at the SRS, and the facilities that it would have
to build, under the dry storage, wet storage, or processing options would be very similar to those
required for the Decentralization alternative (Section 3.2.3). The difference would be that the size of
the storage facilities would be somewhat greater because the amount of fuel to be managed would be
larger [256.6 MTHM (282.9 tons) versus 219.9 MTHM (242.4 tons)]. In addition, DOE would
conduct additional research and development on the other fuel types that SRS would manage under
these options.
3.2.5.3.1 Option 4d - Dry Storage - The actions that the SRS would undertake under
this option, and the new facilities to be constructed, would be similar to those described for
Option 2a - Dry Storage under Alternative 2 (Decentralization) in Section 3.
2.3.2.1. This option
would require an extensive research and development program into capabilities to examine,
characterize, and can the SRS aluminum-clad fuel for dry storage.
3.2.5.3.2 Option 4e - Wet Storage - The SRS would carry out the same actions and
construct the same types of facilities under this option as it would for Option 2b - Wet Storage under
Alternative 2 (Decentralization) as described in Section 3.
2.3.2.2. Research and development activities
would also be similar to those conducted under this Decentralization alternative.
3.2.5.3.3 Option 4f - Processing and Storage - Under this option, the SRS would
process the existing aluminum-clad fuel and place nonaluminum-clad fuel and aluminum-clad fuel
received from offsite in dry storage as described for Option 2c - Processing with storage under
Alternative 2 (Decentralization).
The requirements for new facilities and for research and development
would also be similar.
3.2.5.3.4 Option 4g - Shipment Off the Site - Under this option, the SRS would ship
its current inventory of about 206.
3 MTHM (227.4 tons) to the Oak Ridge Reservation. The activities
and facilities required for this option are the same as those described below for Option 5d of the
Centralization alternative (Section 3.2.6.2.4).
3.2.6 Alternative 5 - Centralization
3.2.6.1 Overview. Under this alternative, DOE would collect all current and future spent
nuclear fuel inventories from DOE sites, the Navy, and other sources at a single location for
management until final disposition. DOE would construct new facilities at the centralized site to
accommodate the increased inventories. The originating sites would characterize and stabilize their
spent nuclear fuel before shipping. They would then close their spent fuel facilities. This alternative
would include the centralization of activities related to the treatment of spent nuclear fuel, including
research and development and pilot programs to support future decisions on its disposition.
3.2.6.2 SRS Options 5a, 5b, 5c, and 5d. DOE analyzed four options for this alternative.
Three deal with shipping all DOE spent nuclear fuel to the SRS for disposition and management in
dry storage (Option 5a), wet storage (Option 5b), or by processing existing aluminum-clad fuel and
storing the remaining fuel (Option 5c). The fourth case involves the shipment of all SRS fuel off the
Site to another location (Option 5d). Options 5a, 5b, and 5c concern the following fuels:
- 65.2 MTHM (71.7 tons) of naval fuel
- 213.1 MTHM (234.9 tons) of aluminum-clad fuel
- 2103.2 MTHM (2,318.4 tons) of Hanford defense fuel
- 27.6 MTHM (30.4 tons) of graphite fuel
- 156.5 MTHM (172.5 tons) of commercial nonaluminum-clad fuel
- 96.5 MTHM (106.4 tons) of experimental stainless steel-clad fuel
- 78.0 MTHM (86.0 tons) of Zircaloy-clad fuel
- 1.7 MTHM (1.9 tons) of other fuel types
DOE would manage a total of about 2,741.8 MTHM (3,022.3 tons) of spent nuclear fuel at the
SRS under the first three options. Options 5a and 5b would involve storing all the fuel on the Site.
Option 5c would require processing the existing aluminum-clad fuel [184.4 MTHM (203.3 tons)] and
placing the remaining nonaluminum-clad SRS fuels and all fuel received from other locations
[2,557.4 MTHM (2,819.0 tons)] into dry storage. The SRS could accept shipments from offsite
sources and place them in storage as it built new facilities and transferred the onsite inventory.
Under Option 5d, shipments leaving the Site would amount to about 206.3 MTHM (227.4 tons),
which is equal to the inventory of spent nuclear fuel at the SRS under Alternative 1.
3.2.6.2.1 Option 5a - Dry Storage - The actions that the SRS would undertake under
this option would be the same as those described for Option 2a - Dry Storage under Alternative 2
(Decentralization) in Section 3.
2.3.2.1. However, the number and size of the new facilities needed to
implement this centralization option would be much greater because of the larger volume of fuel that
the Site would manage. In addition, DOE would have to build a new Expended Core Facility at the
SRS to examine and characterize the naval fuels.
This option would require an extensive research and development program into capabilities to
examine, characterize, and can SRS and other fuel types before their placement in a dry storage vault.
DOE would also carry out research and development into other aspects of the management of the
spent fuels, including those related to its ultimate disposition.
3.2.6.2.2 Option 5b - Wet Storage - Under this option, DOE would undertake actions
similar to those described in Section 3.
2.3.2.2 for Option 2b - Wet Storage under Alternative 2. As
with Option 5a (Dry Storage), the SRS would have to build major new facilities to manage the large
volume of fuel it would receive. DOE would also have to build a new Expended Core Facility at the
SRS. Research and development would be greatly expanded as well.
3.2.6.2.3 Option 5c - Processing and Storage - DOE would process the current
inventory of aluminum-clad spent fuel under this option in the same manner as described for the other
alternatives.
All other fuel onsite and all fuel received from elsewhere would be canned and placed in
new dry storage facilities. The SRS would shut down the F- and H-Area separations facilities after
processing the existing inventory of aluminum-clad fuel. Thereafter, any aluminum-clad fuel sent to
the SRS would be placed in dry storage.
This option would require major new facilities, including a new Expended Core Facility. DOE
would also conduct extensive research and development in spent fuel management.
3.2.6.2.4 Option 5d - Shipment Off the Site - DOE would consolidate and prepare
all spent nuclear fuel on the SRS for shipment to another DOE site; this would require the construction
of a new fuel characterization facility.
Some fuels could require canning before shipment. SRS would
use existing facilities to accomplish this. DOE would then close all SRS spent nuclear fuel-related
facilities.
DOE would conduct research and development into methods of stabilizing, canning, and
transporting aluminum-clad fuels, particularly that which is corroded or otherwise degraded.
3.3 Comparison of Alternatives
Table 3-3 summarizes the environmental consequences of the five alternatives. Chapter 5
presents detailed descriptions of these consequences.
In general, the levels of impacts associated with Alternatives 1 through 4 would be similar
because the amounts of spent nuclear fuel that DOE would manage at the SRS under these cases
would be approximately the same [e.g., about 206 to 257 MTHM (227 to 283 tons)] and activities
would extend throughout the full 40-year management period. The lowest level of impact at SRS
would occur under Option 4g or Option 5d (Regionalization or Centralization at another site) because
DOE would ship the SRS spent fuel off the Site well before the management period ended in 2035.
Alternative 5, under which DOE would ship all spent nuclear fuel to the SRS, would result in the
greatest onsite impacts; the Site would have to manage approximately 2,741.8 MTHM (3,022.3 tons)
of spent fuel.
Table 3-3. Comparison of impacts for the five alternatives.
ALTERNATIVE 1 - NO ACTION
Option 1
Wet Storage
Land Use No new facilities would be required.
Socioeconomics No new operations jobs and only about 50 construction
jobs would be created.
Cultural Resources No new construction would be carried out. No impacts
are anticipated.
Aesthetics and Facilities are in an existing industrial area not
Scenic Resources visible from public access roads or from off the Site.
No impacts are anticipated. Emissions would not impact
visibility.
Geology No minerals of economic value are in affected area. No
impacts are anticipated.
Air Resources Emissions of criteria air pollutants and toxic air
pollutants would be only a small fraction of air quality
standards.
Water Resources This option would not require use of additional surface
water beyond the 75.7 billion liters (20 billion
gallons) per year that the SRS withdraws at present.
This option would not require withdrawals of additional
groundwater beyond the 14.0 billion liters (3.7 billion
gallons) per year the SRS uses. Activities related to
this option currently use about 35.1 million liters (9.3
million gallons) of groundwater per year. Impacts would
be minimal.
No perennial streams or other surface waters would be
affected.
Accidental releases could contaminate shallow
groundwater that is not a source for drinking water or
domestic use. Releases would not affect surface streams
or drinking water aquifers.
Ecological Minor disturbance of wildlife due to traffic would
Resources occur.
No wetlands or threatened or endangered species would be
affected.
Noise The only noise experienced by offsite populations would
be generated by employee traffic and by truck and rail
deliveries. There would be no change in traffic noise
impacts.
Traffic and This option would not increase site traffic.
Transportation
Number of LCFf, normal transport:
Worker: 6.0 x 10-4
Public: 7.0 x 10-5
Occupational and Maximum LCFf probabilities:
Public Health and Worker: 4 x 10-5
Safety Offsite population: 4 x 10-14 (air)
(Radiological) 1 x 10-14 (water)
Annual LCFf incidences:
Worker: 8 x 10-5
Offsite population: 2 x 10-9
Table 3-3. (continued).
Option 1
Wet Storage
Occupational and Hazard index:
Public Health and Worker: 2 x 10-6
Safety Maximally exposed individual: 2 x 10-7
(Nonradiological)
Utilities and Minimal changes in demand for electricity, steam,
Energy domestic water and wastewater treatment would occur.
Current SRS capacities are adequate for these additions.
Impacts would be minimal.
Materials and Waste Annual average volume of waste generated (cubic
Management meters)b:
LLW: 400
TRU: 17
HLW: 0.4
No impact on site waste management capacities.
Accidentsc Greatest point estimate of riskd:
Worker: Data not calculatede
Colocated worker: 7.7 x 10-7
Maximally exposed individual: 1.6 x 10-7
Offsite population: 1.4 x 10-3
a. Not applicable.
b. LLW = low-level waste; TRU = transuranic waste; HLW = high-level
waste.
c. Data is provided as adjusted point estimates of risk by receptor group
to demonstrate a relative comparison of each alternative on an option-
by-option basis. The adjusted values were taken from Tables 5-27
through 5-29.
d. Units for adjusted point estimates of risk are given in terms of
potential fatal cancers per year.
e. The safety analysis reports from which information was extracted were
written before issuance of DOE Order 5480.23; previous orders did not
require the inclusion of workers.
f. LCF = latent cancer fatalities.
Table 3-3. (continued).
ALTERNATIVE 2 - DECENTRALIZATION
Option 2a Option 2b Option 2c
Dry Storage Wet Storage Processing
Land Use Most new Same as Option 2a. Same as Option 2a.
construction would
be in parts of F-
and H-Areas already
dedicated to
industrial use.
Impacts would be
minimal.
Socioeconomic Operations jobs Same as Option 2a. Operations jobs
s would be filled by would be filled by
current employees. current employees.
A maximum of about A maximum of about
600 construction 550 construction
jobs would be jobs would be
created. created.
Cultural Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Aesthetics Same as Option 1. Same as Option 1. Same as Option 1.
and Scenic
Resources
Geology Same as Option 1. Same as Option 1. Same as Option 1.
Air Resources Same as Option 1. Same as Option 1. Same as Option 1.
Water New withdrawals of New withdrawals of New withdrawals of
Resources approximately 6.1 approximately 7.2 approximately 311
million liters (1.6 million liters (1.9 million liters
million gallons) million gallons) (82.2 million
per year of cooling per year of cooling gallons) per year
water from Savannah water from Savannah of cooling water
River would be River would be from Savannah
required. Impacts required. Impacts River would be
would be minimal. would be minimal. required.
Impacts would be
Additional minimal.
Additional groundwater
groundwater withdrawals would Same as Option 2a.
withdrawals would total about
total about 50.6 million liters
48.7 million liters (13.4 million
(12.9 million gallons) per year.
gallons) per year. Impacts would be
Impacts would be minimal. No perennial
minimal. streams or other
No perennial surface waters
No perennial streams or other would be affected.
streams or other surface waters
surface waters would be affected. Accidental
would be affected. releases could
Accidental releases contaminate
Accidental releases could contaminate shallow
could contaminate shallow groundwater groundwater that
shallow groundwater that is not used as is not used as a
that is not used as a source for source for
a source for drinking water or drinking water or
drinking water or domestic use. domestic use.
domestic use. Releases would not Releases would not
Releases would not affect surface affect surface
affect surface streams or drinking streams or
streams or drinking water aquifers. drinking water
water aquifers. aquifers.
Table 3-3. (continued).
Option 2a Option 2b Option 2c
Dry Storage Wet Storage Processing
Ecological Small increase in Same as Option 2a. Small increases in
Resources traffic would cause traffic would cause
slight increase in small increase in
road kills and in road kills and in
disturbance of disturbance of
wildlife due to wildlife due to
noise. Impacts Same as Option 2a. noise. Impacts
would be minimal. would be minimal.
No wetlands or Same as Option 2a.
threatened or
endangered species
would be affected.
Noise Only noise Same as Option 2a. Same as Option 2a.
experienced by
communities would be
generated by
employee traffic and
by truck and rail
deliveries.
Changes in traffic
levels are expected
to result in only
very small changes
in noise impacts.
Traffic and This option would Same as Option 2a. This option would
Transportati increase site increase site
on traffic slightly. traffic slightly.
Number of LCFg, Number of LCFg,
normal transport: normal transport:
Worker: 1.0 x 10-3 Worker: 2.1 x 10-4
Public: 1.2 x 10-4 Public: 1.9 x 10-5
Occupational Maximum LCFg Maximum LCFg Maximum LCFg
and Public probabilities: probabilities: probabilities:
Health and Worker: 3 x 10-5 Worker: 4 x 10-5 Worker: 6 x 10-5
Safety Offsite population: Offsite Offsite
(Radiologica 4 x 10-14 (air) population: population:
l) 1 x 10-14 (water) 5 x 10-14 (air) 2 x 10-7 (air)
2 x 10-14 (water) 6 x 10-8 (water)
Annual LCFg
incidences: Annual LCFg Annual LCFg
Worker: 7 x 10-5 incidences: incidences:
Offsite population: Worker: 8 x 10-5 Worker: 3 x 10-2
2 x 10-9 Offsite Offsite
population: 2 x 10-9 population: 8 x 10-3
Occupational Same as Option 1. Same as Option 1. Hazard index:
and Public Worker: 6 x 10-3
Health and Maximally exposed
Safety individual: 5 x
(Nonradiolog 10-4
ical)
Utilities Requirements would Same as Option 2a. Very similar to
and Energy increase 3 to 7 Option 2a.
percent above
present levels.
Current SRS
capacities are
adequate for these
increases.
Materials Annual average Same as Option 2a. Annual average
and Waste volume of waste volume of waste
Management generated (cubic generated (cubic
meters)b: meters)b:
LLW: 400 LLW: 800
TRU: 18 TRU: 19
HLW: 0.4 HLW: 2.3c
No impact on site No impact on site
capacities. capacities.
Table 3-3. (continued).
Option 2a Option 2b Option 2c
Dry Storage Wet Storage Processing
Accidentsd Greatest point Greatest point Greatest point
estimate of riske: estimate of riske: estimate of riske:
Worker: Data not Worker: Data not Worker: Data not
calculatedf calculatedf calculatedf
Colocated worker: Colocated worker: Colocated worker:
1.6 x 10-6 1.7 x 10-6 7.7 x 10-7
Maximally exposed Maximally exposed Maximally exposed
individual: individual: individual: 1.6 x
3.3 x 10-7 3.5 x 10-7 10-7
Offsite population: Offsite Offsite
2.8 x 10-3 population: 3.0 x population: 1.4 x
10-3 10-3
a. NA = not applicable.
b. LLW = low-level waste; TRU = transuranic waste; HLW = high-level
waste.
c. High-level waste will be generated only during approximately the first
10 years.
d. Data is provided as adjusted point estimates of risk by receptor group
to demonstrate a relative comparison of each alternative on an option-
by-option basis. The adjusted values were taken from Tables 5-27
through 5-29.
e. Units for adjusted point estimates of risk are given in terms of
potential fatal cancers per year.
f. The safety analysis reports from which information was extracted were
written before issuance of DOE Order 5480.23; previous orders did not
require the inclusion of workers.
g. LCF = latent cancer fatalities.
Table 3-3. (continued).
ALTERNATIVE 3 - 1992/1993 PLANNING BASIS
Option 3a Option 3b Option 3c
Dry Storage Wet Storage Processing
Land Use Same as Option 2a. Same as Option 2a. Same as Option 2a.
Socioeconomi Same as Option 2a. Operations jobs Same as Option 2c.
cs would be filled by
current employees.
A maximum of about
650 construction
jobs would be
created.
Cultural Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Aesthetics Same as Option 1. Same as Option 1. Same as Option 1.
and Scenic
Resources
Geology Same as Option 1. Same as Option 1. Same as Option 1.
Air Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Water Same as Option 2a. Same as Option 2b. Same as Option 2c.
Resources
Ecological Same as Option 2a. Same as Option 2a. Same as Option 2c.
Resources
Noise Same as Option 2a. Same as Option 2a. Same as Option 2a.
Traffic and Same as Option 2a. Same as Option 2a. Same as Option 2c.
Transportati
on
Occupational Same as Option 2a. Same as Option 2b. Same as Option 2c.
and Public
Health and
Safety
(Radiologica
l)
Occupational Same as Option 1. Same as Option 1. Same as Option 2c.
and Public
Health and
Safety
(Nonradiolog
ical)
Utilities Same as Option 2a. Same as Option 2a. Very similar to
and Energy Option 2a.
Materials Same as Option 2a. Same as Option 2a. Annual average
and Waste volume of waste
Management generated (cubic
meters)a:
LLW: 750
TRU: 19
HLW: 1.7b
No impact on site
capacities.
Accidentsc Greatest point Same as Option 3a. Greatest point
estimate of riskd: estimate of riskd:
Worker: Data not Worker: Data not
calculatede calculatede
Colocated worker: Colocated worker:
1.9 x 10-6 1.1 x 10-6
Maximally exposed Maximally exposed
individual: individual:
4.0 x 10-7 2.3 x 10-7
Offsite Offsite
population: 3.4 x population: 2.0 x
10-3 10-3
a. LLW = low-level waste; TRU = transuranic waste; HLW = high-level
waste.
b. High-level waste will be generated only during approximately the first
10 years.
c. Data is provided as adjusted point estimates of risk by receptor group
to demonstrate a relative comparison of each alternative on an option-
by-option basis. The adjusted values were taken from Tables 5-27
through 5-29.
d. Units for adjusted point estimates of risk are given in terms of
potential fatal cancers per year.
e. The safety analysis reports from which information was extracted were
written before issuance of DOE Order 5480.23; previous orders did not
require the inclusion of workers.
Table 3-3. (continued).
ALTERNATIVE 4 - REGIONALIZATION A (By Fuel Type)
Option 4a Option 4b Option 4c
Dry Storage Wet Storage Processing
Land Use Same as Option 2a. Same as Option 2a. Same as Option 2a.
Socioeconom Same as Option 3b. Same as Option 3b. Same as Option 2c.
ics
Cultural Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Aesthetics Same as Option 1. Same as Option 1. Same as Option 1.
and Scenic
Resources
Geology Same as Option 1. Same as Option 1. Same as Option 1.
Air Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Water Same as Option 2a. Same as Option 2b. Very similar to
Resources Option 2c.
Ecological Same as Option 2a. Same as Option 2a. Same as Option 2c.
Resources
Noise Same as Option 2a. Same as Option 2a. Same as Option 2a.
Traffic and Same as Option 2a. Same as Option 2a. Same as Option 2c.
Transportat
ion
Occupationa Same as Option 2a. Same as Option 2b. Maximum LCFa
l and probabilities:
Public Same as Option 2c.
Health and
Safety Annual LCFa
(Radiologic incidences:
al) Worker: 3 x 10-2
Offsite
population: 9 x 10-3
Occupationa Same as Option 1. Same as Option 1. Same as Option 2c.
l and
Public
Health and
Safety
(Nonradiolo
gical)
Utilities Very similar to Same as Option 2a. Very similar to
and Energy Option 2a. Option 2a.
Materials Same as Option 1. Same as Option 1. Annual average
and Waste volume of waste
Management generated (cubic
meters)b:
LLW: 790
TRU: 18
HLW: 2.3c
No impact on site
capacities.
Table 3-3. (continued).
Option 4a Option 4b Option 4c
Dry Storage Wet Storage Processing
Accidentsd Greatest point Same as Option 3a. Greatest point
estimate of riske: estimate of riske:
Worker: Data not Worker: Data not
calculatedf calculatedf
Colocated worker: Colocated worker:
2.1 x 10-6 1.3 x 10-6
Maximally exposed Maximally exposed
individual: individual:
4.4 x 10-7 2.8 x 10-7
Offsite population: Offsite population:
3.7 x 10-3 2.4 x 10-3
a. LCF = latent cancer fatalities.
b. LLW = low-level waste; TRU = transuranic waste; HLW = high-level
waste.
c. High-level waste will be generated only during approximately the first
10 years.
d. Data is provided as adjusted point estimates of risk by receptor group
to demonstrate a relative comparison of each alternative on an option-
by-option basis. The adjusted values were taken from Tables 5-27
through 5-29.
e. Units for adjusted point estimates of risk are given in terms of
potential fatal cancers per year.
f. The safety analysis reports from which information was extracted were
written before issuance of DOE Order 5480.23; previous orders did not
require the inclusion of workers.
Table 3-3. (continued).
ALTERNATIVE 4 - REGIONALIZATION B (By Location)a
Option 4d Option 4e Option 4f
Dry Storage Wet Storage Processing
Land Use Same as Option 2a. Same as Option 2a. Same as Option 2a.
Socioeconom Operations jobs Operations jobs Same as Option 3b.
ics would be filled by would be filled by
current employees. current employees.
A maximum of about A maximum of about
700 construction 800 construction
jobs would be jobs would be
created. created.
Cultural Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Aesthetics Same as Option 1. Same as Option 1. Same as Option 1.
and Scenic
Resources
Geology Same as Option 1. Same as Option 1. Same as Option 1.
Air Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Water Same as Option 2a. Same as Option 2b. Very similar to
Resources Option 2c.
Ecological Same as Option 2a. Same as Option 2a. Same as Option 2c.
Resources
Traffic and Same as Option 2a. Same as Option 2a. Same as Option 2c.
Transportat
ion
Occupationa Maximum LCFe Maximum LCFe Maximum LCFe
l and probabilities: probabilities: probabilities:
Public Worker: 4 x 10-5 Worker: 5 x 10-5 Worker: 7 x 10-5
Health and Offsite population: Offsite population: Offsite population:
Safety 5 x 10-14 (air) 6 x 10-14 (air) 2 x 10-7 (air)
(Radiologic 2 x 10-14 (water) 2 x 10-14 (water) 6 x 10-8 (water)
al)
Annual LCFe Annual LCFe Annual LCFe
incidences: incidences: incidences:
Worker: 8 x 10-5 Worker: 1 x 10-4 Worker: 3 x 10-2
Offsite Offsite Offsite
population: 2 x 10-9 population: 2 x 10-9 population: 9 x 10-3
Occupationa Hazard index: Same as Option 4d. Hazard index:
l and Worker: 2 x 10-6 Worker: 8 x 10-3
Public Maximally exposed Maximally exposed
Health and individual: 3 x individual: 6 x
Safety 10-7 10-4
(Nonradiolo
gical)
Utilities Same as Option 2a. Very similar to Very similar to
and Energy Option 2a. Option 2a.
Materials Same as Option 1. Same as Option 1. Same as Option 4c.
and Waste
Management
Table 3-3. (continued).
Option 4d Option 4e Option 4f
Dry Storage Wet Storage Processing
Accidentsb Greatest point Same as Option 4d Greatest point
estimate estimate
of riskc: of riskc:
Worker: Data not Worker: Data not
calculatedd calculatedd
Colocated worker: Colocated worker:
2.0 x 10-6 1.2 x 10-6
Maximally exposed Maximally exposed
individual: individual: 2.5
4.1 x 10-7 x 10-7
Offsite population: Offsite population:
3.5 x 10-3 2.1 x 10-3
a. Impacts for Option 4g, Ship Offsite, would be the same as for Option
5d as described in the last entry in this table.
b. Data is provided as adjusted point estimates of risk by receptor group
to demonstrate a relative comparison of each alternative on an option-
by-option basis. The adjusted values were taken from Tables 5-27
through 5-29.
c. Units for adjusted point estimates of risk are given in terms of
potential fatal cancers per year.
d. The safety analysis reports from which information was extracted were
written before issuance of DOE Order 5480.23; previous orders did not
require the inclusion of workers.
e. LCF = latent cancer fatalities.
Table 3-3. (continued).
ALTERNATIVE 5 - CENTRALIZATION
Option 5a Option 5b Option 5c
Dry Storage Wet Storage Processing
Land Use Most new Same as Option 5a. Same as Option 5a.
construction would
be in parts of F-
and H-Areas already
dedicated to
industrial use.
Additional maximum
of 0.4 square
kilometer (100
acres) would be
converted from pine
plantation to
industrial use.
Impacts would be
minimal.
Socioeconom Operations jobs Operations jobs Operations jobs
ics would be filled by would be filled by would be filled by
present employees. present employees. present employees.
A maximum of about A maximum of about A maximum of about
2,550 construction 2,700 construction 2,550 construction
jobs would be jobs would be jobs would be
created. created. created.
Cultural No known historical, Same as Option 5a. Same as Option 5a.
Resources archeological, or
paleontological
resources are in
areas to be
affected. All areas
are classified as
having low or
moderate probability
of containing
archeological site.
Impact is unlikely.
Aesthetics Same as Option 1. Same as Option 1. Same as Option 1.
and Scenic
Resources
Geology Same as Option 1. Same as Option 1. Same as Option 1.
Air Same as Option 1. Same as Option 1. Same as Option 1.
Resources
Water Same as Option 2a. Same as Option 2b. Same as Option 2c.
Resources
Additional Additional Same as Option 5a.
groundwater groundwater
withdrawals would withdrawals would
total about 67.7 total about
million liters (17.9 69.6 million liters
million gallons) per (18.4 million
year. Impacts would gallons) per year. Same as Option 5a.
be minimal. Impacts would be
minimal.
No perennial streams
or other surface Same as Option 5a. Accidental releases
waters would be could contaminate
affected. shallow groundwater
that is not used as
Accidental releases Accidental releases a source for
could contaminate could contaminate drinking water or
shallow groundwater shallow groundwater domestic use.
that is not used as that is not used as Releases would not
a source for a source for affect surface
drinking water or drinking water or streams or drinking
domestic use. domestic use. water aquifers.
Releases would not Releases would not
affect surface affect surface
streams or drinking streams or drinking
water aquifers. water aquifers.
Table 3-3. (continued).
Option 5a Option 5b Option 5c
Dry Storage Wet Storage Processing
Ecological Same as Option 2a, Same as Option 5a. Same as Option 5a,
Resources plus plus
Loss of up to 0.4 Increased
square kilometer disturbance due to
(100 acres) of more worker
loblolly pine. traffic. Impacts
Impacts would be would be minor.
minor.
Noise Same as Option 2a. Same as Option 2a. Same as Option 2a.
Traffic and Same as Option 2a. This option would Same as Option 2c.
Transportati increase site
on traffic by about 17
percent. Impacts
would be small.
Number of LCFsg
would be same as
for Option 2b for
normal transport.
Occupational Maximum LCFg Maximum LCFg Maximum LCFg
and Public probabilities: probabilities: probabilities:
Health and Worker: 4 x 10-4 Worker: 5 x 10-4 Worker: 6 x 10-4
Safety Offsite Offsite Offsite
(Radiologica population: population: population:
l) 5 x 10-13 (air) 6 x 10-13 (air) 2 x 10-7 (air)
2 x 10-13 (water) 2 x 10-13 (water) 6 x 10-8 (water)
Annual LCFg Annual LCFg Annual LCFg
incidences: incidences: incidences:
Worker: 9 x 10-4 Worker: 1 x 10-3 Worker: 3 x 10-2
Offsite Offsite Offsite
population: 2 x 10-8 population: 3 x 10-8 population: 9 x 10-3
Occupational Same as Option 1. Same as Option 1. Same as Option 2c.
and Public
Health and
Safety
(Nonradiolog
ical)
Utilities Similar to Option Similar to Option Requirements for
and Energy 2a. 2a. electricity would
increase by about
17 percent. Other
increases would be
similar to Option
2c. Impacts would
be minor.
Materials Annual average Annual average Annual average
and Waste volume of waste volume of waste volume of waste
Management generated (cubic generated (cubic generated (cubic
meters)b: meters)b: meters)b:
LLW: 400 LLW: 400 LLW: 800
TRU: 16 TRU: 20 TRU: 20
HLW: 0 HLW: 2.3c HLW: 2.3c
No impact on site No impact on site No impact on site
capacities. capacities. capacities.
Table 3-3. (continued).
Option 5a Option 5b Option 5c
Dry Storage Wet Storage Processing
Accidentsd Greatest point Same as Option 5a. Greatest point
estimate of riske: estimate of riske:
Worker: Data not Worker: Data not
calculatedf calculatedf
Colocated worker: Colocated worker:
4.0 x 10-6 3.3 x 10-6
Maximally exposed Maximally exposed
individual: individual: 6.8 x
8.4 x 10-7 10-7
Offsite Offsite
population: 7.2 x population: 5.8 x
10-3 10-3
a. NA = not applicable.
b. LLW = low-level waste; TRU = transuranic waste; HLW = high-level
waste.
c. High-level waste will be generated only during approximately the first
10 years.
d. Data is provided as adjusted point estimates of risk by receptor group
to demonstrate a relative comparison of each alternative on an option-
by-option basis. The adjusted values were taken from Tables 5-27
through 5-29.
e. Units for adjusted point estimates of risk are given in terms of
potential fatal cancers per year.
f. The safety analysis reports from which information was extracted were
written before issuance of DOE Order 5480.23; previous orders did not
require the inclusion of workers.
g. LCF = latent cancer fatalities.
Table 3-3. (continued).
ALTERNATIVE 5 - CENTRALIZATION
ALTERNATIVE 4 - REGIONALIZATION B
Option 4g and Option 5db
Ship Out
Land Use Same as Option 1.
Socioeconomics No new operations jobs and only about 200 construction
jobs would be created.
Cultural Resources Same as Option 1.
Aesthetics and Same as Option 1.
Scenic Resources
Geology Same as Option 1.
Air Resources Same as Option 1.
Water Resources This option would require new withdrawals of
approximately 3.0 million liters (790 thousand gallons)
per year of cooling water from the Savannah River.
Impacts would be minimal.
It also would require additional groundwater withdrawals
of about 38.1 million liters (10.1 million gallons) per
year. Impacts would be minimal.
Impacts to surface water and groundwater would be
similar to those from Option 1.
Ecological Same as Option 1.
Resources
Noise Same as Option 2a.
Traffic and NAa
Transportation
Occupational and Less than Option 1.
Public Health and
Safety
(Radiological)
Occupational and Same as Option 1.
Public Health and
Safety
(Nonradiological)
Utilities and Requirements would increase 2 to 6 percent above current
Energy levels during first 10 years. Current SRS capacities
are adequate for these increases.
Materials and Waste Annual average volume of waste generated initial 10
Management years only (cubic meters)c:
LLW: 400
TRU: 18
HLW: 0
Table 3-3. (continued).
Option 4g and Option 5db
Ship Out
Accidentsd Greatest point estimate of riske:
Worker: Data not calculatedf
Colocated Worker:
Option 4g: 8.1 x 10-7
Option 5d: 8.2 x 10-7
Maximally exposed individual:
Option 4g: 1.7 x 10-7
Option 5d: 1.7 x 10-7
Offsite population:
Option 4g: 1.4 x 10-3
Option 5d: 1.4 x 10-3
a. NA = not applicable.
b. Impacts for Option 4g (Regionalization-B) are the same as for Option 5d.
c. LLW = low-level waste; TRU = transuranic waste; HLW = high-level waste.
d. Data is provided as adjusted point estimates of risk by receptor group
to demonstrate a relative comparison of each alternative on an option-
by-option basis. The adjusted values were taken from Tables 5-27
through 5-29.
e. Units for adjusted point estimates of risk are given in terms of
potential fatal cancers per year.
f. The safety analysis reports from which information was extracted were
written before issuance of DOE Order 5480.23; previous orders did not
require the inclusion of workers.
4. AFFECTED ENVIRONMENT
4.1 Overview
This section describes the existing environment at the Savannah River Site (SRS) and nearby
areas. Its purpose is to support the assessment of environmental consequences of the alternative
actions regarding spent nuclear fuels described in Chapter 3. Chapter 5 describes the environmental
consequences in detail.
4.2 Land Use
The SRS occupies an area of approximately 198,000 acres (800 square kilometers) in western
South Carolina, in a generally rural area about 25 miles (40 kilometers) southeast of Augusta, Georgia.
The SRS, which is bordered by the Savannah River to the southwest, includes portions of Aiken,
Barnwell, and Allendale Counties (Figure 2-1).
Land use on the SRS falls into three major categories: forest/undeveloped, water/wetlands, and
developed facilities. About 181,500 acres (735 square kilometers) of the SRS area are undeveloped
(USDA 1991a). Approximately 90 percent of this undeveloped area is forested (Cummins et al. 1991).
In 1952, an interagency agreement between the U.S. Department of Energy [DOE, which was then the
Atomic Energy Commission (AEC)] and the Forest Service, U.S. Department of Agriculture, created
an SRS forest management program. In 1972, the AEC designated the SRS as a National
Environmental Research Park (NERP); at present, approximately 14,000 acres (57 square kilometers or
7 percent) of the SRS area are designated as "Set-Asides," areas specifically protected for
environmental research activities that are coordinated either through the University of Georgia
Savannah River Ecology Laboratory (SREL) or the Savannah River Technology Center (SRTC; Davis
1994). Administrative, production, and support facilities occupy approximately 5 percent of the total
SRS land area.
DOE is considering decisions that could affect the long-range land use of the SRS.
Programmatic decisions on the reconfiguration of the nuclear weapons complex, spent nuclear fuel
interim strategies, and waste management and environmental restoration activities that could result in
significant changes in the SRS mission are in the early stages of discussion. In the shorter term,
however, a Land Use Technical Committee consisting of representatives from DOE, Westinghouse
Savannah River Company, and various stakeholder groups is evaluating alternative land use strategies
and potential future uses. These activities are consistent with the guidelines for land use plans
contained in DOE Order 4320.1B, "Site Development Planning," and in the Resource Conservation
and Recovery Act (RCRA) and the Comprehensive Environmental Response, Compensation, and
Liability Act (CERCLA).
Land use bordering SRS is primarily forest and agricultural. There is also a significant amount
of open water and nonforested wetlands along the Savannah River valley. Incorporated and industrial
areas are the only other significant use of land in the vicinity (Figure 4-1). None of the three counties
in which the SRS is located has zoned any of the Site land. The only adjacent area with any zoning is
the Town of New Ellenton, which has two zoning categories for lands that bound SRS - urban
development and residential development. The closest residences to the SRS boundary include several
within 200 feet (61 meters) of the Site perimeter to the west, north, and northeast.
Various industrial, manufacturing, medical, and farming operations are conducted in areas
surrounding the Site. Major industrial and manufacturing facilities in the area include textile mills,
plants producing polystyrene foam and paper products, chemical processing plants, and a commercial
nuclear power plant. Farming is diversified in the region and includes crops such as peaches,
watermelon, cotton, soybeans, corn, and small grains.
There is a wide variety of public outdoor recreation facilities in the SRS region (Figure 4-2).
Federal outdoor recreation facilities include portions of the Sumter National Forest [47 miles
(75 kilometers) to the northwest of the Site], the Santee National Wildlife Refuge [50 miles
(80 kilometers) to the east], and the Clarks Hill/Strom Thurmond Reservoir, a U.S. Army Corps of
Engineers impoundment [43 miles (70 kilometers) to the northwest]. There are also a number of state,
county, and local parks in the region, most notably Redcliffe Plantation, Rivers Bridge, Barnwell and
Aiken County State Parks in South Carolina, and Mistletoe State Park in Georgia (HNUS 1992a).
The SRS is a controlled area with public access limited to through traffic on South Carolina
Highway 125 (SRS Road A), U.S. Highway 278, SRS Road 1, and the CSX railway. The SRS does
not contain any public recreation facilities. However, the SRS conducts controlled deer hunts each
fall, from mid-October through mid-December; hunters can also kill feral hogs during these hunts.
Figure 4-1. Generalized land use at the Savannah River Site and vicinity. Figure 4-2. Federal and state forests and parks within a 2-hour drive from Savannah River Site. The intent of the hunts is to control the resident populations of these animals and to reduce
animal-vehicle accidents on SRS roads.
No onsite areas are subject to Native American treaty rights. The SRS does not contain any
prime farmland.
4.3 Socioeconomics
This section discusses baseline socioeconomic conditions within a region of influence where
approximately 90 percent of the SRS workforce lived in 1992. The SRS region of influence includes
Aiken, Allendale, Bamberg, and Barnwell Counties in South Carolina, and Columbia and Richmond
Counties in Georgia (Figure 4-2).
4.3.1 Employment and Labor Force
The labor force living in the region of influence increased from about 150,550 to 209,000
between 1980 and 1990. In 1990, approximately 75 percent of the total labor force in the region of
influence lived in Richmond and Aiken Counties. Assuming a constant unemployment rate of 5.8
percent, the regional labor force is likely to increase to approximately 257,000 by 1995 (Table 4-1).
Between 1980 and 1990, total employment in the region of influence increased from 139,504 to
199,161, an average annual growth rate of approximately 5 percent. Table 4-1 lists projected
employment data for the six-county region of influence. As shown, by 1995 employment levels
should increase 22 percent to approximately 242,000. The unemployment rates for 1980 and 1990
were 7.3 percent and 4.7 percent, respectively (HNUS 1992a).
In 1990, employment at the SRS was 20,230 (DOE 1993a), representing 10 percent of the
employment in the region of influence. In Fiscal Year 1992, employment at the SRS increased
approximately 15 percent to 23,351, with an associated payroll of more than $1.1 billion. Due to
planned budget reductions, Site employment could decline by as many as 4,200 jobs (Fiori 1995). As
shown in Table 4-1, this would reduce Site employment to approximately 15,800 by 1996.
Table 4-1. Forecast employment and population data for the Savannah River Site and the region of
influence.
Labor Force Employment Population
Year (Region) (Region) SRS Employmentb (Region)
1994 254,549 239,785 21,500 456,892
1995 256,935 242,033 20,000 461,705
1996 258,500 243,507 15,800 465,563
1997 260,680 245,561 15,800 468,665
1998 263,121 247,860 15,800 471,176
1999 265,694 250,284 15,800 473,186
2000 268,430 252,861 15,800 474,820
2001 271,265 255,532 15,800 476,179
2002 274,238 258,332 15,800 477,332
2003 277,318 261,234 15,800 478,340
2004 280,415 264,151 15,800 479,182
a. Source: HNUS (1993).
b. Sources: Turner (1994), Fiori (1995).
4.3.2 Personal Income
Personal income in the six-county region has doubled during the past two decades, increasing from
approximately $3.4 billion in 1970 to almost $6.9 billion by 1989 (in constant 1991 dollars).
Together, Richmond and Aiken Counties accounted for 75.4 percent of the personal income in the
region of influence in 1989, because these two counties provide most of the employment opportunities
in the region. Personal income in the region is likely to increase 3 percent to approximately
$7.1 billion by 1995 and to almost $8.2 billion by 2000 (HNUS 1992a).
4.3.3 Population
Between 1980 and 1990, the population in the region of influence increased 13 percent from
376,058 to 425,607. More than 88 percent of the 1990 population lived in Aiken (28.4 percent),
Columbia (15.5 percent), and Richmond (44.6 percent) Counties. Table 4-1 also lists population data
for the region of influence forecast to 2004. According to census data, in 1990 the estimated average
number of persons per household in the six-county region was 2.72, and the median age of the
population was 31.2 years (HNUS 1992a).
4.3.4 Housing
From 1980 to 1990, the number of year-round housing units in the six-county region increased
23.2 percent from 135,866 to 167,356. In 1990, approximately 68 percent of the total housing units
were single-family units, 18 percent were multifamily units, and 14 percent were mobile homes. In
the same year, the region had a 4.7-percent vacancy rate with 7,818 available unoccupied housing
units. Of the available unoccupied units, 29 percent (2,267) were available for sale and 71 percent
(5,551) were available for rent (HNUS 1992a).
4.3.5 Community Infrastructure and Services
Public education facilities in the six-county region include 95 elementary and intermediate
schools and 25 high schools. Aside from the public school systems, 42 private schools and 16 post-
secondary facilities are available to residents in the region (HNUS 1992a).
Based on a combined average daily attendance for elementary and high school students in the
region of influence in 1988, the average number of students per teacher was 16. The highest ratio was
in Columbia County high schools where there were 19 students per teacher (1987-1988). The lowest
ratio occurred in Barnwell County's District 29 high school, which had only 12 students per teacher
(1988-1989) (HNUS 1992a).
The six-county region has 14 major public sewage treatment facilities with a combined design
capacity of 302.2 million liters (79.8 million gallons) per day. In 1989, these systems were operating
at approximately 56 percent of capacity, with an average daily flow of 170 million liters (44.9 million
gallons) per day. Capacity utilization ranged from 45 percent in Aiken County to 80 percent in
Barnwell County (HNUS 1992a).
There are approximately 120 public water systems in the region of influence. About 40 of these
county and municipal systems are major facilities, while the remainder serve individual subdivisions,
water districts, trailer parks, and miscellaneous facilities. In 1989, the 40 major facilities had a
combined total capacity of 576.3 million liters (152.2 million gallons) per day. With an average daily
flow rate of approximately 268.8 million liters (71 million gallons) per day, these systems were
operating at 47 percent of total capacity in 1989. Facility utilization rates ranged from 13 percent in
Allendale County to 84 percent in the City of Aiken (HNUS 1992a).
Eight general hospitals operate in the six-county region with a combined bed capacity in 1987 of
2,433 (5.7 beds per 1,000 population). Four of the eight general hospitals are in Richmond County;
Aiken, Allendale, Bamberg, and Barnwell Counties each have one general hospital. Columbia County
has no hospital. In 1989, there were approximately 1,295 physicians serving the regional population,
which represents a physician-to-population ratio of 3 to 1,000. This ratio ranged from 0.8 physician
per 1,000 people in Aiken and Allendale Counties to 5.4 physicians per 1,000 people in Richmond
County (HNUS 1992a).
Fifty-six fire departments provide fire protection services in the region of influence. Twenty-
seven of these are classified as municipal fire departments, but many provide protection to rural areas
outside municipal limits. The average number of firefighters in the region in 1988 was 3.8 per
1,000 people, ranging from 1.6 per 1,000 in Richmond County to 10.2 per 1,000 in Barnwell County
(HNUS 1992a).
The county sheriff departments and municipal police departments provide most law enforcement
services in the region of influence. In addition, state law enforcement agents and state troopers
assigned to each county provide protection and assist county and municipal law enforcement officers.
In 1988, the average ratio in the region of full-time police officers employed by state, county, and
local agencies per 1,000 population was 2.0. This ratio ranged from 1.4 per 1,000 in Columbia
County to 2.5 per 1,000 in Richmond County (HNUS 1992a).
4.3.6 Government Fiscal Structure
This section discusses the fiscal structure of Aiken and Barnwell Counties because these two
counties would have the greatest potential for fiscal impacts from changes at SRS.
Public services provided by Aiken County are funded principally through the county's general
fund. In Fiscal Year 1988, revenues and expenditures of this fund were $15.5 million and
$18 million, respectively. The current property tax rate is 55.8 mills for county operations and
8.0 mills for debt service. Long-term general obligation bond indebtedness was $9.3 million at the
end of Fiscal Year 1988, and reserve general obligation bond indebtedness was $5.5 million. The
assessed value of property in the county was $182.5 million in Fiscal Year 1988 (HNUS 1992a).
Assuming revenues and expenditures increase in proportion to projected growth in the
employment and population, estimated revenues and expenditures for Aiken County over the period
from Fiscal Year 1990 to Fiscal Year 2000 will be $15.6 million to $17.0 million (in constant 1988
dollars) (HNUS 1992a).
Public services provided by Barnwell County also are funded principally through the county's
general fund. In Fiscal Year 1988, revenues and expenditures of this fund were $4.0 million and
$4.9 million, respectively. The property tax rate is 23.9 mills of assessed valuation. Budgeted Fiscal
Year 1990 revenues were approximately $4.5 million (HNUS 1992a).
4.4 Cultural Resources
4.4.1 Archeological Sites and Historic Structures
Field studies conducted under an ongoing program over the past two decades by the South
Carolina Institute of Archeology of the University of South Carolina, under contract to DOE and in
consultation with the South Carolina State Historic Preservation Officer, have provided considerable
information about the distribution and content of archeological and historic resources on the SRS. By
the end of Fiscal Year 1992, approximately 60 percent of the Site had been examined, and 858
archeological (historic and prehistoric) sites had been identified; these include 706 prehistoric and
350 historic components, some of which are mixed (i.e., contain elements of both). Of the 858 sites,
53 have been determined to be eligible for the National Register of Historic Places; 650 have not been
evaluated. Approximately 21 of the 53 (40 percent) are historic sites, such as building foundations;
none are standing structures. These sites provide knowledge of the area's history before 1820. The
remainder are primarily prehistoric sites and some are mixed (historic and prehistoric). No SRS
facilities have been nominated for eligibility to the National Register for Historic Places and there are
no plans for such a nomination at this time (Brooks 1993; Brooks 1994). The existing SRS nuclear
production facilities are not likely to be eligible for the National Register, either because they might
lack architectural integrity, might not represent a particular architectural style, or might not contribute
to the broad historic theme of the Manhattan Project and initial nuclear materials production
(DOE 1993a).
Archeologists have divided areas of the SRS into three sensitivity zones related to their potential
for containing sites with multiple archeological components or dense or diverse artifacts, and their
potential for eligibility to the National Register of Historic Places (SRARP 1989).
- Zone 1 is the zone of the highest archeological site density with a high probability of
encountering large archeological sites with dense and diverse artifacts, and high potential for
nomination to the National Register of Historic Places.
- Zone 2 covers areas of moderate archeological site density that should contain sites of
similar composition. Activities in this zone have a moderate probability of encountering
archeological sites, but a low probability of encountering large sites with more than three
prehistoric components. All areas within the zone are conducive to site preservation. The
zone has moderate potential for encountering sites that would be eligible for nomination to
the National Register of Historic Places.
- Zone 3 covers areas of low archeological site density. Activities in this zone have a low
probability of encountering archeological sites and virtually no chance of encountering large
sites with more than three prehistoric components; potential for site preservation is low.
Some exceptions to this definition have been discovered in Zone 3, so some sites in the
zone could be considered eligible for nomination to the National Register of Historic Places.
4.4.2 Native American Cultural Resources
In conjunction with 1991 studies related to a proposed New Production Reactor, DOE conducted
an investigation of Native American concerns over religious rights in the Central Savannah River
Valley. During this study three Native American groups - the Yuchi Tribal Organization, the National
Council of Muskogee Creek, and the Indian People's Muskogee Tribal Town Confederacy - expressed
concerns over sites and items of religious significance on the SRS. DOE has included these
organizations on its environmental mailing list and sends them documents about SRS environmental
activities (NUS 1991a).
Native American resources in the region include villages or townsites, ceremonial lodges, burial
sites, cemeteries, and areas containing traditional plants for certain rituals. Villages or townsites might
contain a variety of sensitive features associated with different ceremonies and rituals. The Yuchi and
Muskogee Creek tribes have expressed concerns that the area might contain several plants traditionally
used in tribal ceremonies (DOE 1993a).
4.4.3 Paleontological Resources
Invertebrate fossil remains occur within the McBean, Barnwell, and Congaree formations of the
Eocene Age (54 million to 39 million years ago) on the SRS. Relatively large quantities of marine
invertebrate fossils have been recorded for the McBean and Barnwell Formations. Relative assessment
of fossil localities is difficult because the South Carolina Geological Survey has not established criteria
for, or registry of, important paleontological locations (DOE 1991b).
4.5 Aesthetics and Scenic Resources
The dominant aesthetic setting in the vicinity of the SRS consists mainly of agricultural land and
forest, with some limited residential and industrial areas. Because of the distance to the Site boundary,
the rolling terrain, normally hazy atmospheric conditions, and heavy vegetation, SRS facilities are not
generally visible from off the Site. The few locations that have views of some of the SRS structures
are quite distant from the facility [5 miles (8 kilometers) or more].
SRS land is heavily wooded, and developed areas occupy only approximately 5 percent of the
total land area. The facilities are scattered across the SRS and are brightly lit at night. Typically, the
reactors and principal processing facilities are large concrete structures as much as 100 feet
(30 meters) high and usually colocated with lower administrative and support buildings and parking
lots. The facilities are visible in the direct line-of-sight when approaching them from SRS access
roads. A 500-foot cooling tower is located in K-Area. Otherwise, heavily wooded areas that border
the SRS road system and public highways that cross the Site limit views of the facilities.
4.6 Geology
The SRS is on the Upper Atlantic Coastal Plain of South Carolina, which consists of 213 to
366 meters (700 to 1,200 feet) of sands, clays, and limestones of Tertiary and Cretaceous age. These
sediments are underlain by sandstones of Triassic age and older metamorphic and igneous rocks
(Arnett et al. 1993). There are no known capable faults on the SRS or volcanic activities within
800 kilometers (500 miles) of the Site.
4.6.1 General Geology
The SRS is in the Coastal Plain physiographic province of western South Carolina,
approximately 32 kilometers (20 miles) southeast of the Fall Line, which separates the Piedmont and
Coastal Plain provinces (Figure 4-3). The Coastal Plain province is underlain by a wedge of
seaward-dipping and thickening unconsolidated and semiconsolidated sediments that extend from the
Fall Line to the Continental Shelf (Figure 4-4).
In South Carolina, the Coastal Plain province is divided into the Upper Coastal Plain and the
Lower Coastal Plain. Subdivisions of the Coastal Plain in the State include the Aiken Plateau and the
Congaree Sand Hills in the Upper Coastal Plain, and the Coastal Terraces in the Lower Coastal Plain.
The Congaree Sand Hills trend along the Fall Line northeast and north of the Aiken Plateau. The
Savannah and Congaree Rivers bound the Aiken Plateau, on which the SRS is located; the plateau
extends from the Fall Line to the Coastal Terraces. The surface of the plateau is highly dissected and
characterized by broad interfluvial areas with narrow steep-sided valleys. The plateau is generally well
drained, although poorly drained depressions (Carolina bays) do exist (DOE 1991b). Because of the
proximity of the SRS to the Piedmont province, it has more relief than areas that are nearer to the
coast, with onsite elevations ranging from 27 to 128 meters (89 to 420 feet) above mean sea level.
The sediments of the Atlantic Coastal Plain of South Carolina overlie a basement complex
composed of Paleozoic crystalline and Triassic sedimentary rocks. These sediments dip gently
seaward from the Fall Line and range in age from Late Cretaceous to Recent. The sedimentary
sequence thickens from essentially zero at the Fall Line to more than 1,219 meters (4,000 feet) at the
coast. Regional dip is to the southeast. Coastal Plain sediments underlying the SRS consist of sandy
clays and clayey sands, although occasional beds of clean sand, gravel, clay, or carbonate occur
(Figure 4-5). Two clastic limestone zones occur within the Tertiary age sequence. These calcareous
zones vary in thickness from about 0.6 meter (2 feet) to approximately 24 meters (80 feet). Most of
the clastic sediments are unconsolidated, but thin semiconsolidated beds also occur (DOE 1991b).
Underlying sediments are dense crystalline igneous and metamorphic rock or younger consolidated
sediments of the Triassic Period. The Triassic formations and older igneous and metamorphic rocks
are separated hydrologically from the overlying Coastal Plain sediments by a regional aquitard, the
Appleton Confining System (Arnett et al. 1993). Section 4.8.2 contains a detailed discussion of
hydrogeology on the SRS.
SRS construction activities have used clay, sand, and gravel to a limited extent. These materials
are not of major economic value due to their abundance throughout the region. The SRS historically
has been a major user of groundwater in the region, withdrawing about 33 million liters (9 million
gallons) per day. Section 4.8.2 describes the groundwater resources at the SRS.
4.6.3 Seismic and Volcanic Hazards
The closest offsite fault system of significance is the Augusta Fault Zone, approximately
40 kilometers (25 miles) from the SRS. In this fault zone, the Belair Fault has experienced the most
recent movement, but it is not considered capable of generating major earthquakes (DOE 1987a).
There is no conclusive evidence of recent displacement along any fault within 320 kilometers (200
miles) of the SRS, with the possible exception of the buried faults in the epicentral area of the 1886
earthquake at Charleston, South Carolina, approximately 145 kilometers (90 miles) away (DOE
1991b). Faulting in the subsurface Coastal Plain sediments in the Charleston vicinity has been
suggested, based on structure contour mapping of the Eocene-Oligocene unconformity, which lies at a
depth of about 30 to 61 meters (100 to 200 feet) below ground surface (WSRC 1994a). However,
because it is not known if these faults offset sediments younger than Eocene-Oligocene, these shallow
faults cannot be related to modern earthquakes that occur at depths greater than about 1.9 kilometers
(1.2 miles). Figure 4-6 shows the geologic structures within 150 kilometers (95 miles) from the SRS,
some of which are discussed above.
Several Triassic-Jurassic basins, 140 to 230 million years old, have been identified in the Coastal
Plain province of South Carolina and Georgia. The Dunbarton Triassic basin, which underlies a
portion of the SRS, was formed by fault movement resulting from extensional forces operating during
the formation of the Atlantic Ocean. After the erosion of basin margins and infilling of the basin with
Triassic age sediments, possible movement of an opposite sense to that during basin formation
occurred along the fault during the Late Cretaceous age. Geophysical data indicate minimal movement
on faults at the basement-Coastal Plain interface, with the exception of possible reverse fault motion
along the Pen Branch Fault up into the Tertiary (WSRC 1994a).
Figure 4-6. Geologic structures within 150 km of SRS (Source: DOE 1991b). Researchers have mapped the Pen Branch Fault for at least 24 kilometers (15 miles) across the
central portion of the SRS (Snipes et al. 1993). This fault is probably a continuation of the northern
boundary fault of the Triassic age Dunbarton basin and is interpreted as being at least a
Cretaceous/Tertiary (144-1.6 million years) reactivation of that fault (WSRC 1994a). Observed
displacements of the Coastal Plain sediments range from about 26 meters (85 feet) at the
Basement/Cretaceous contact to about 9 meters (30 feet) in the shallower sediments (WSRC 1994a).
Based on the available data, there is no evidence to indicate that the Pen Branch is a "capable fault" as
defined by the U.S. Nuclear Regulatory Commission (NRC). Under the NRC definition, a fault is
capable if it has moved within the last 35,000 years, has had recurring movement within the last
500,000 years, is related to any earthquake activity, or is associated with another capable fault. A
recent study (Snipes et al. 1993) examined a Quaternary light tan soil horizon in SRS railroad cuts.
The soil horizon, which has a thickness of 3 to 6 meters (10 to 20 feet), revealed no detectable offset,
indicating that there has been no recent Pen Branch Fault activity. Figure 4-7 shows the locations of
the Pen Branch Fault and other known or suspected faults within the Paleozoic and Triassic Basement
(DOE 1991b).
Seismicity in the Coastal Plain of South Carolina occurs in three distinct seismic zones near the
Charleston area (WSRC 1994a): Middleton Place-Summerville, about 19 kilometers (12 miles)
northwest of Charleston; Bowman, about 59 kilometers (37 miles) northwest of the Middleton
Place-Summerville; and Adams Run, about 30 kilometers (19 miles) southwest of the Middleton
Place-Summerville (WSRC 1994a). Of the distinct seismic zones within the Coastal Plain province,
the Charleston area has been and remains the most seismically active. The Charleston area is also the
most significant source of seismicity affecting the SRS, both in terms of maximum historic site
intensity and the number of earthquakes felt in the area (WSRC 1994a).
Tables 4-2 and 4-3 summarize the historic information on earthquakes that have occurred in the
SRS region. Two notable earthquakes have occurred within 320 kilometers (200 miles) of the SRS.
The first was a major earthquake in 1886 centered in the Charleston area about 145 kilometers
(90 miles) from the Site; it had an estimated Richter magnitude of 6.8. DOE estimates that the SRS
would have felt a tremor with an estimated Modified Mercalli Intensity (MMI) of VI to VII and an
estimated peak horizontal acceleration of 10 percent of gravity, or 0.10g, due to that earthquake
(WSRC 1994a). The second earthquake was the Union County, South Carolina, earthquake of 1913,
which had an estimated Richter magnitude of 6.0 and occurred about 160 kilometers (100 miles) from
the SRS (WSRC 1994a). This earthquake, which is the closest significant event to the SRS other than
Figure 4-7. Geologic faults of the Savannah River Site. Table 4-2. Earthquakes in the SRS region with a Modified Mercalli Intensity greater than V.
Reported or
Coordinates Estimated Estimated
Maximum Distance from Intensity at Richter Acceleration
Dateb Location Intensity SRS (km)c SRS Magnitude at SRS(g)
Lat. Long.
(yN) (yW)
1811 Jan 13 Burke Co., Ga. 33.2 82.2 V 55 III-IV NAd 0.02
1811-1812 New Madrid, Mo. 36.3 89.5 XI-XII 850 V-VI NA 0.05
(3 shocks)
1875 Nov 02 Lincolnton, Ga. 33.8 82.5 VI 100 III-IV NA 0.02
1886 Sep 02 Charleston, S.C. 32.9 80.0 X 145 VI 6.8 0.10
1886 Oct 22 Charleston, S.C. 32.9 80.0 VII 155 III-IV NA 0.02
1897 May 31 Giles Co., Va. 33.0 80.7 VIII 455 III NA 0.02
1913 Jan 01 Union Co., S.C. 34.7 81.7 VII-VIII 160 IV 6.0e 0.02
1920 Aug 01 Charleston, S.C. 33.1 80.2 VII 135 III-IV NA 0.02
1972 Feb 03 Bowman, S.C. 33.5 80.4 V 115 IV 4.5 0.02
1974 Aug 02 Willington, S.C. 33.9 82.5 VI 105 IV 4.1 0.02
1974 Nov 22 Charleston, S.C. 32.9 80.1 VI 145 III-IV 4.3 0.02
a. Source: DOE (1991b).
b. Based on Greenwich mean time.
c. Conversion factor: 1 kilometer = 0.6214 mile.
d. NA = data not available.
e. Estimated.
Table 4-3. Earthquakes in the SRS region with a Modified Mercalli Intensity greater than IV or a magnitude greater than 2.0.
Reported or
Coordinates Estimated Estimated
Maximum Distance from Intensity at Richter Acceleration
Dateb Intensity SRS (km)c SRS Magnitude at SRS(g)
Lat. Long.
(yN) (yW)
1811 Jan 13d 33.2 82.2 V 55 III-IV NAe 0.02
1853 May 20 34.0 81.2 VI 102 NA NA NA
1945 Jul 26 33.8 81.4 V 77 NA 4.4 NA
1964 Mar 07 33.7 82.4 NA 85 NA 3.3 NA
1964 Apr 20 33.8 81.1 V 96 NA 3.5 NA
1968 Sep 22 34.1 81.5 IV 102 NA 3.5 NA
1972 Aug 14 33.2 81.4 NA 27 NA 3.0 NA
1974 Oct 28 33.8 81.9 IV 72 NA 3.0 NA
1974 Nov 05 33.7 82.2 III 77 NA 3.7 NA
1976 Sep 15 33.1 81.4 NA 25 NA 2.5 NA
1977 Jun 05 3.1 81.4 NA 35 NA 2.7 NA
1982 Jan 28 32.9 81.4 NA 40 NA 3.4 NA
1985 Jun 08 33.2 81.7 III Onsite III 2.6 NA
1988 Feb 17f 33.6 81.7 III 45 NA 2.6 NA
1988 Aug 05 33.1 81.4 NA Onsite II 2.0 NA
1993 Aug 08 NA NA NA NA NA 3.2 NA
a. Source: DOE (1991b).
b. Based on Greenwich mean time.
c. Conversion factor: 1 kilometer = 0.6214 mile.
d. Located in Burke County, Ga.
e. NA = data not available.
f. Located at Aiken, S.C.
the Charleston-area earthquake, produced an estimated intensity of II to III (MMI) in the City of
Aiken, which is approximately 19 kilometers (12 miles) north of the Site (DOE 1991b; WSRC 1994a).
Two earthquakes have occurred on the SRS during recent years (see Figure 4-7). On June 8,
1985, onsite instruments recorded an earthquake with a Richter magnitude of 2.6 and a focal depth of
about 1.0 kilometer (0.6 mile) (WSRC 1994a). The epicenter was just west of the C- and K-Areas.
The ground acceleration from this event did not activate instrumentation in the reactor areas (detection
limits of 0.002g). On August 5, 1988, an earthquake with a Richter magnitude of 2.0 and a focal
depth of approximately 2.7 kilometers (1.7 miles) occurred (Stephenson 1988); earthquakes of Richter
magnitude 2.0 are normally detected only by specialized instrumentation. The epicenter for this event
was just northeast of K-Area. Although this event was not felt by workers on the SRS, it was
recorded by sensors within 96 kilometers (60 miles) of the Site. A report on the August 1988
earthquake (Stephenson 1988) also reviewed the latest earthquake history for the region. This report
predicts recurrence period of 1 year for a magnitude 2.0 event for the southeast Coastal Plain.
However, the report notes that historic data to calculate recurrence rates accurately are sparse. SRS
workers did feel the effects of two other events that occurred in the area within the past 7 years. A
Richter magnitude 2.6 earthquake occurred in the City of Aiken, approximately 19 kilometers
(12 miles) north of the SRS on February 17, 1988. Reports indicate that this event was felt in the
Aiken area and on the SRS (DOE 1991b). Most recently, a Richter magnitude 3.2 earthquake
occurred on August 8, 1993, approximately 16 kilometers (10 miles) east of the City of Aiken near
Couchton, South Carolina. Residents reported feeling this earthquake in Aiken, New Ellenton
(immediately north of the SRS), North Augusta (approximately 40 kilometers [25 miles] northwest of
the SRS), and the Site.
Based on seismic activity information in the past 300 years, this analysis does not project
earthquakes greater than a Richter magnitude 6.0, which corresponds to a Modified Mercalli Intensity
of VII, to occur on the SRS. The design-basis earthquake for the SRS is a Modified Mercalli
Intensity VIII event, which corresponds to a horizontal peak ground acceleration of 0.2g. Based on
current technology, as applied in various probabilistic evaluations of the seismic hazard in the SRS
region, the 0.2g peak ground acceleration can be associated with a 2 x 10-4 annual probability of
exceedance (5,000-year return period). DOE Standards 1020 (DOE 1994a) and 1024 (DOE 1992)
summarize the results of recent seismic analyses at DOE sites and show that maximum horizontal
ground accelerations for the Savannah River Site for 500 year, 1,000 year, 2,000 year, and 5,000 year
seismic events are 0.10g, 0.13g, 0.18g, and 0.19g respectively. The seismic hazard information
presented in this EIS is for general seismic hazard comparisons across DOE sites. Potential seismic
hazards for existing and new facilities should be evaluated on a facility-specific basis consistent with
DOE Orders and standards and site-specific standards.
Historically, DOE has generally selected the more conservative 0.20g as the peak ground
acceleration for the 5,000 year seismic event when preparing safety analysis reports and environmental
impact statements for the SRS. For consistency with these existing analyses, this environmental
impact statement assumes 0.20g to be the peak horizontal ground acceleration that would result from
the 5,000 year seismic event. Figure 4-8 shows seismic hazard curves for the SRS.
A number of paleoliquefaction sites have been identified in Beaufort County, South Carolina,
some 50 miles (80 kilometers) southeast of the SRS, indicating a likelihood of prehistoric seismic
events outside of the currently-active Charleston seismic zone (Rajendran and Talwani 1993). There is
no evidence to suggest that seismically-induced liquefaction of soils represents a hazard at SRS,
however. Weak subsurface zones are encountered occasionally during drilling. These zones are
associated with carbonate materials and appear to be related to dissolution of these materials.
Engineering investigations have been conducted on granular soils underlying the Defense Waste
Processing Facility [in S-Area just north of H-Area (see Figure 2-3)] to evaluate the cyclic mobility
(liquefaction under cyclic stresses) of these soils (WSRC 1992b). These investigations determined that
the sands and clayey sands throughout the subgrade will not experience liquefaction (strength loss
leading to bearing capacity failures) and will not develop cyclic mobility (significant cyclic or
accumulate deformations) under the safe shutdown earthquake with a peak horizontal ground surface
acceleration of 0.20g (9.8 meters/second2 or 32.1 feet/second2).
4.7 Air Resources
4.7.1 Meteorology and Climatology
The SRS collects wind data from instruments mounted on seven onsite 61-meter (200-foot)
meteorological towers. Figure 4-9 shows a wind rose that represents annual wind direction frequencies
and wind speeds for the SRS from 1987 through 1991. The maximum wind directional frequencies
are from the northeast and west-southwest. The average wind speed for this 5-year period was
3.8 meters per second (8.5 miles per hour). Calm winds (less than 1 meter per second or 2.2 miles
per hour) occurred less than 10 percent of the time during the 5-year period. Seasonally, wind speeds
Figure 4-8. Seismic hazard curves for the SRS. Figure 4-9. Wind Rose. were greatest during the winter at 4.1 meters per second (9.5 miles per hour) and lowest during the
summer at 3.4 meters per second (7.6 miles per hour) (WSRC 1994a).
The annual average temperature at the SRS is 18 degrees C (64 degrees F); monthly averages
range from a low of 7 degrees C (45 degrees F) in January to a high of 27 degrees C (81 degrees F)
in July. Relative humidity readings taken four times each day range from 36 percent in April to
98 percent in August (DOE 1991a).
The average annual precipitation at the SRS is approximately 122 centimeters (48 inches).
Precipitation distribution is fairly even throughout the year, with the highest precipitation in the
summer [36.1 centimeters (14.2 inches)] and the lowest in autumn [22.4 centimeters (8.8 inches)].
Snowfall has occurred in the months of October through March, with the average annual snowfall at
3.0 centimeters (1.2 inches). Large snowfalls are rare (DOE 1991a).
Winter storms in the SRS area occasionally bring strong and gusty surface winds with speeds as
high as 32 meters per second (72 miles per hour). Thunderstorms can generate winds with speeds as
high as 18 meters per second (40 miles per hour) and even stronger gusts. The fastest 1-minute wind
speed recorded at Augusta between 1950 and 1986 was 37 meters per second (83 miles per hour)
(DOE 1991a).
4.7.1.1 Occurrence of Violent Weather. The SRS area experiences an average of 56
thunderstorm days per year. From 1954 to 1983, 37 tornadoes were reported for a 1-degree square of
latitude and longitude that includes the SRS (DOE 1991a). This frequency of occurrence is equivalent
to an average of about one tornado per year. The estimated probability of a tornado striking a point
on the SRS is 7 x 10-5 per year (DOE 1991a). Since operations began at the SRS in 1953, nine
confirmed tornadoes have occurred on or near the Site. They caused nothing more than light damage,
with the exception of a tornado in October 1989 that caused considerable damage to forest resources in
an undeveloped southeastern sector of the SRS (WSRC 1994a).
From 1700 to 1992, 36 hurricanes occurred in South Carolina, resulting in an average frequency
of about one hurricane every 8 years. Three hurricanes were classified as major. Because SRS is
about 160 kilometers (100 miles) inland, the winds associated with hurricanes have usually diminished
below hurricane force [i.e., equal to or greater than a sustained wind speed of 33.5 meters per second
(75 miles per hour)] before reaching the SRS. Winds exceeding hurricane force have been observed
only once at SRS (Hurricane Gracie in 1959) (WSRC 1994a).
4.7.1.2 Atmospheric Stability. Based on measurements at onsite meteorological stations, the
atmosphere in the SRS region is unstable approximately 56 percent of the time, neutral 23 percent of
the time, and stable about 21 percent of the time. On an annual basis, inversion conditions occur
21 percent of the time at the SRS (WSRC 1994a).
4.7.2 Nonradiological Air Quality
4.7.2.1 Background Air Quality. The SRS is in the Augusta (Georgia) - Aiken (South
Carolina) Interstate Air Quality Control Region (AQCR). This Air Quality Control Region, which is
designated as a Class II area, is in compliance with National Ambient Air Quality Standards (NAAQS)
for criteria pollutants. The criteria pollutants include sulfur dioxide, nitrogen oxides reported as
nitrogen dioxide, particulate matter (less than or equal to 10 microns), carbon monoxide, ozone, and
lead (CFR 1993a). The closest nonattainment area to the SRS is the Atlanta, Georgia, air quality
region, 233 kilometers (145 miles) to the west, which is in nonattainment of the standard for ozone.
The SRS will have to comply with Prevention of Significant Deterioration (PSD) Class II
requirements if there is a significant increase in emissions of criteria air pollutants due to a
modification at the Site (CFR 1993b). Development at the SRS has not yet triggered Prevention of
Significant Deterioration permitting requirements. If a permit were required, the SRS would have to
address several requirements, including impacts on the air quality of Class I areas within 10 kilometers
(6.2 miles) of the Site (CFR 1993b). The nearest Class I area to the SRS is the Congaree Swamp
National Monument in South Carolina, approximately 73 kilometers (45 miles) to the east-northeast of
the Site. Therefore, a Prevention of Significant Deterioration permit, if required for the SRS, would
not have to address Class I areas.
4.7.2.2 Air Pollutant Source Emissions. The SRS utilized the 1990 comprehensive
emissions inventory data to establish the baseline year for showing compliance with State and Federal
air quality standards - calculating both maximum potential and actual emission rates. The air quality
compliance demonstration also included sources forecast for construction or operation in this decade
(for which the SRS had obtained air quality construction permits through December 1992). The SRS
based its calculated emission rates for the sources on process knowledge, source testing, permitted
operating capacity, material balance, and U.S. Environmental Protection Agency (EPA) Air Pollution
Emission Factors (AP-42; EPA 1985).
4.7.2.3 Ambient Air Monitoring. At present, the SRS performs no onsite ambient air quality
monitoring. State agencies operate ambient air quality monitoring sites in Barnwell, Aiken, and
Richmond Counties. These areas, which include the SRS, are in attainment with National Ambient
Air Quality Standards for sulfur dioxide, nitrogen oxides, carbon monoxide, particulate matter, ozone,
and lead (CFR 1993a).
4.7.2.4 Atmospheric Dispersion Modeling. The SRS has performed atmospheric
dispersion modeling for criteria and toxic air pollutants for both maximum potential and actual
emissions for the base year 1990, using the EPA Industrial Source Complex Short Term No. 2 Model.
The SRS used 1991 meteorological data collected at the Site meteorological stations for input to the
model.
4.7.2.5 Summary of Nonradiological Air Quality. The SRS is in compliance with
National Ambient Air Quality Standards and with the gaseous fluoride and total suspended particulate
standards required by South Carolina Department of Health and Environmental Control (SCDHEC)
Regulation R.61-62.5, Standard 2, "Ambient Air Quality Standards" (AAQS) (see Table 4-4).
The SCDHEC has non-radiological air quality regulatory authority over the SRS. The
Department determines SRS ambient air quality compliance based on SRS air pollutant emissions
modeled at the Site perimeter (excluding SC Highway 125, which crosses the southwestern quadrant of
the SRS).
The SRS is in compliance with SCDHEC Regulation R.61-62.5, Standard 8, "Toxic Air
Pollutants," which regulates the emission of 257 toxic substances. The SRS has identified emission
sources for 139 of the 257 regulated substances; the modeled results indicate that the Site is within
applicable Department of Health and Environmental Control standards (WSRC 1993a). Table 4-5 lists
SRS emissions of toxic air pollutants of concern related to the SRS spent nuclear fuel alternatives,
based on 1990 baseline data and the potential sources of air pollution permitted for construction or
operation in December 1992.
4.7.3 Radiological Air Quality
4.7.3.1 Background and Baseline Radiological Conditions. In the SRS region, airborne
radionuclides originate from natural resources (terrestrial or cosmic), worldwide fallout, and Site
operations. The SRS maintains a network of air monitoring stations on and around the Site to
Table 4-4. Estimated ambient concentration contributions of criteria air pollutants from existing SRS
sources and sources planned for construction or operation through 1995 (-g/m3). ,b
Maximum
SRS Maximum Most stringent Potential
Averaging Potential AAQSd (Federal Concentration
Pollutantc time Concentration Actual or state) as a Percent of
AAQSe
SO2 Annual 18 10 80f 22.5
24-hour 356 185 365f,g 97.5
3-hour 1,210 634 1,300f,g 93
NOx Annual 30 4 100f 30
CO 8-hour 818 23 10,000f,g 8
1-hour 3,553 180 40,000f,g 9
Gaseous fluorides 12-hour 2.40 0.62 3.7e 65
(as HF) 24-hour 1.20 0.31 2.9e 41
1-week 0.6 0.15 1.6e 38
1-month 0.11 0.03 0.8e 14
PM10 Annual 9 3 50f 18
24-hour 93 56 150f 62
O3 1-hour NA NA 235f,g NA
TSP Annual 20 11 75e 2.7
geometric
mean
Lead Calendar 0.0015 0.0003 1.5e 0.1
quarter
mean
a. Source: WSRC (1994b).
b. The contributions listed are the maximum values at the SRS boundary.
c. SO2 = sulfur dioxide; NOx = nitrogen oxides; CO = carbon monoxide; PM10 = particulate matter <
10-m in diameter; TSP = Total Suspended Particulates, O3 = Ozone.
d. AAQS = Ambient Air Quality Standard.
e. Source: SCDHEC (1976).
f. Source: 40 CFR Part 50.
g. Concentration not to be exceeded more than once a year.
NA = Not available.
Table 4-5. Baseline 24-hour average modeled concentrations at the SRS boundary - toxic air
pollutants regulated by South Carolina from existing SRS sources and sources planned for construction
or operation through 1995 (yg/m3).
Maximum
Maximum Potential
Regulatory Potential Actual Concentration as a
Pollutantb Limit Concentrationc Concentrationd Percent of AAQSe
Nitric acid 125 51 4.0 41
1,1,1-Trichloroethane 9,550 81 22 1
Benzene 150 32 31 21
Ethanolamine 200 <0.01 <0.01 <0.1
Ethyl benzene 4,350 0.58 0.12 <0.1
Ethylene glycol 650 0.20 0.08 <0.1
Formaldehyde 7.5 <0.01 <0.01 <0.1
Glycol ethers Pending <0.01 <0.01 -
Hexachloronapthalene 1 <0.01 <0.01 <0.1
Hexane 200 0.21 0.072 <0.1
Manganese 25 0.82 0.10 3
Methyl alcohol 1,310 2.9 0.51 0.2
Methyl ethyl ketone 14,750 6.0 0.99 <0.1
Methyl isobutyl ketone 2,050 3.0 0.51 <0.1
Methylene chloride 8,750 10.5 1.8 <0.1
Naphthalene 1,250 0.01 0.01 <0.1
Phenol 190 0.03 0.03 <0.1
Phosphorus 0.5 <0.001 <0.001 <0.1
Sodium hydroxide 20 0.01 0.01 <0.1
Toluene 2,000 9.3 1.6 <0.1
Trichloroethylene 6,750 4.8 1.0 <0.1
Vinyl acetate 176 0.06 0.02 <0.1
Xylene 4,350 39 3.8 0.9
a. Source: WSRC (1994b).
b. Pollutants listed include compounds of interest regarding spent nuclear fuel alternatives.
c. Maximum potential emissions from all SRS sources for 1990 plus maximum potential emissions
for sources permitted in 1991 and 1992.
d. Actual emissions from all SRS sources plus maximum potential emissions for sources permitted for
construction through December 1992.
e. AAQS = Ambient Air Quality Standard.
determine concentrations of radioactive particulates and aerosols in the air (Arnett et al. 1992).
Table 4-6 lists average and maximum atmospheric radionuclide concentrations at the SRS boundary
and background [160-kilometer (100-mile) radius] monitoring locations during 1991. Table 4-7 lists
the average concentrations of tritium in the atmosphere, as measured at on- and offsite monitoring
locations.
Table 4-6. Radioactivity in air at SRS perimeter and at 160-kilometer (100-mile) radius (pCi/m3).
Gross Nonvolatile
Location Alpha Beta Sr-89,90b Pu-238b Pu-239b
Site perimeter
Average 2.61x10-3 1.78x10-2 4.90x10-5 1.22x10-6 2.11x10-6
Maximum 1.07x10-2 4.63x10-2 5.11x10-4 1.94x10-5 5.40x10-5
Background
(160-kilometer
radius)
Average 2.60x10-3 1.76x10-2 2.00x10-4 1.44x10-6 6.10x10-7
Maximum 9.31x10-3 5.26x10-2 2.08x10-3 2.39x10-5 5.40x10-6
a. Source: Arnett et al. (1992).
b. Monthly composite.
Table 4-7. Average atmospheric tritium concentrations on and around the Savannah River Site
(pCi/m3).
Location 1991 1990 1989
Onsite 250 430 640
Site perimeter 21 32 37
40-kilometer radius 11 12 14
160-kilometer radius 8.5 8.8 9
a. Source: Arnett et al. (1992).
4.7.3.2 Sources of Radiological Emissions. Table 4-8 lists groups of facilities that
released radionuclides to the atmosphere in 1992; the facilities are grouped according to the principal
function that resulted in the release of radioactive materials.
Table 4-9 lists both the identified radionuclides that contributed to the SRS dose and the percent
contribution of each radionuclide to the total site effective dose equivalent.
Table 4-8. Operational groupings and function of radionuclide sources.
Group Function
Reactor Materials Production of fuel and targets
Reactors Irradiation of fuel and targets
Separations Separation of useful radionuclides (other than tritium)
Analytical Laboratories Process Control Laboratories
Tritium Extraction, purification, and packaging
Waste Management Management of radioactive waste
Savannah River Technology Center Research and development to support SRS processes
4.8 Water Resources
4.8.1 Surface Water
The Savannah River bounds the SRS on its southwestern border for about 20 miles
(32 kilometers), approximately 160 river miles (260 kilometers) from the Atlantic Ocean. At the SRS,
river flow averages about 10,000 cubic feet (283 cubic meters) per second. River flows range from
3,960 cubic feet (112 cubic meters) per second to 71,700 cubic feet (2,030 cubic meters) per second.
Five upstream reservoirs - Jocassee, Keowee, Hartwell, Richard B. Russell, and Strom Thurmond
- minimize the effects of droughts and the impacts of low flow on downstream water quality and fish
and wildlife resources in the river.
At the SRS, a swamp occupies the floodplain along the Savannah River for a distance of
approximately 10 miles (17 kilometers); the swamp is about 1.5 miles (2.5 kilometers) wide. A
natural levee separates the river from the swampy floodplain. Figure 4-10 shows the 100-year
floodplain of the Savannah River in the vicinity of the SRS as well as the floodplains of major
tributaries draining the SRS. A 500-year floodplain map of the SRS has not been completed, but
would be required prior to the siting of any spent nuclear fuel management facilities, in compliance
with DOE regulations (CFR 1979). These regulations require DOE to evaluate the potential effects of
flooding to proposed "critical actions" (for example, the storage of highly toxic or water-reactive
materials), which it defines as those for which even a slight chance of flooding would be unacceptable.
The five principal tributaries to the river on the SRS are Upper Three Runs Creek, Fourmile
Branch, Pen Branch, Steel Creek, and Lower Three Runs Creek (Figure 4-10). These tributaries drain
Table 4-9. Annual quantity of radionuclide emissions from the Savannah River Site. ,b
Radionuclide Annual Quantity (curies) Percent of Total Site Dose
H-3 (oxide) 1.00x105 98.0
Pu-239 7.45x10-4 0.6
U-235,238 1.58x10-3 0.4
Pu-238 4.46x10-4 0.3
Ar-41 2.51x102 0.3
I-129 3.50x10-3 0.2
Am-241,243 1.13x10-4 0.1
Sr-89,90 (Y-90) 2.03x10-3 0.02
Cm-242,244 2.31x10-5 0.01
Cs-137 (Ba-137m) 2.50x10-4 0.01
C-14 1.86x10-1 0.01
H-3 (elemental) 5.59x104 <0.01
I-135 1.34x10-1 <0.01
Kr-85 4.99x101 <0.01
I-131 9.99x10-5 <0.01
Ru-106 (Rh-106) 1.81x10-6 <0.01
I-133 1.15x10-3 <0.01
Co-60 3.60x10-7 <0.01
Xe-135 2.43x10-3 <0.01
Cs-134 3.75x10-8 <0.01
Ce-144 (Pr-144,144m)1.16x10-7 <0.01
Eu-154 3.44x10-13 <0.01
Eu-155 1.63x10-13 <0.01
Sb-125 7.27x10-15 <0.01
Zr-95 (Nb-95) 2.39x10-14 <0.01
a. Source: Arnett et al. (1993).
b. Includes emissions to the atmosphere and surface water.
Figure 4-10. Savannah River Site, showing major stream systems and facilities. almost all of the SRS. Each of these streams originates on the Aiken Plateau in the Coastal Plain and
descends 50 to 200 feet (15 to 60 meters) before discharging into the river. The streams, which
historically have received varying amounts of effluent from various SRS operations, are not
commercial sources of water. The natural flow of SRS streams ranges from less than 10 cubic feet
(l cubic meter) per second in smaller streams such as Pen Branch to 240 cubic feet (6.8 cubic meters)
per second in Upper Three Runs Creek.
4.8.1.1 SRS Streams. This section describes the pertinent physical and hydrologic properties
of Upper Three Runs Creek and Fourmile Branch, which are the streams closest to most SRS spent
nuclear fuel management locations (Figure 4-10). These two streams are among the largest on the
SRS, and they border the areas where DOE is most likely to locate new spent nuclear fuel facilities.
Upper Three Runs Creek is a large, cool [annual maximum temperature of 26.1 degrees C
(79 degrees F)] blackwater stream in the northern part of the SRS. It drains an area of approximately
210 square miles (545 square kilometers), and has an average discharge of 330 cubic feet (9.3 cubic
meters) per second at the mouth of the creek. Upper Three Runs Creek is approximately 25 miles
(40 kilometers) long, with its lower 17 miles (28 kilometers) inside the boundaries of the SRS. This
creek receives more water from underground sources than the other SRS streams and, therefore, has
low conductivity, hardness, and pH values. Upper Three Runs Creek is the only major tributary on
the SRS that has never received thermal discharges.
Fourmile Branch is about 15 miles (24 kilometers) long and drains an area of approximately
34 square miles (89 square kilometers). In its headwaters, Fourmile Branch is a small blackwater
stream that receives relatively few impacts from SRS operations. The water chemistry in the
headwater area of the creek is very similar to that of Upper Three Runs Creek, with the exception of
nitrate concentrations, which are an order of magnitude higher than those in Upper Three Runs Creek
(WSRC 1994a). These elevated nitrate concentrations are probably the result of groundwater transport
and outcropping from the F- and H-Area seepage basins. In its lower reaches, Fourmile Branch
broadens and flows through a delta formed by the deposition of sediments. Although most of the flow
through the delta is in one main channel, the delta has many standing dead trees, logs, stumps, and
cypress trees that provide structure and reduce the water velocity in some areas. Downstream of the
delta, the creek flows in one main channel and most of the flow discharges into the Savannah River at
River Mile 152 (kilometer 245), while a small portion of the creek flows west and enters Beaver Dam
Creek, a small onsite tributary.
4.8.1.2 Surface Water Quality. The Savannah River, which forms the boundary between the
States of Georgia and South Carolina, supplies potable water to several users. Upstream of the SRS,
the river supplies domestic and industrial water needs for Augusta, Georgia, and North Augusta, South
Carolina. The river also receives sewage treatment plant effluent from Augusta, Georgia; North
Augusta, Aiken, and Horse Creek Valley, South Carolina; and as described above from a variety of
SRS operations via onsite stream discharges. Approximately 130 river-miles (210 kilometers)
downstream of the SRS, the river supplies domestic and industrial water needs for Savannah, Georgia,
and Beaufort and Jasper Counties in South Carolina through intakes located at about River Mile 29
and River Mile 39. In addition, Georgia Power's Vogtle Electric Generating Plant withdraws an
average of 1.3 cubic meters per second (46 cubic feet per second) for cooling and returns an average
of 0.35 cubic meters per second (12 cubic feet per second) of cooling tower blowdown. Also, the
Urquhart Steam Generating Station at Beech Island, South Carolina withdraws approximately 7.5 cubic
meters per second (265 cubic feet per second) for once-through cooling water.
The South Carolina Department of Health and Environmental Control regulates the physical
properties and concentrations of chemicals and metals in SRS effluents under the National Pollutant
Discharge Elimination System (NPDES) program. This agency also regulates chemical and biological
water quality standards for SRS waters. On April 24, 1992, the agency changed the classification of
the Savannah River and SRS streams from "Class B waters" to "Freshwaters." The definitions of
Class B waters and Freshwaters are the same, but the Freshwaters classification imposes a more
stringent set of water quality standards (Arnett et al. 1993). Tables 4-10 and 4-11 list the
characteristics of SRS surface-water quality upstream and downstream, respectively, due to
contributions from SRS and possibly other sources. A comparison of these results indicates that
influences from SRS or other sources are not seriously degrading Savannah River water quality.
4.8.2 Groundwater Resources
4.8.2.1 Hydrostratigraphic Units. There are two hydrogeologic provinces in the subsurface
beneath SRS (WSRC 1994a). The first, referred to as the Piedmont hydrogeologic province
(Figure 4-11), includes Paleozoic metamorphic and igneous basement rocks and Triassic-aged lithified
mudstone, sandstone, and conglomerate contained within the Dunbarton Basin. The second, referred
to as the Southeastern Coastal Plain hydrogeologic province, represents the major aquifer systems and
consists of a wedge of unconsolidated Coastal Plain sediments of Late Cretaceous and Tertiary age
(Figure 4-11). These two units are overlain by the vadose or unsaturated zone, which extends from
Table 4-10. Water quality in the Savannah River above the confluence with Upper Three Runs near
the Savannah River Site in 1990. ,b
Existing Water-Body Concentrationf
Parameter Unit of Measure MCL c,d or DCGe Average Maximum
Aluminum mg/L 0.05-0.2g NCi 1.1
Ammonia mg/L NAj 0.1 0.2
Cadmium mg/L 0.005g NC <0.01
Calcium mg/L NA NC 4.4
Cesium-137 pCi/L 120e 0.0088 0.030
Chemical oxygen demand mg/L NA 9.7 17
Chloride mg/L 250h 7.8 11
Chromium mg/L 0.1d NC <0.02
Copper mg/L 1.0d NC <0.01
Dissolved oxygen mg/L >5 8.0 9.6
Fecal coliform Colonies per 100/ml 1,000g 54 197
Gross alpha pCi/L 15g 0.04 0.36
Ironc mg/L 0.3h NC 1.5
Lead mg/L 0.015g NC 0.27
Magnesium mg/L NA NC 1.4
Manganesec mg/L 0.05g NC 0.12
Mercury mg/L 0.002d NC <0.0002
Nickel mg/L 0.1c NC <0.05
Nitrite/Nitrate mg/L 10g 0.32 0.99
Nonvolatile beta (dissolved)pCi/L 50g 1.9 3.6
pH pH Units 6.5-8.5g Not reported 7.4
Phosphate mg/L N/A 0.09 0.16
Plutonium-238 pCi/L 1.6e 0.0006 0.0021
Plutonium-239 pCi/L 1.2e 0.0005 0.0021
Sodium mg/L NA NC 11
Strontium-89 pCi/L 800e 0.23 1.0
Strontium-90 pCi/L 8c 0.09 0.22
Sulfate mg/L 250h 7.8 11
Suspended solids mg/L NA 13 22
Temperature Degrees Celsius 32.2k 18.0 27
Total dissolved solids mg/L 500h 62 76
Tritium pCi/L 20,000c 150 1,110
Zinc mg/L 5h NC 0.02
a. Source: Cummins et al. (1991).
b. Parameters are those for which DOE routinely measures as a regulatory requirement or as part of ongoing monitoring programs.
c. Maximum Contaminant Level (MCL), EPA National Primary Drinking Water Regulations (CFR 1974).
d. Maximum Contaminant Level (MCL); South Carolina (1976).
e. U.S. Department of Energy Derived Concentration Guides (DCGs) for Water (DOE 1993b). DCG values are based on committed
effective dose of 100 millirem per year; however, because drinking water MCL is based on 4 millirem per year, number listed is 4
percent of DCG.
f. Average concentration of samples taken at downstream monitoring station. Maximum is highest sampled concentration along reach of
river potentially affected by site activities. Less than (<) indicates concentration below analysis detection limit.
g. Concentration exceeded water quality criteria; however, these criteria are listed for comparison only. Similarly, drinking water standards
and DOE DCGs are listed. Water Quality Criteria (WQCs) and secondary standards are not legally enforceable.
h. Secondary Maximum Contaminant Level (SMCL), EPA National Secondary Drinking Water Regulations (CFR 1991).
i. NC = Not calculated due to insufficient number of samples.
j. NA = None applicable.
k. Shall not exceed weekly average of 32.2 degrees Celsius after mixing nor rise more than 2.8 degrees Celsius in 1 week unless appropriate
temperature criterion mixing zone has been established.
Table 4-11. Water quality in the Savannah River below the confluence with Lower Three Runs near
the Savannah River Site in 1990. ,b
Existing Water-Body Concentrationf
Parameter Unit of Measure MCL c,d or DCGe Average Maximum
Aluminum mg/L 0.05-0.2g NCi 1.1
Ammonia mg/L NAj 0.1 0.2
Cadmium mg/L 0.005g NC <0.01
Calcium mg/L NA NC 4.4
Cesium-137 pCi/L 120e 0.028 0.037
Chemical oxygen demand mg/L NA 9.8 14
Chloride mg/L 250h 8 10
Chromium mg/L 0.1d NC <0.02
Copper mg/L 1.0d NC <0.01
Dissolved oxygen mg/L >5 7.7 9.5
Fecal coliform Colonies per 100/ml 1,000g 54 197
Gross alpha pCi/L 15g 0.08 1.48
Ironc mg/L 0.3h NC 1.5
Lead mg/L 0.015g NC 0.01
Magnesium mg/L NA NC 1.3
Manganesec mg/L 0.05h NC 0.1
Mercury mg/L 0.002d NC <0.0002
Nickel mg/L 0.1c NC <0.05
Nitrite/Nitrate mg/L 10g 0.28 0.43
Nonvolatile beta (dissolved)pCi/L 50g 2.1 5.1
pH pH Units 6.5-8.5h Not reported 8.2
Phosphate mg/L N/A 0.1 0.16
Plutonium-238 pCi/L 1.6e 0.0006 0.0029
Plutonium-239 pCi/L 1.2e 0.0014 0.0079
Sodium mg/L NA NC 11
Strontium-89 pCi/L 800e 0.25 0.98
Strontium-90 pCi/L 8c 0.13 0.30
Sulfate mg/L 250h 8.5 12
Suspended solids mg/L NA 12 19
Temperature Degrees Celsius 32.2k 18.0 27
Total dissolved solids mg/L 500h 63 71
Tritium pCi/L 20,000c 900 6,810
Zinc mg/L 5h NC 0.02
a. Source: Cummins et al. (1991).
b. Parameters are those for which DOE routinely measures as a regulatory requirement or as part of ongoing monitoring programs.
c. Maximum Contaminant Level (MCL), EPA National Primary Drinking Water Regulations (CFR 1974).
d. Maximum Contaminant Level (MCL); South Carolina (1976).
e. U.S. Department of Energy Derived Concentration Guides (DCGs) for Water (DOE 1993b). DCG values are based on committed
effective dose of 100 millirem per year; however, because drinking water MCL is based on 4 millirem per year, number listed is 4
percent of DCG.
f. Average concentration of samples taken at downstream monitoring station. Maximum is highest sampled concentration along reach of
river potentially affected by site activities. Less than (<) indicates concentration below analysis detection limit.
g. Concentration exceeded water quality criteria; however, these criteria are listed for comparison only. Similarly, drinking water standards
and DOE DCGs are listed. Water Quality Criteria (WQCs) and secondary standards are not legally enforceable.
h. Secondary Maximum Contaminant Level (SMCL), EPA National Secondary Drinking Water Regulations (CFR 1991).
i. NC = Not calculated due to insufficient number of samples.
j. NA = None applicable.
k. Shall not exceed weekly average of 32.2 degrees Celsius after mixing nor rise more than 2.8 degrees Celsius in 1 week unless appropriate
temperature criterion mixing zone has been established.
Figure 4-11. Comparison of lithostratigraphy and hydrostratigraphy for the SRS region (not to scale). the ground surface to the water table. The unsaturated zone is a heterogeneous unit of clean, clayey,
or silty sand through which recharge takes place.
The sediments that make up the Southeastern Coastal Plain hydrogeologic province in
west-central South Carolina are grouped into three major aquifer systems divided by two major
confining systems, all of which are underlain by the Appleton confining system (Figure 4-11). The
Appleton system separates the Southeastern Coastal Plain hydrogeologic province from the underlying
Piedmont hydrogeologic province. Locally, each of the major aquifer systems contains individual
aquifer and confining units. Figure 4-11 shows the regional lithostratigraphy of the geologic province
with the attendant primary hydrostratigraphic subdivision of the province. The complexly interbedded
strata that form the three aquifer systems consist primarily of fine- to coarse-grained sand and local
gravel and limestone deposited under relatively high energy conditions in fluvial to shallow marine
environments (WSRC 1994a).
Figure 4-11 shows the current aquifer/aquitard terminology at the SRS. Aquifers, in ascending
order, include the McQueen Branch, the Crouch Branch, and the Steed Pond. For comparison, the
figure also includes the corresponding aquifer terminology used on the Georgia side of the Savannah
River. These include the Midville, Dublin, and Floridan aquifer systems. In addition, the three
aquifers are separated by confining layers which include, in ascending order, the Appleton, Allendale,
and Meyers Branch confining systems (WSRC 1994a).
4.8.2.2 Groundwater Flow. Excellent quality groundwater is abundant in this region of
South Carolina from many local aquifer units. As a result, the South Carolina Department of Health
and Environmental Control has classified all aquifers in the state as Class GB (South Carolina 1976),
or U.S. Environmental Protection Agency (EPA) Class II, meaning that the aquifers can provide
resource-quality water, but are not the sole source of supply (South Carolina Class GA or EPA Class I
aquifers) (DOE 1991b).
The main source of recharge to the vadose zone is rainfall. The annual precipitation at the SRS
is 48 inches (121.9 centimeters), with an estimated 16 inches (41 centimeters) designated as surface
recharge at the center of the SRS, in bare and grass-covered areas (WSRC 1994a). The direction of
groundwater flow in the vadose zone is predominantly downward. However, given the lenses of silt
and clay that exist, there is significant lateral spread in some areas. In general, the vadose zone
thickness ranges from approximately 130 feet (40 meters) in the northernmost portion of the SRS to
0 feet where the water table intersects wetlands, streams, or creeks.
The following discussion of groundwater flow in the Coastal Plain hydrogeologic province
begins with the deepest aquifers at the SRS and proceeds to shallower units. It does not address flow
in the confining units because few hydraulic head measurements are available for these units and, to a
good approximation, flow in aquitards is limited predominantly to vertical flow between aquifer units.
The Midville or McQueen Branch aquifer (which has also been called the Middendorf, the Lower
Cretaceous, the Tuscaloosa, and Aquifer IA) is highly transmissive and, therefore, serves in part as the
production aquifer for much of the SRS. This aquifer flows horizontally, predominantly toward the
Savannah River. In the past, groundwater production wells at the SRS were screened in both the
Midville (McQueen Branch) and Dublin (Crouch Branch) aquifers. In 1985 DOE committed to the
South Carolina Department of Health and Environmental Control to complete production wells only in
the McQueen Branch aquifer to minimize the potential for contamination to reach such wells and
spread in the deeper aquifers.
Flow in the Dublin or Crouch Branch aquifer (which has also been called the Black Creek, the
Tuscaloosa, the Upper Cretaceous, and Aquifer IB) is more complicated than flow in the deeper
McQueen Branch aquifer because of the apparent communication with Upper Three Runs Creek on the
SRS. Nonetheless, horizontal flow in the Dublin (Crouch Branch) aquifer is predominantly toward the
Savannah River. However, there is an upward vertical flow component near the river and Upper
Three Runs Creek. Recharge to the Dublin-Midville aquifer system occurs in areas exposed at the
ground surface near the Fall Line (see Figure 4-3).
Horizontal flow in the Gordon aquifer (previously called the Congaree, the Tertiary, and
Aquifer II) is toward Upper Three Runs Creek and the Savannah River, depending on the area of the
SRS. Both the river and Upper Three Runs Creek intercept this aquifer. The Gordon aquifer receives
most of its recharge from groundwater that originates on the SRS.
Previous SRS studies have called the Upper Three Runs aquifer the "water table aquifer"; others
have defined it as both the Barnwell/McBean and water table aquifers in the central portion of the SRS
where those aquifers were thought to be separated by a "tan clay." The Upper Three Runs aquifer is
the shallowest aquifer at the SRS. The horizontal groundwater flow is generally toward the nearest
surface-water feature that is in communication with the water table. Most SRS streams, except Tims
Branch in the northeastern part of the Site, are in communication with the water table. Tims Branch is
a "losing stream," meaning it provides, or "loses," water to the Upper Three Runs aquifer. However,
the Upper Three Runs aquifer receives most of its recharge from precipitation. The Upper Three Runs
aquifer is not a source of domestic or production water on the SRS because the lower aquifers provide
a more abundant supply of higher quality water (WSRC 1994a).
4.8.2.3 Groundwater Quality. The quality of groundwater in the principal hydrologic
systems beneath the SRS depends on both the source of the water and the inorganic and biochemical
reactions that take place along its flowpath. Quality is strongly influenced by the chemical
composition and mineralogy of the enclosing geologic materials (WSRC 1994a).
In general, the quality of the groundwater in the Coastal Plain sediments at the SRS and the
surrounding areas is suitable for most domestic and industrial purposes. The waters have low
concentrations of total dissolved solids (TDS), ranging from less than 10 milligrams per liter to about
150 to 200 milligrams per liter. The pH values range from 4.9 to 7.7 (where the groundwater is in
contact with limestone). Much of the groundwater is corrosive to metal surfaces due to its low solids
content and frequently low pH values. High dissolved iron concentrations can also be of concern in
some groundwater units. The SRS uses degasification and filtration processes to raise the pH and
remove iron in domestic water supplies where necessary (WSRC 1994a).
Table 4-12 summarizes groundwater quality data from 85 existing waste sites on the SRS
compared to drinking water standards; Table 4-13 lists similar information for selected radiological
constituents. The data in these tables are from ongoing monitoring programs on the Site.
EPA-accepted methods and guidelines for sampling and analysis are an integral part of this monitoring
program. Several of the facilities discussed below have state-approved sampling and analysis plans.
The shallow aquifers beneath 5 to 10 percent of the SRS have been contaminated by industrial
solvents, metals, tritium, or other constituents used or generated on the Site. Figure 4-12 shows the
locations of facilities where the SRS monitors groundwater and areas with constituents that exceeded
drinking water standards in 1992; the concentrations shown on Figure 4-12 represent the maximum
data from one monitoring well on at least one occasion at a given area. Contamination is limited to
the shallow aquifers, with one exception (see next paragraph). Most contaminated groundwater at the
SRS is beneath a few facilities; contaminants reflect the operations and chemical processes those
facilities perform. For example, contaminants in the groundwater beneath A- and M-Areas include
chlorinated volatile organics, radionuclides, metals, and nitrate. At F- and H-Areas, contaminants in
the groundwater include tritium and other radionuclides, metals, nitrate, chlorinated volatile organics at
values much smaller than those found at A- and M-Areas, and sulfate. The groundwater beneath the
Sanitary Landfill contains chlorinated volatile organics, radionuclides, and metals. The groundwater
Table 4-12. Representative groundwater quality data for nonradioactive constituents from the
Savannah River Site.
Parameter (Unit) Standard Maximum Value
Alkalinity (as CaCO3) (mg/L) 100 1,360b
pH (pH units) 8.5c 13b
Antimony (mg/L) 0.005 0.013
Arsenic (mg/L) 0.05 0.1
Beryllium (mg/L) 0.011d 0.0043
Cadmium (mg/L) 0.005c 0.34
Chromium (mg/L) 0.1c 0.82
Mercury (mg/L) 0.002c 0.12
Lead (mg/L) 0.015e 1.0
Nitrate-N (mg/L) 10c 278b
Sulfate (mg/L) 400c 73,500b
Pentachlorophenol (mg/L) 0.001c 0.0032
Lindane (mg/L) 0.0002c 0.00048
Carbon tetrachloride (mg/L) 0.005 0.43
1,2-Dichloroethane (mg/L) 0.005c 0.27
1,1,1-Trichloroethane (mg/L) 0.2c 0.21
1,1-Dichloroethylene (mg/L) 0.007c 0.15
Trichlorethylene (mg/L) 0.005c 147
Tetrachloroethylene (mg/L) 0.005c 101
a. Data compiled from 85 existing wastes sites (Arnett et al. 1993).
b. The elevated values for alkalinity and pH might be due to faulty well installation; the elevated
sulfate and nitrate values might be due to acid spills near wells.
c. National secondary drinking water regulations (CFR 1991).
d. National primary drinking water regulations (CFR 1974).
e. Action level at which providers of public drinking water apply treatment technique to reduce lead
levels (CFR 1991).
Table 4-13. Representative groundwater data for radioactive constituents from the Savannah River
Site (pCi/liter).
Maximum
Constituent Standardb Concentration
Gross alpha 15 2,700
Nonvolatile beta 50 19,000
Tritium 20,000 1.8 x 108
Cesium-137 200 980
Cobalt-60 100 290
Iodine-129 1 72
Ruthenium-106 30 170
Total radium (radium-226 and 5 50
radium-228)
Strontium-90 8 5,300
a. Source: Arnett et al. (1993).
b. National Primary Drinking Water Regulations (CFR 1974), (56 FR 33052).
beneath all the reactor areas except R-Area contains tritium, other nuclides, metals, and chlorinated
volatile organics. At R-Area, groundwater contaminants include radionuclides and cadmium. The
groundwater beneath D-Area contains metals, radionuclides, sulfate, and chlorinated volatile organics.
At TNX-Area, the groundwater contains chlorinated volatile organics, radionuclides, and nitrate (Arnett
et al. 1993). None of these cases indicated the presence of groundwater contamination beyond Site
boundaries. With the ongoing and expanding "pump and treat" system at the A-/M-Area
(Figure 4-12), concentrations in the volatile organic compound plume are likely to decrease with time.
Contamination of groundwater in a drinking water aquifer has been found in only one relatively-
small area north of A-Area, in the northwest portion of the site. In the early 1980s, SRS monitors
found low concentrations of trichloroethylene (11.7 microgram per liter) in water from one production
well (53A) completed to the Dublin-Midville Aquifer System (formerly called the Tuscaloosa
Formation) in M-Area. The monitors found the contamination only at 430 and 480 feet (131 and
146 meters) in this well, which is 670 feet (204 meters) deep. The well is screened intermittently
from 387 feet (118 meters) to the bottom. DOE concluded that the contamination is probably
migrating down the outside well casing from soils near the surface that are contaminated with
trichloroethylene. This contaminated water enters the well through screens set in the Dublin-Midville
Figure 4-12. Groundwater contamination at the Savannah River Site. System (Du Pont 1983). In addition, in 1992 trichloroethylene and tetrachloroethylene were detected
above Primary Drinking Water Standards in cretaceous zone (Dublin-Midville) well MSB 55TA,
which is approximately 3,500 feet west of well 53A and 1,500 feet north of A-Area (Arnett et al.
1993).
4.8.2.4 Groundwater Use. The McQueen Branch aquifer, which becomes shallower toward
the Fall Line, forms the base for most municipal and industrial water supplies in Aiken County.
Toward the coast, in Allendale and Barnwell Counties, this aquifer exists at increasingly greater
depths. As a consequence, the shallower Gordon aquifer supplies some municipal, industrial, and
agricultural users (Arnett et al. 1993).
DOE has identified 56 major municipal, industrial, and agricultural groundwater users within
20 miles (32 kilometers) of the center of the SRS (DOE 1987a). The total pumpage for these users is
about 49 billion liters (13 billion gallons) per year. The SRS withdraws approximately 14.0 billion
liters (3.7 billion gallons) of groundwater per year for domestic and industrial uses (DOE 1990).
4.9 Ecological Resources
The U.S. Government acquired the SRS in 1951. At that time, the Site was approximately
two-thirds forested and one-third cropland and pasture (Dukes 1984). At present, more than
90 percent of the SRS is forested. An extensive forest management program conducted by the
Savannah River Forest Station, which is operated by the U.S. Forest Service, has converted many
pastures and croplands to pine plantations. With the exception of the SRS production and support
areas, natural succession has reclaimed previously disturbed areas. Table 4-14 lists SRS land cover,
other than the land used for nuclear reactors and support facilities.
The SRS is important to maintaining the biodiversity of the region. Satellite imagery of the Site
shows a circle of wooded habitat within a matrix of cleared uplands and narrow forested riparian
corridors. The SRS provides more than 734 square kilometers (181,000 acres) of contiguous forested
cover broken only by unpaved secondary roads, transmission line corridors in various stages of
succession, and a few paved primary roads. Carolina bays, the Savannah River swamp, and several
relatively intact longleaf pine-wiregrass communities provide important contributions to the
biodiversity of the SRS and of the entire region.
Table 4-14. Land cover of undeveloped areas on the Savannah River Site.
Percent of
Land cover types Square kilometer total
Longleaf pine 150 20
Loblolly pine 258 35
Slash pine 117 16
Mixed pine/hardwood 23 3
Upland hardwood 20 3
Bottomland hardwood 117 16
Savannah River swamp 49 7
Total 734 100.0
a. Source: USDA (1991a).
b. To convert square kilometers to acres, multiply by 247.1.
F- and H-Areas, located near the center of the SRS and approximately 1.6 kilometers (1 mile)
southeast of Upper Three Runs Creek, are heavily industrialized with little natural vegetation
remaining inside the fenced areas. These areas are dominated by buildings, paved parking lots,
gravelled construction areas, and laydown yards. While some grassed areas occur around the
administration buildings and some vegetation is present along the ditches that drain the area, the
majority of the site contains no vegetation. Wildlife is absent except for occasional crows (Corvus
brachyrhynchos) and nesting barn swallows (Hirundo rustica) around the buildings.
Figure 2-3 shows the location of a representative host site at the SRS for potential spent nuclear
fuel activities. F- and H-Areas (and developed areas immediately adjacent to them) would house most
spent nuclear fuel management facilities, while the undeveloped area south and east of H-Area would
be used for the construction of new facilities that F- and H-Areas could not accommodate. The
undeveloped area, which was 98 percent cleared fields in 1951, is now almost completely forested, for
the most part with 5- to 40-year-old upland pine stands that are actively managed by the Savannah
River Forest Station. Most of these stands are loblolly pine (Pinus taeda), but there are small stands
of slash pine (P. elliottii), upland hardwoods (predominantly oaks and hickories), and bottomland
hardwoods (most commonly sweetgum, Liquidambar styraciflua, and yellow poplar, Liriodendron
tulipifera) associated with two small Carolina bays located south of H-Area. The area south of H-Area
lies in the Fourmile Branch watershed, while the area east of H-Area is in the McQueen Branch (a
tributary of Upper Three Runs Creek) watershed. Neither area is likely to contain any threatened or
endangered species or their habitats.
The general area of the representative host site contains suitable habitat for white-tailed deer and
feral hogs as well as other faunal species common to the mixed pine/hardwood forests of South
Carolina. Additional wildlife species found in the area include gray squirrel (Sciurus carolinensis), fox
squirrel (S. niger), wild turkey (Meleagris gallopovo), cottontail rabbit (Sylvilagus floridanus), raccoon
(Procyon lotor), bobcat (Felix rufus), and gray fox (Urocyon cinereoargenteus).
4.9.1 Terrestrial Ecology
The SRS is near the transition area between the oak-hickory-pine forest and the southern mixed
forest. As a consequence, species typical of both associations occur (Dukes 1984). In addition,
farming, fire, soil features, and topography have strongly influenced existing SRS vegetation patterns.
A variety of vascular plant communities occurs in the upland areas (Dukes 1984). Typically,
scrub oak communities occur on the drier, sandier areas. Longleaf pine (Pinus palustrus), turkey oak
(Quercus laevis), bluejack oak (Q. incana), blackjack oak (Q. marilandica), and dwarf post oak
(Q. margaretta) dominate these communities, which typically have understories of wire grass (Aristida
stricta) and huckleberry (Vaccinium sp.). Oak-hickory communities occur on more fertile, dry
uplands; characteristic species are white oak (Q. alba), post oak (Q. stellata), southern red oak
(Q. falcata), mockernut hickory (Carya tomentosa), pignut hickory (C. glabra), and loblolly pine, with
an understory of sparkleberry (Vaccinium arboreum), holly (Ilex sp.), greenbriar (Smilax sp.), and
poison ivy (Rhus radicans).
The removal of human residents in 1951 and the subsequent restoration of forest cover has
provided the wildlife of the SRS with excellent habitat. Furbearers such as gray fox, raccoon,
opossum (Didelphis virginiana), bobcat, beaver (Castor canadensis), and otter (Lutra canadensis) are
relatively common throughout the Site. Game species such as gray squirrel and fox squirrel,
white-tailed deer (Odocoileus virginianus), cottontail rabbit, and wild turkey are also common. The
Savannah River Ecology Laboratory has conducted numerous studies of reptile and amphibian use of
the wetlands and adjacent uplands of the SRS.
DOE allows carefully regulated public hunting for white-tailed deer and feral hogs (Sus scrofa)
on most of the SRS to reduce the incidence of animal/vehicle collisions and maintain healthy
populations within the carrying capacity of the range. SRS personnel monitor all animals removed
from the Site for contamination before releasing them to the hunters (WSRC 1992a).
Before releasing any animal to a hunter, SRS technicians perform field analyses for cesium-137
at the hunt site. In 1992, hunters collected 1,519 deer and 168 hogs. The maximum 1992 cesium-137
field measurement for deer was 22.4 picocuries per gram; the average was 6.4 picocuries per gram
(Arnett et al. 1993). For hogs, the maximum value was 22.9 picocuries per gram and the average was
3.5 picocuries per gram. The field technicians determine estimated doses from consumption of the
venison and pork and make this information available to the hunters.
In 1992, the estimated maximum dose received by a hunter was 49 millirem per year. The basis
for this unique hypothetical maximum dose, which was for a hunter who harvested eight deer and one
hog, is the assumption that the hunter consumed the entire edible portion of each animal. An
additional hypothetical model involved a hunter whose total meat consumption for the year consisted
of SRS deer [81 kilograms (179 pounds) per year] (Arnett et al. 1993). Based on these
low-probability assumptions and on the average concentration of cesium-137 (6.4 picocuries in deer
harvested on the SRS), the estimated potential maximum dose from this pathway is 26 millirem; this is
26 percent of the annual 100-millirem DOE Derived Concentration Guide. Although a large
percentage of this hypothetical dose is probably due to cesium-137 from worldwide fallout, the
estimated total contains this background cesium-137 for conservatism.
4.9.2 Wetlands
The SRS has extensive, widely distributed wetlands, most of which are associated with
floodplains, creeks, and impoundments. In addition, approximately 200 Carolina bays occur on the
Site (Shields et al. 1982; Schalles et al. 1989).
The southwestern SRS boundary adjoins the Savannah River for approximately 32 kilometers
(20 miles). The river floodplain supports an extensive swamp, covering about 49 square kilometers
(12,148 acres) of the Site; a natural levee separates the swamp from the river. Timber was cut in the
swamp in the late 1800s. At present, the swamp forest consists of second-growth bald cypress
(Taxodium distichum), black gum (Nyssa sylvatica), and other hardwood species (Workman and
McLeod 1990; USDA 1991a).
Five major streams drain the SRS and eventually flow into the Savannah River. Each stream has
floodplains characterized by bottomland hardwood forests or scrub-shrub wetlands in varying stages of
succession. Dominant species include red maple (Acer rubrum), box elder (A. negundo), bald cypress,
water tupelo (Nyssa aquatica), sweetgum, and black willow (Salix nigra) (Workman and McLeod
1990).
Carolina bays are unique wetland features of the southeastern United States. They are islands of
wetland habitat dispersed throughout the uplands of the SRS. The approximately 200 bays on the Site
exhibit extremely variable hydrology and a range of plant communities from herbaceous marsh to
forested wetland (Shields et al. 1982; Schalles et al. 1989). SRS scientists have studied Carolina bay
ecology extensively, particularly in relation to the construction of the Defense Waste Processing
Facility (DWPF; SREL 1980).
4.9.3 Aquatic Ecology
The aquatic resources of the SRS have been the subject of intensive study for more than
30 years. Research has focused on the flora and fauna of the Savannah River and the five tributaries
of the river that drain the Site. Section 4.8.1.1 describes those portions of the aquatic systems that
spent nuclear fuel management activities could affect. In addition, several monographs (Patrick et al.
1967; Dahlberg and Scott 1971; Bennett and McFarlane 1983), the eight-volume Comprehensive
Cooling Water Study (Du Pont 1987), and three EISs (DOE 1984; DOE 1987b; DOE 1990) that
evaluated operations of SRS production reactors describe the aquatic biota and aquatic systems of the
SRS.
4.9.4 Threatened and Endangered Species
Threatened, Endangered, and Candidate Plant and Animal Species of the Savannah River Site
(HNUS 1992b) describes threatened, endangered, and candidate plant and animal species that are
known to occur or that might occur on the SRS. Table 4-15 lists these species.
The following Federally listed endangered animals are known to occur on the SRS or in the
Savannah River adjacent to the Site: the red-cockaded woodpecker (Picoides borealis), the southern
bald eagle (Haliaeetus leucocephalus), the wood stork (Mycteria americana), and the shortnose
sturgeon (Acipenser brevirostrum) (HNUS 1992b). Researchers have found one Federally listed
endangered plant species, the smooth coneflower (Echinacea laevigata), on the Site, several Federally
Table 4-15. Threatened, endangered, and candidate plant and animal species of the SRS.
Common Name (Scientific Name) Status
Animals
Rafinesques (= Southeastern) big-eared bat (Plecotus rafinesquii) FC2
Loggerhead Shrike (Lanius ludovicianus) FC2
Bachman's sparrow (Aimophila aestivalis) FC2
Carolina crawfish (= Gopher) frog (Rana areolata capito) FC2
Southern hognose snake (Heterodon simus) FC2
Northern pine snake (Pituophis melanoleucus melanoleucus) FC2
Bald eagle (Haliaeetus leucocephalus) E
Wood stork (Mycteria americana) E
Red-cockaded woodpecker (Picoides borealis) E
American alligator (Alligator mississippiensis) T/SA
Shortnose sturgeon (Accipenser brevirostrum) E
Plants
Smooth coneflower (Echinacea laevigata) E
Bog spice bush (Lindera subcoriacea) FC2
Boykin's lobelia (Lobelia boykinii) FC2
Loose watermilfoil (Myriophyllum laxum) FC2
Nestronia (Nestronia umbellula) FC2
Awned meadowbeauty (Rhexia aristosa) FC2
Key: E = Federal endangered species.
T/SA = Threatened due to Similarity of Appearance.
FC2 = Under review (a candidate species) for listing by the Federal government.
listed Category 2 species, and several state listed species (Knox and Sharitz 1990). At present, the
SRS is implementing strategies for the protection of these species.
F- and H-Areas and the representative host site contain no habitat suitable for any of the
Federally listed threatened or endangered species found on the SRS. The Southern bald eagle and the
wood stork feed and nest near wetlands, streams, and reservoirs, and thus would not be attracted to the
host site, a densely forested upland area. Shortnose sturgeon, typically residents of large coastal rivers
and estuaries, have never been collected in Fourmile Branch or any of the tributaries of the Savannah
River that drain the SRS.
Red-cockaded woodpeckers prefer open pine forests with mature trees (older than 80 years) for
foraging and nesting. The pines of the undeveloped host site are 5 to 40 years old, thus red-cockaded
woodpeckers probably would not forage or nest in the area.
The Red-cockaded Woodpecker Management Standards and Guidelines, Savannah River Site
(USDA 1991b) describes the SRS management strategy for the red-cockaded woodpecker. The most
significant element of this management strategy is the conversion of slash (and some loblolly) pine in
a designated red-cockaded woodpecker management area to longleaf pine, with a harvest rotation of
120 years.
4.10 Noise
The major noise sources at the SRS occur primarily in developed operational areas and include
various facilities, equipment, and machines (e.g., cooling towers, transformers, engines, pumps, boilers,
steam vents, paging systems, construction and materials-handling equipment, and vehicles). Major
noise sources outside the operational areas consist primarily of vehicles and railroad operations.
Previous studies have assessed noise impacts of existing SRS operational activities (NUS 1991b; DOE
1991b; DOE 1990; DOE 1993a). These studies concluded that, because of the remote locations of the
SRS operational areas, there are no known conditions associated with existing onsite noise sources that
adversely affect individuals at offsite locations. Some disturbance of wildlife activities might occur on
the SRS as a result of operational and construction activities.
Existing SRS-related noise sources of importance to the public are those resulting from the
transportation of people and materials to and from the Site. These sources include trucks, private
vehicles, helicopters, and freight trains. In addition, a portion of the air cargo and business travel
using commercial air transport through the airports at Augusta, Georgia, and Columbia, South
Carolina, are attributable to SRS operations.
The States of Georgia and South Carolina and the counties in which the SRS is located have not
established any regulations that specify acceptable community noise levels with the exception of Aiken
County. A provision of the Aiken County Nuisance Ordinance limits daytime and nighttime noise by
frequency band (Aiken County 1991).
During a normal week in 1995, about 20,000 employees are likely to travel to the SRS each day
in private vehicles from surrounding communities. Both government-owned and private trucks pick up
and deliver materials at the Site. Most private vehicles and trucks traveling to and from the Site each
day use South Carolina Highways (SC) 125 and 19. The contribution of SRS operations to traffic
volumes along SC 125 and SC 19, especially during peak traffic periods, affects noise levels through
the towns of New Ellenton and Jackson and the City of Aiken.
Noise measurements taken during 1989 and 1990 along SC 125 in the Town of Jackson at a
point about 15 meters (50 feet) from the roadway indicate that the 1-hour equivalent sound level from
traffic ranged from 48 to 72 decibels (A-weighted). The estimated day/night average sound level
along this route was 66 decibels for summer and 69 decibels for winter. Similarly, noise
measurements along SC 19 in the town of New Ellenton at a point about 15 meters (50 feet) from the
roadway indicate that the 1-hour equivalent sound level from traffic ranged from 53 to 71 decibels.
The estimated day/night average sound level along this route was 68 decibels for summer and
67 decibels for winter (NUS 1990). Employment at the SRS has increased slightly since 1989,
potentially causing small increases in traffic noise, especially during peak traffic periods
(approximately between 6:30 and 8:30 a.m. and between 3:30 and 5:30 p.m., corresponding to the
major shift changes). Because some residences and at least two schools are within 100 to 200 feet of
these routes, some annoyance to members of the public residing along these highways might occur
based on the relationship between the day/night average sound level and the "percent highly annoyed"
(Schultz 1978; Fidell et al. 1989; FICON 1992).
Noise sources from rail transport include diesel engines, wheel-track contact, and
whistle-warnings at rail crossings.
4.11 Traffic and Transportation
4.11.1 Regional Infrastructure
The SRS is surrounded by a system of Interstate highways, U.S. highways, state highways, and
railroads. The regional transportation networks service the four South Carolina counties (Aiken,
Allendale, Bamberg, and Barnwell) and two Georgia counties (Columbia and Richmond) that generate
about 90 percent of SRS commuter traffic (HNUS 1992a). Two major railroads - CSX Transportation
and Norfolk Southern Corporation - also serve the SRS vicinity. Although barge traffic is possible on
the Savannah River, neither the SRS nor commercial shippers normally use barges. Figure 4-13 shows
the regional transportation infrastructure.
4.11.1.1 Regional Roads. Two Interstate highways serve the SRS area. Interstate 20 (I-20)
provides a primary east-west corridor and I-520 links I-20 with parts of Augusta, Georgia.
U.S. Highways 1 and 25 are principal north-south routes and U.S. 78 provides east-west connections.
Several other highways - U.S. 221, U.S. 301, U.S. 321, and U.S. 601 - provide additional transport
routes in the region.
Several state routes provide direct access to the SRS. Running northwest/southeast is SC 125.
Access to the Site is provided from the north by SC 19, from the northeast by SC 39, and from the
east by SC 64.
U.S. 278 bisects the northern part of the SRS and is available to public access without restriction.
The SRS maintains barricades at site entries and exits on SC 125 to control public access if necessary,
although it is generally open to unrestricted public travel. The public also has direct access to Site
Road 1. All other site roads have restricted access.
4.11.1.2 Regional Railroads. Norfolk Southern serves Augusta and Savannah, Georgia, as
well as Columbia and Charleston, South Carolina. CSX serves the same locations and the SRS.
4.11.2 SRS Infrastructure
The SRS transportation infrastructure consists of more than 143 miles (230 kilometers) of
primary roads, 1,200 miles (1,931 kilometers) of unpaved secondary roads, and 103 kilometers
(64 miles) of railroad track (WSRC 1993b). These roads and railroads provide connections among the
various SRS facilities and to offsite transportation linkages. Figure 4-14 shows the SRS network of
primary roadways and access points. Figure 4-15 shows the SRS railway system.
4.11.2.1 SRS Roads. Two major public highways traverse the Site: SC 125 and U.S. 278.
SC 125 connects Allendale, South Carolina, to Augusta, Georgia, by crossing the Site in a
northwest-to-southeast direction. U.S. 278 also connects Augusta and Allendale, but its route
approximately follows the northern and eastern SRS boundaries.
Figure 4-13. Regional transportation infrastructure. Figure 4-14. Major SRS road and access points. Figure 4-15. SRS railroads lines. Ten barricades around the Site limit access from public roads. Five barricades limit SRS access
from SC 125; three limit access from SC 19, SC 39, and SC 64; and two limit access from the public
areas of the administrative complex near the northern SRS boundary (A-Area).
In general, the primary SRS roadways are in good condition and are smooth and free from
potholes. Typically, wide, firm shoulders border roads that are either straight or have wide gradual
turns. Intersections are well marked for both traffic and safety identification and are sufficiently
cleared of trees and brush that might obstruct a driver's view of oncoming traffic. Railings along the
side of the roadways offer protection at appropriate locations from dropoffs or other hazards. In
general, the roadways are lighted only at gate areas and near major facilities. The SRS has two
overpasses, one at the cloverleaf intersection of Roads 2 and C, and the other where SC 125
overpasses the CSX railroad tracks in the southern part of the Site. The 60 bridges on the Site have
been inspected and evaluated for safe loading, with some bridges rated as high as 200 tons (181 metric
tons) under controlled conditions. The steepest roadway gradient is on Road C at the east bank of
Upper Three Runs Creek, where the road drops more than 100 feet (30 meters) in about 0.25 miles
(0.4 kilometer). At the base of the dropoff is a bridge over the creek and an immediate turn in the
road. This area presents a relatively hazardous roadway condition.
In general, heavy traffic occurs early in the morning and late in the afternoon when workers from
surrounding communities commute to and from the Site. During working hours, official vehicles and
logging trucks constitute most of the traffic. At any time, as many as 60 logging trucks, which can
impede traffic, might be operating on the Site, with an annual average of about 25 trucks per day.
Table 4-16 provides data on traffic counts for various roads and access points around the SRS.
4.11.2.2 SRS Railroads. Railroads on the Site include both CSX tracks and SRS rolling
stock and tracks. Two routes of the CSX distribution system run through the Site: a line between
Florence, South Carolina, and Augusta, Georgia, and a line between Yemassee, South Carolina, and
Augusta, Georgia. The two lines join on the Site just south of L-Lake (Figure 4-15). Early in 1989
CSX discontinued service on the line from the SRS junction to Florence.
The 64 miles (103 kilometers) of SRS railroads are well maintained. The rails and crossties are
in good condition, and the track lines are clear of vegetation and debris. Significant clear areas border
the tracks on both sides. Intersections of railroads and roadways are marked by railroad crossing signs
with lights where appropriate.
Table 4-16. SRS traffic counts - major roads.
Average
Day Peak speed
Measurement point Date DirectionTotal Peakb timec (mph)d
Road 2 between Roads C and D 2-23-93 East 3,031 800 1530 47
4-21-93 West 3,075 864 0630 NAe
Road 4 between Roads E and C 12-9-92 East 1,624 352 1530 NA
12-9-92 West 1,553 306 0615 NA
Road 8 at Pond C 2-23-92 East 634 274 1530 58
2-23-92 West 662 331 0615 56
Road C between landfill and R12-16-92 North 6,931 2,435 1530 53
12-16-92 South 6,873 2,701 0630 58
Road C north of Road 7 1-20-93 North 742 288 0630 53
1-20-93 South 763 223 1530 54
Road D 9-29-93 North 1,779 218 1500 43
9-29-93 South 1,813 220 0845 52
Road E at E-Area 8-25-93 North 3,099 669 1530 35
8-25-93 South 3,054 804 0630 38
Road F at Upper Three Runs Cr2-2-93 North 3,239 1,438 1530 53
2-2-93 South 3,192 1,483 0630 51
H-Area Exit 12-2-92 Outbound 2,181 406 1530 12
a. Source: Swygert (1993).
b. Number of vehicles in peak hour.
c. Start of peak hour.
d. mph = miles per hour; to convert to kilometers per hour multiply by 1.6093.
e. NA = data not available.
The SRS rail classification yard is east of P-Reactor. This eight-track facility sorts and redirects
rail cars. Deliveries of SRS shipments occur at two onsite rail stations at the former towns of Ellenton
and Dunbarton. From these stations, an SRS engine moves the railcars to the appropriate receiving
facility. The Ellenton station, which is on the main Augusta-Yemassee line, is the preferred delivery
point. The Dunbarton station, which is on the discontinued portion of the Augusta-Florence line,
receives less use.
4.12 Occupational and Public Radiological Health and Safety
The sources of radiation exposure to individuals consist of natural background radiation from
cosmic, terrestrial, and internal body sources; radiation from medical diagnostic and therapeutic
practices; and radiation from manmade sources, including consumer and industrial products, nuclear
facilities, and weapons test fallout.
All radiation doses discussed in this document are effective dose equivalents (i.e., organ dose
equivalents weighted for biological effect and summed to yield a whole-body dose equivalent with the
same risk as irradiation of individual organs) as defined by the International Commission on
Radiological Protection, Publication 26 (ICRP 1977), unless specifically identified otherwise (e.g.,
thyroid dose, bone dose).
Natural background radiation contributes about 83 percent of the annual dose of 380 millirem
received by an average member of the population within 50 miles (80 kilometers) of the Site. Based
on national averages, medical exposure accounts for 14 percent of the annual dose, and the combined
doses from weapons test fallout, consumer and industrial products, and air travel account for
approximately 3 percent (Arnett et al. 1993).
4.12.1 Occupational Health and Safety
SRS maintains a network of air monitoring stations on and around the Site to determine the
concentrations of radioactive particulates and aerosols in the air (Arnett et al. 1993). Table 4-17 lists
average and maximum radionuclide particulate concentrations found in 1992 in air at the F- and
H-Areas, SRS boundary, and background [100-mile (160-kilometer) radius] monitoring locations.
Table 4-18 lists average and maximum concentrations of tritium in atmospheric moisture during 1992
for the F- and H-Areas, SRS boundary, and background monitoring locations.
Gamma radiation levels measured by thermoluminescent dosimeters in 1992 at the F- and H-Area
fences averaged 70 and 74 millirem per year, respectively. Gamma radiation levels, including natural
background (terrestrial and cosmic) radiation, measured at the Site perimeter in 1992 yielded an
average dose of 35 millirem per year (Arnett et al. 1993).
Table 4-17. Radioactivity in air at the Savannah River Site and vicinity (pCi/m3).
Gross Nonvolatile
Location Alpha Beta SR-89,90b Pu-238b Pu-239b
F-Area
Average 1.80x10-3 1.94x10-2 0.62x10-4 1.26x10-5 8.15x10-6
Maximum 3.55x10-3 5.56x10-2 6.02x10-4 2.64x10-5 2.48x10-5
H-Area
Average 1.80x10-3 1.93x10-2 2.69x10-4 2.03x10-5 5.14x10-6
Maximum 4.24x10-3 5.39x10-2 2.83x10-3 6.03x10-5 1.41x10-5
Site perimeter
Average 1.80x10-3 2.30x10-2 0.13x10-4 0.01x10-7 2.40x10-7
Maximum 4.04x10-2 4.95x10-2 4.54x10-4 2.21x10-6 2.76x10-6
Background
(100-mile radius)
Average 1.67x10-3 1.73x10-2 0.49x10-4 0.72x10-6 <1.00x10-6
Maximum 3.83x10-3 4.37x10-2 6.89x10-4 1.98x10-5 6.15x10-6
a. Arnett et al. (1993).
b. Monthly composite.
Table 4-18. Tritium measured in air at the Savannah River Site (pCi/cc).
Location Average Maximum
F-Area 8.67x10-5 2.98x10-4
H-Area 0.99x10-3 6.77x10-3
Site boundary 2.65x10-5 1.03x10-4
Background (100-mile radiu8.32x10-6 1.08x10-5
a. Arnett (1993).
Soil samples from uncultivated areas provide a measure of the quantity of particulate radioactivity
deposited from the atmosphere. Table 4-19 lists maximum measurements of radionuclides in the soil
for 1992 at F- and H-Areas, SRS boundary, and background [100-mile (160-kilometer)-radius]
monitoring locations. The SRS measured elevated concentrations of plutonium-238 and plutonium-239
around F- and H-Areas, reflecting releases from these areas. From 1955 through 1992, total
atmospheric plutonium releases from the F- and H-Areas were approximately 0.7 curie of
plutonium-238 and 3 curies of plutonium-239 (Arnett et al. 1992; 1993).
The SRS workers investigated for purposes of assessing occupational radiation exposures belong to
the group of involved workers assigned to F- and H-Area facilities. The investigation selected these
facilities because they process materials with radiological characteristics similar to the materials being
Table 4-19. Maximum radioactivity concentrations in soil at the Savannah River Site (pCi/g).
Location Sr-90 Cs-137 Pu-238 Pu-239
F-Area 2.16x10-2 7.19x10-1 4.03x10-1 5.31x10-1
H-Area 2.89x10-2 8.22x10-1 2.13x10-2 5.54x10-2
Site perimeter (b) 4.84x10-1 2.19x10-3 1.36x10-2
Background (100-mile radius)1.46x10-2 (b) 2.34x10-4 1.93x10-2
a. Arnett et al. (1992).
b. None detected.
analyzed in this EIS. The dosimetry results for these two involved worker groups are most useful
because they depict occupational impacts that are directly relevant to each alternative. The
investigation selected two dosimetry periods of record for this analysis: 1983 - 1987 and 1993. The
earlier 5-year period included times when materials processing was occurring at a rate that was
accelerated in comparison with recent years. The later period includes processing rates that better
reflect near-term DOE mission initiatives.
Tables 4-20 and 4-21 list the involved worker dosimetry data for 1983 - 1987 and 1993,
respectively. This analysis adapted these data from monitoring data statistics (Matheny 1994a;
Matheny 1994b) for operations, maintenance, laboratory, and health protection personnel assigned to
the F- and H-Area Canyons and the associated B-Line facilities. The calculated incidences of excess
fatal cancer attributable to each facility's collective worker dose are approximately 0.11 and 0.037 for
the earlier and later time periods, respectively. Similarly, the highest calculated excess fatal cancer
probabilities attributable to average individual worker doses are approximately 0.0003 and 0.0001,
respectively. The analysis estimated these health effects using risk coefficients adopted by DOE
(DOE 1993).
4.12.2 Public Health and Safety
Table 4-22 summarizes the major sources of exposure for the population within 50 miles
(80 kilometers) of the SRS and for the Savannah River water-consuming population in Beaufort and
Jasper Counties, South Carolina, and Port Wentworth, Georgia. Most of the sources, such as natural
background dose and medical dose, are independent of the presence of the SRS.
Atmospheric releases of radioactive material to the environment from SRS operations from 1990 to
1992 resulted in an average dose of approximately 0.02 millirem per year to individuals in the 50-mile
Table 4-20. Annual involved worker doses, 1983 - 1987.
Total Collective
Average Worker Worker Dose
Facility Dose (rem) (person-rem)
H-Canyon 0.41 36.28
HB-Line 0.49 21.84
F-Canyon 0.48 87.25
FB-Line 0.74 124.68
Facilities Average0.53 NA
Facilities Total NA 270.05
NA = Not applicable.
Table 4-21. Annual involved worker doses, 1993.
Total Collective
Average Worker Worker Dose
Facility Dose (rem) (person-rem)
H-Canyon 0.17 11.07
HB-Line 0.24 21.97
F-Canyon 0.22 9.16
FB-Line 0.24 51.16
Facilities Average0.22 NA
Facilities Total NA 93.36
NA = Not applicable.
Table 4-22. Major sources of radiation exposure to the public in the vicinity of the Savannah River
Site.
Dose to average
individual Percentage of
Source of Exposure (mrem/yr) exposure
Natural background radiation 315 83
Medical radiation 54 14
Consumer and industrial products, fallout, air travel 10 3
Savannah River Site operations 0.22 0.06
Grand Total 380 100
a. Arnett et al. (1993).
(80-kilometer)-radius population. The collective effective dose equivalent due to atmospheric releases
from 1992 SRS operations to the population of 620,100 within 50 miles (80 kilometers) was
approximately 6.4 person-rem per year. Atmospheric releases of tritium accounted for more than
90 percent of the offsite population dose; tritium is the only radionuclide of SRS origin that is
routinely detected in offsite air (Cummins et al. 1991; Arnett et al. 1992, 1993). Table 4-23 lists
average annual atmospheric tritium concentrations in the vicinity of SRS for the three years ending in
1992.
Table 4-23. Average atmospheric tritium concentrations in the vicinity of the Savannah River Site
(pCi/m3).
Location 1992 1991 1990
Onsite 340 250 430
Site perimeter 27 21 32
25-mile radius 11 11 12
100-mile radius 8.3 8.5 8.8
a. Arnett et al. (1993).
From 1990 to 1992, the calculated maximum individual average annual dose from atmospheric
releases to a hypothetical individual residing at the SRS boundary was 0.12 millirem (Cummins et al.
1991; Arnett et al. 1992, 1993).
In general, liquid releases of tritium account for more than 99 percent of the total radioactivity
introduced into the Savannah River from SRS activities (Arnett et al. 1993). The calculated average
annual dose to the maximally exposed individual resulting from liquid releases from 1990 to 1992 was
0.21 millirem (Cummins et al. 1991; Arnett et al. 1992; 1993). From 1990 to 1992 liquid releases of
radioactive material to the environment from SRS operations resulted in an average dose of 0.04
millirem per year and 0.05 millirem per year to downstream consumers of drinking water from the
Beaufort-Jasper and Port Wentworth water treatment plants, respectively. These doses to the current
Beaufort-Jasper river-water-consuming population of about 51,000 and the current Port Wentworth
river-water-consuming population of about 20,000 would yield a collective effective dose equivalent to
these populations of approximately 3 person-rem per year (Cummins et al. 1991; Arnett et al. 1992,
1993).
The SRS analyzes samples from other environmental media that onsite releases might affect and
that might provide a pathway for radiation exposure to the public and Site employees; these include
samples of milk, food products, drinking water, wildlife, rainwater, soil, sediment, and vegetation.
The 1992 SRS Environmental Report (Arnett et al. 1993) describes the sampling program, monitoring
locations, and monitoring results for each of these media.
Major nuclear facilities within 50 miles (80 kilometers) of the SRS include a low-level waste
burial site operated by Chem-Nuclear Systems, Inc., near the eastern SRS boundary in Barnwell, South
Carolina, and the Georgia Power Company Alvin W. Vogtle Electric Generating Plant, directly across
the Savannah River from the SRS. Plant Vogtle began commercial operation in 1987, and its releases
are controlled to meet U.S. Nuclear Regulatory Commission requirements.
4.13 Utilities and Energy
This section describes SRS electricity consumption, water consumption, fuel usage, and domestic
and industrial wastewater treatment. Table 4-24 contains information on the current status of these
items at SRS.
Table 4-24. Current capacities and usage of utilities and energy at SRS.
ELECTRICITY
Consumption 659,000 megawatt hours per year
Load 75 megavolt-amperes
Peak Demand 130 megavolt-amperes
Capacity 340 megavolt-amperes
WATER
Groundwater usage 12,490 million liters (3.3 billion gallons) per year
Surface water usage (cooling) 75,700 million liters (20 billion gallons) per year
FUEL
Oil 28.4 million liters (7.5 million gallons) per year
Coal 210,000 metric tons (230,000 tons) per year
Gasoline 4.7 million liters (1.24 million gallons) per year
WASTEWATER
Domestic capacity 3.97 million liters (1.05 million gallons) per day
Domestic load 1.89 million liters (0.50 million gallons) per day
Industrial capacitya,b 1.64 million liters (433,244 gallons) per day
Industrial loada 44,000 liters (11,580 gallons) per day
a. F/H Effluent Treatment Facility only.
b. Design capacity; permitted capacity is about 67 percent of this value.
4.13.1 Electricity
The SRS purchases electric power from the South Carolina Electric and Gas Company (SCE&G)
through three purchased power-line interconnects to the SRS transmission grid. The recent total
annual power consumption for the SRS was approximately 659,000 megawatt-hours. The average load
was 75 megavolt-amperes and the peak demand was about 130 megavolt-amperes. South Carolina
Electric and Gas sources can supply as much as 340 megavolt-amperes to the SRS grid with existing
direct connections. The SRS generating station in D-Area can produce an additional
80 megavolt-amperes capacity, although that plant currently produces only process steam. The SRS
transmission grid that would provide power to any spent nuclear fuel facilities consists of more than
145 kilometers (90 miles) of 115-kilovolt lines, four switching stations, and 15 substations. Electric
service to all major production areas provides parallel redundant capacity to ensure maximum
availability and reliability (WSRC 1993c).
4.13.2 Water Consumption
Groundwater from a deep confined aquifer supplies domestic and process water for the SRS
through approximately 100 production wells. The aquifer system sustains single well yields of about
10.2 million liters (2.7 million gallons) per day. Current usage from this source is about 14.0 billion
liters (3.7 billion gallons) per year (DOE 1990). The SRS withdraws cooling water for its facilities
from the Savannah River at an annual rate of about 75.7 billion liters (20 billion gallons)
(WSRC 1993c).
4.13.3 Fuel Consumption
Fuels consumed at SRS include oil, coal, and gasoline. SRS facilities and equipment burn
approximately 28.4 million liters (7.5 million gallons) of oil each year. This total includes diesel fuel,
No. 6 oil, and No. 2 oil. The SRS burns coal and some waste oils in the D-Area powerhouse to
produce steam for Site facilities. Current coal usage is about 208,655 metric tons (230,000 tons) per
year. SRS vehicles use approximately 4.7 million liters (1.24 million gallons) of gasoline annually.
Under the provisions of the Energy Policy Act of 1992, natural gas will replace gasoline on the SRS
within the next 10 years. At that time, SRS usage of natural gas would be approximately 12.2 million
cubic meters (429 million cubic feet) per year. At present, the SRS consumes no natural gas
(WSRC 1993c).
4.13.4 Wastewater Treatment
By 1995, the SRS Centralized Sanitary Wastewater Treatment Facility will process most of the
domestic effluent on the Site. This centrally located facility has a design capacity of 4 million liters
(1.05 million gallons) per day. Once operational, the plant will use about 50 percent of this capacity.
In addition, five smaller sanitary treatment plants serve more remote areas of the Site. Facilities for
spent nuclear fuel management would use the centralized facility.
The F/H Effluent Treatment Facility (ETF), which decontaminates routine process effluents and
accidental radioactive releases from operations, treats industrial wastewater in the F- and H-Areas,
where the spent fuel management activities would occur.
Effluent Treatment Facility process operations performed on the waste liquids include
neutralization (adjusts pH), submicron filtration (removes suspended solids), activated carbon
absorption (removes dissolved organic chemicals), reverse osmosis membrane deionization (removes
salts), ion exchange (removes heavy metals), and evaporation (separates radionuclides from aqueous
condensate). This facility releases two different streams. The treated water stream is sampled and
analyzed to ensure that it meets discharge requirements and then is released to Upper Three Runs
Creek via a permitted outfall. The waste concentrate (i.e., bottoms from the evaporator process) is
transferred to the H-Area waste tank farm for treatment and disposal in the Z-Area Saltstone facility.
The design capacity for the Effluent Treatment Facility is approximately 600 million liters (158
million gallons) per year. The maximum permitted treatment capacity is about 400 million liters
(105.7 million gallons) per year. Under normal operating conditions, the facility treats more than
16,000 cubic meters (26 million gallons) of liquid waste per year (WSRC 1993d).
The influent water load to processes discharging to the permitted outfall includes as much as 205
million liters (54 million gallons) per year of F-Area Canyon process wastewater, 120 million liters
(32 million gallons) per year of H-Area Canyon process wastewater, 34 million liters (9 million
gallons) per year from the F-Area collection and retention basins, 34 million liters (9 million gallons)
per year from the H-Area collection and retention basins, 68 million liters (18 million gallons) per year
of Effluent Treatment Facility acid, caustic, flush and rinse water, and similar wastewater from other
SRS facilities.
4.14 Materials and Waste Management
The historic national defense mission of the SRS has resulted in the generation of high-level
radioactive waste, transuranic waste, low-level radioactive waste (low-activity and intermediate-level),
hazardous waste, mixed waste (radioactive and hazardous combined), and sanitary waste
(nonhazardous, nonradioactive solid waste). This section discusses the treatment, storage, and disposal
of waste at the SRS. Section 4.13 discusses domestic and industrial wastewater treatment.
DOE is preparing an environmental impact statement on Waste Management at the Savannah
River Site (DOE 1995). The purpose of the EIS is to provide a basis for DOE to select a sitewide
strategic approach to managing present and future SRS waste generated as a result of ongoing
operations, environmental restoration activities, transition from nuclear production to other missions,
and decontamination and decommissioning programs. The Waste Management EIS will support
project-level decisions on the operation of specific treatment, storage, and disposal facilities within the
near term (10 years or less). In addition, the EIS will provide a baseline for analyses of future waste
management activities and a basis for the evaluation of the specific waste management alternatives.
The Waste Management EIS will not include management of spent nuclear fuel which is addressed in
this document.
DOE treats and stores waste generated from onsite operations in waste management facilities
located primarily in E-, F-, H-, N-, S-, and Z-Areas (Figure 4-16). These facilities include the F- and
H-Area Effluent Treatment Facility, the High-Level Waste Tank Farms, and the Solid Waste Disposal
Facility. The Defense Waste Processing Facility is nearly operational and the Consolidated
Incineration Facility is under construction. The SRS places sanitary and inert waste in the Interim
Sanitary Landfill and the Burma Road Landfill, respectively.
DOE continues to reduce the amount of waste generated and disposed of at the SRS through
waste minimization and treatment programs. DOE accomplishes waste minimization by reducing the
volume, toxicity, or mobility of waste before storing or disposing of it. These activities also include
more intensive surveying, waste segregation, and use of administrative and engineering controls.
The waste that DOE presently stores on the SRS includes high-level, transuranic, hazardous,
mixed waste and some low-level waste. The Site stores high-level waste in underground storage tanks
that have received South Carolina Department of Health and Environmental Control industrial
wastewater permits, and manages them in accordance with Clean Water Act, Resource Conservation
and Recovery Act, and DOE requirements. The SRS stores transuranic mixed waste on interim-status
storage pads in accordance with South Carolina Department of Health and Environmental Control
requirements and DOE Orders. Hazardous and mixed waste is placed in permitted or interim-status
Figure 4-16. Waste management facilities at the Savannah River Site. storage in the Hazardous Waste Storage Facilities (both buildings and pads) and in the mixed waste
storage buildings.
Figure 4-17 shows the high-level liquid waste management process at the SRS. Figure 4-18
shows the process for handling all other forms of solid waste at the Site.
Table 4-25 is a forecast of annual waste generation for all waste forms except sanitary and
high-level waste (WSRC 1994c). The volumes listed do not include waste related to decontamination
and decommissioning because DOE has not yet completed the planning of these activities.
Section 5.14 discusses potential consequences of spent nuclear fuel activities as they relate to the
alternative interim storage and treatment scenarios.
4.14.1 High-Level Waste
The SRS generated high-level waste from the recovery of nuclear materials from spent fuel and
target processing in the F- and H-Areas. It is stored in 50 underground tanks. These tanks also store
other radioactive waste effluents (primarily low-level radioactive waste such as aqueous process waste,
including purge water from storage basins for irradiated reactor fuel or fuel elements). The high-level
waste is stored to permit the decay of short-lived radionuclides and allow separation of solids (sludge)
from soluble waste. Evaporators concentrate soluble waste to reduce original volumes and to
immobilize it as crystallized salt by successive evaporations of the liquid supernate. The SRS treats
the evaporator overheads in cesium removal columns before transferring them to the F- and H-Area
Effluent Treatment Facility. The SRS processes the sludge and salt to prepare them for vitrification at
the Defense Waste Processing Facility (high-level waste), when it becomes operational, or stabilization
at the Z-Area Saltstone Facility (low-level waste). DOE has prepared a Supplemental EIS related to
Defense Waste Processing Facility operations (DOE 1994d).
By December 31, 1991, DOE had stored approximately 127.9 million liters (33.8 million gallons)
of high-level radioactive waste on the Site. Estimates of current tank capacity and high-level waste
forecasts should be available in 1995. In general, however, due to a number of factors, the most
important of which has been the extended outage of the evaporators, the estimated inventory of waste
in the high-level tanks is greater than 90 percent of existing capacity (WSRC 1994d). DOE is
constructing a replacement high-level waste tank evaporator to augment or replace existing
evaporators.
Figure 4-17. Flow diagram for high-level radioactive waste. Figure 4-18. Flow diagram for waste handling at the SRS. Table 4-25. Average annual waste generation forecast for Savannah River Site (cubic meters). ,b
Waste Type FY94 FY95 FY96
Transuranic 670 860 760
Low-Level
Low-Activity 21,350 17,680 17,970
Intermediate-Level 940 580 740
Hazardous 140 130 100
Mixed 120 130 110
a. Source: WSRC (1994c).
b. To convert cubic meters to cubic feet, multiply by 35.314.
4.14.2 Transuranic Waste
At present, DOE uses three methods of retrievable storage for transuranic waste at SRS, based on
the time of generation. Transuranic waste generated before 1974 is buried in approximately
120 belowgrade concrete culverts in the Solid Waste Disposal Facility. Transuranic waste generated
from 1974 to 1985 is stored on five concrete pads and one asphalt pad that have been covered with
approximately 1.2 meters (4 feet) of native soil. DOE stores waste generated since 1985 on
13 additional concrete pads that are not covered with soil. Pads 1 through 17 operate under Interim
Status approved by the South Carolina Department of Health and Environmental Control. DOE uses
Pads 18 through 19, which are not required to have interim status, to manage nonhazardous transuranic
wastes only.
The SRS stores wastes containing 10 to 100 nanocuries per gram of transuranic material with
transuranic waste until it can complete Site-specific radiological performance assessments, which will
provide disposal limits for transuranic isotopes. SRS transuranic waste inventories and forecasts
include both transuranic waste and the 10- to 100-nanocuries-per-gram transuranic wastes.
At the end of 1993, the SRS had approximately 9,900 cubic meters (350,000 cubic feet) of
transuranic waste in storage (WSRC 1994e). Based on the 1994-to-1996 average annual generation
rate forecast, the Site generates approximately 760 cubic meters (27,000 cubic feet) of transuranic
waste annually. Transuranic mixed waste (transuranic and hazardous combined) accounts for
approximately 110 cubic meters (3,900 cubic feet) of this volume (WSRC 1994c). DOE is evaluating
available storage space for transuranic mixed waste to alleviate any storage capacity deficit.
4.14.3 Mixed Low-Level Waste
The SRS mixed waste program consists primarily of providing safe storage until treatment and
disposal facilities are available. The current volume of mixed low-level waste at the SRS is
1,700 cubic meters (60,000 cubic feet) (WSRC 1994e). Based on the 1994-to-1996 average annual
generation forecast, the Site generates approximately 118 cubic meters (4,170 cubic feet) of mixed
low-level waste annually (WSRC 1994c). DOE is evaluating available storage space to determine
when the SRS will exceed its capacity. However, DOE is constructing a Consolidated Incineration
Facility in H-Area, which will treat mixed, hazardous, and low-level waste. When the incinerator is
operational, existing inventory will be reduced and more storage capacity will become available.
4.14.4 Low-Level Waste
The SRS packages low-level waste for disposal on the Site in accordance with the waste category
and its estimated surface dose rate. The Site places low-activity waste in carbon steel boxes and
deposits it in an Engineered Low-Level Trench (ELLT). The trenches are several acres in size by
6 meters (20 feet) deep and have sloped sides and floor, allowing drainage to a collection sump.
When the trenches are full, DOE backfills and covers them with at least 1.8 meters (6 feet) of soil.
The Site packages intermediate-level wastes according to the waste form and disposes of them in slit
trenches. DOE will store long-lived wastes, such as resins, until the Long-Lived Waste Storage
Building, currently under construction, becomes operational. This building will provide storage until
DOE develops treatment and disposal technologies.
The SRS is developing a new disposal facility, known as the E-Area Vault (EAV). This facility
will include vaults for low-activity waste, intermediate-level non-tritium waste, and intermediate-level
tritium waste.
Based on the 1994-to-1996 average annual generation forecast, the Site generates approximately
19,000 cubic meters (671,400 cubic feet) of low-activity waste and 750 cubic meters (26,600 cubic
feet) of intermediate-level waste annually. DOE expects that the Consolidated Incineration Facility
will begin operations by the second quarter of Fiscal Year 1996; this facility will have the capability
of annually processing as much as 15,850 cubic meters (560,000 cubic feet) of boxed low-activity
waste and approximately 186 cubic meters (6,600 cubic feet) of hazardous and mixed waste.
4.14.5 Hazardous Waste
DOE stores hazardous wastes generated at various SRS facilities in buildings in the B- and
N-Areas, and on the Solid Waste Storage Pads. The Resource Conservation and Recovery Act
regulates these wastes.
The inventory of hazardous waste in storage at the SRS is about 1.6 million kilograms (3.6 million
pounds), occupying a volume of about 2,430 cubic meters (86,000 cubic feet) (WSRC 1994e). Based
on the 1994-to-1996 average annual generation rate forecast, the Site generates approximately
124 cubic meters (4,370 cubic feet) of hazardous waste annually (WSRC 1994c).
4.14.6 Sanitary Waste
The SRS disposes of most of its solid sanitary waste in onsite landfills, the most recent of which
began operation in 1985. Current disposal operations include the Interim Sanitary Landfill. About
30 trucks per work day arrive at this facility carrying approximately 18,125 kilograms (40,000 pounds)
of waste that, after compaction, occupies approximately 115 cubic meters (150 cubic yards) of landfill
space. The recent implementation of SRS paper and aluminum can recycling programs and disposal of
office waste off the Site in a commercial landfill has increased the projected life of the landfill to the
fourth quarter of 1996 (WSRC 1994e).
DOE also maintains an inert material landfill on the Site near Burma Road. This facility receives
demolition and construction debris. DOE is evaluating the construction of a new SRS sanitary landfill
or the use of a commercial landfill.
4.14.7 Hazardous Materials
The SRS 1993 Tier II emergency and hazardous chemical inventory lists 205 reportable hazardous
substances present on the Site in excess of the 10,000-pound (4,536-kilogram) threshold quantity
(WSRC 1994f). The number and the total weight of any hazardous chemicals used on the Site change
daily in response to use. The annual Superfund Amendments and Reauthorization Act (SARA) reports
for the SRS include listings of hazardous materials used or stored on the Site during each year.
5. ENVIRONMENTAL CONSEQUENCES
5.1 Overview
This chapter discusses the potential environmental consequences for each spent nuclear fuel
management alternative described in Chapter 3. The representative host site locations, as described in
Chapter 2, are the F- and H-Areas and an undeveloped site close to H-Area. These sites are
representative of available areas that could support spent fuel management missions. Based on generic
facility characteristics, this chapter analyzes representative consequences in terms of the environmental
attributes of the potential host areas and the Savannah River Site (SRS) at large, as described in
Chapter 4. Table 3-2 compares the environmental consequences of each alternative. The impacts
associated with the construction and operation of a Navy Expended Core Facility are not included in
this chapter, but are included in Appendix D of Volume 1 of this Environmental Impact Statement.
5.2 Land Use
Overall environmental impacts on land use by any of the alternatives would be small because the
U.S. Department of Energy (DOE) would construct most new facilities in F- and H-Areas, which are
already dedicated to industrial use and which previous activities have disturbed. New construction on
the undeveloped representative host site near H-Area would probably be necessary only for the
construction of a dry storage vault.
The Centralization Alternative (Alternative 5), under which DOE would transfer all spent nuclear
fuel to the SRS, would result in the greatest changes in land use. Under this alternative, the SRS
would dedicate between 70 and 100 acres (0.3 and 0.4 square kilometer) for use in spent nuclear fuel
management; the exact location and size of the area affected would depend on whether DOE chose to
use the wet storage, dry storage, or processing option. Of this affected area, a maximum of
approximately 100 acres (0.4 square kilometer) would change from managed pine forest to industrial
use.
DOE would retain under its control any lands supporting the spent nuclear fuel management
program for the life of the project. No alternative would require the acquisition of public lands.
5.3 Socioeconomics
Socioeconomic consequences resulting from the implementation of any of the alternatives would
relate primarily to changes in employment within the region of influence (ROI). DOE has based the
analysis in the following section on estimated employment and population data for each SRS spent
nuclear fuel alternative, as listed in Table 5-1. The population within the region of influence in 1995
is estimated to be approximately 462,000. The labor force will be about 257,000 persons of which
about 242,000 will be employed.
DOE expects the employment level at the Site to decline from about 20,000 (in 1995) to about
15,800 (in 2004) as the SRS mission is redefined. This anticipated decline would be somewhat offset
by the jobs created by the spent nuclear fuel management activities. Therefore, none of the
alternatives would require additional operations employees because the SRS could fill all operational
positions through the reassignment of existing workers. Consequently, this analysis addresses only
employment impacts from construction activities. Given the natural variation in construction
employment levels, the analysis could not accurately determine the reassignment of existing
construction workers. As a result, this assessment analyzed the maximum potential impact, which
assumes that all construction employment would represent new jobs that in-migrating workers would
fill.
DOE estimated total employment impacts using the Regional Input-Output Modeling System that
the U.S. Bureau of Economic Analysis developed for the SRS region of influence. This assessment
also analyzed changes in population based on historic data that indicate that 90 percent of SRS
employees live in the six-county region.
5.3.1 Potential Impacts
Table 5-1 lists direct increases in construction employment for each alternative and the
corresponding change in population. As listed, potential impacts to socioeconomic resources would be
smallest under Alternative 1 (No Action) and would be greatest under Option 5b (Centralization - Wet
Storage). Therefore, Option 5b provides the bounding case for maximum potential impacts to
socioeconomic resources.
Table 5-1. Direct construction employment and total population changes by alternative, 1995-2004.
Alternative 1995a 1996a 1997a 1998a 1999a 2000 2001 2002 2003 2004
Alternative 1- 50 50 50 50 50 50 50 50 50 50
Employmenta 200 150 150 100 100 100 100 100 100 100
Population
Option 2a- 50 50 50 50 50 200 400 600 500 200
Employment 200 150 150 100 100 850 1,550 2,250 2,000 750
Population
Option 2b- 50 50 50 50 50 200 400 600 500 200
Employment 100 150 150 100 100 850 1,550 2,250 2,000 750
Population
Option 2c- 50 50 50 50 50 200 350 550 500 150
Employment 200 150 150 100 100 700 1,350 2,050 1,850 600
Population
Option 3a- 50 50 50 50 50 200 400 600 500 200
Employment 200 150 150 100 100 850 1,550 2,250 2,000 750
Population
Option 3b- 50 50 50 50 50 200 400 650 600 250
Employment 200 150 150 100 100 800 1,600 2,550 2,400 900
Population
Option 3c- 50 50 50 50 50 200 350 550 500 150
Employment 200 150 150 100 100 700 1,350 2,050 1,850 600
Population
Option 4a- 50 50 50 50 50 200 400 650 600 250
Employment 200 150 150 100 100 800 1,600 2,550 2,400 900
Population
Option 4b- 50 50 50 50 50 200 400 650 600 250
Employment 200 150 150 100 100 800 1,600 2,550 2,400 900
Population
Option 4c- 50 50 50 50 50 200 350 550 500 150
Employment 200 150 150 100 100 700 1,350 2,050 1,850 600
Population
Option 4d- 50 50 50 50 50 300 500 700 650 250
Employment 200 200 150 150 150 1,100 1,900 2,800 2,500 900
Population
Option 4e- 50 50 50 50 50 250 500 800 800 300
Employment 200 200 150 150 150 1,000 2,000 3,200 3,000 1,100
Population
Option 4f- 50 50 50 50 50 200 450 650 600 200
Employment 200 200 150 150 150 850 1,700 2,550 2,350 700
Population
Option 4g- 50 50 50 50 50 100 150 200 100 100
Employment 200 150 150 100 100 250 500 700 450 300
Population
Alternative 1995a 1996a 1997a 1998a 1999a 2000 2001 2002 2003 2004
Option 5a- 50 50 50 50 50 900 1,750 2,550 2,500 2,450
Employment 200 150 150 100 100 3,500 6,800 9,900 9,700 9,450
Population
Option 5b- 50 50 50 50 50 1,000 1,900 2,700 2,650 2,600
Employment 200 150 150 100 100 3,850 7,450 10,550 10,350 10,100
Population
Option 5c- 50 50 50 50 50 900 1,750 2,550 2,500 2,450
Employment 200 150 150 100 100 3,500 6,800 9,900 9,700 9,500
Population
Option 5d- 50 50 50 50 50 100 150 200 100 100
Employment 200 150 150 100 100 250 500 700 450 300
Population
a. Construction is related to renovation of reactor basin and Receiving Basin for Offsite Fuels.
Table 5-2 lists indirect employment and corresponding population changes associated with
construction phase activities under Option 5b. As listed, the number of full-time construction workers
required to support the implementation of this option from 1995 to 2004 would range from
approximately 50 to 2,700. When added to the indirect employment of 1,600 jobs in the peak year
(2002), the total employment impact in the region would be approximately 4,300 employees.
Table 5-2. Estimated increases in employment and population related to construction activities for
Option 5b, from 1995 to 2004. ROI refers to the six-county region of influence.
Factor 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004
Direct 50 50 50 50 50 1,000 1,900 2,700 2,650 2,600
employment
Secondary 30 30 30 30 30 600 1,100 1,600 1,550 1,500
employment
Total employment 80 80 80 80 80 1,600 3,000 4,300 4,200 4,100
change
% Change in ROI 0.03 0.03 0.03 0.03 0.03 0.54 1.00 1.41 1.36 1.32
labor force
% Change in ROI 0.03 0.03 0.03 0.03 0.03 0.57 1.06 1.50 1.45 1.40
employment
Population change 200 150 150 100 100 3,850 7,450 10,550 10,350 10,100
(in region)
% Change in ROI 0.04 0.03 0.03 0.02 0.02 0.81 1.56 2.21 2.16 2.11
population
Assuming in-migrating workers filled all jobs, the regional labor force and employment would increase
by 1.4 percent and 1.5 percent, respectively. These changes would be temporary and would have no
adverse impact on the region. After 2004, employment would gradually decline to a relatively
constant level of about 50 jobs.
Based on historic data, approximately 90 percent of new employees would live within the
six-county region of influence. Assuming each new employee represented one household with 2.72
persons per household, there would be approximately 10,550 additional people in the region during the
peak year (2002). These changes would be temporary and would represent an estimated 2.2 percent
increase in baseline population levels. Given this minor change in population, DOE expects potential
impacts on the demand for community resources and services such as housing, schools, police, health
care, and fire protection to be negligible.
Because all the other alternatives would require fewer employees, they would result in smaller
changes than those listed in Table 5-2, and would have no adverse impacts on socioeconomic
resources in the region of influence.
5.4 Cultural Resources
A Programmatic Memorandum of Agreement (SRARP 1989) between the DOE Savannah River
Operations Office, the South Carolina State Historic Preservation Office, and the Advisory Council on
Historic Preservation, ratified on August 24, 1990, is the instrument for the management of cultural
resources at the SRS. DOE uses this memorandum to identify cultural resources, assess them in terms
of eligibility for the National Register of Historic Places, and develop mitigation plans for affected
resources in consultation with the State Historic Preservation Officer. DOE would comply with the
terms of the memorandum for all activities needed to support spent nuclear fuel management actions.
The potential for adverse impacts on cultural resources would be smallest under Alternative 1
(No Action) and would be greatest under Alternative 5 (Centralization). Any facilities that DOE
would construct in F- and H-Areas, north of Road E (Alternatives 1-5), would be in Sensitivity
Zones 2 and 3. Section 4.4 describes these zones. The undeveloped representative host site south and
east of H-Area (Alternative 5) is in Sensitivity Zone 3. Although there are no known archeological
sites in the area, it has never been surveyed. Surveying being conducted near F-Area (north of
Road C and west of Road 4 along Upper Three Runs Creek) has recorded some historic and
prehistoric sites. However, DOE expects no impacts in F- and H-Areas due to their extensive
industrial development. Until DOE has determined the precise locations of facilities connected with
any of the alternatives, it cannot predict impacts on cultural resources in the undeveloped site area
(Sassaman 1994). However, DOE would mitigate, through avoidance or removal, impacts to
potentially significant resources that future site surveys might discover.
5.5 Aesthetic and Scenic Resources
None of the alternatives for spent nuclear fuel management at the SRS would have adverse
consequences on scenic resources or aesthetics. Most new construction would be in F- or H-Area,
both of which are already dedicated to industrial use. New construction on the undeveloped site,
which would occur primarily under Alternative 5, would be adjacent to H-Area in an already heavily
industrialized portion of the SRS. In all cases, new construction would not be visible off the Site or
from public access roads on the Site. No alternative would produce emissions to the atmosphere that
would be visible or would indirectly reduce visibility.
5.6 Geologic Resources
The SRS contains no unique geologic features or minerals of economic value. Therefore, DOE
anticipates no impacts to geologic resources at the SRS from any of the spent nuclear fuel
management alternatives.
Other sections in this chapter consider the relationships of the Site's specific geology and the
region's historic and analyzed seismicity to the local environment and to SRS spent nuclear fuel-
related structures and facilities. Section 5.8 discusses the consequences of analyzed seismic events on
both surface-water and groundwater resources. Section 5.15 describes estimates of risk that consider
both the probability of and the consequences from a wide range of seismic events, ranging from local
and regional historically documented earthquakes to postulated lower probability, higher consequence
events.
The accident analyses in this chapter, which DOE based on information from approved safety
analysis reports for applicable facilities, address the frequency and consequences of historic
earthquakes, as well as postulated less likely, but more damaging, seismic events. DOE has evaluated
the consequences from seismic challenges to the facilities and structures up to 0.20g lateral ground
acceleration.
5.7 Air Quality Consequences
The SRS is in compliance with both Federal and state ambient air quality standards for criteria
and toxic air pollutants. As shown in the following tables, the predicted incremental air pollutant
impacts would not contribute to exceeding either the National Ambient Air Quality Standards or South
Carolina's Ambient Air Quality Standards.
DOE performed analyses using computer models in order to assess the potential air quality
impacts of operations under each of the spent nuclear fuel management alternatives. This section
describes the results of these analyses. All the concentrations discussed below are ground-level
estimations based on results from the ISC2 and FDM models for nonradiological pollutants, and
MAXIGASP- and POPGASP SRS-climatology-specific models for radionuclides. The analyses
assume that facility operations would result in both radiological and nonradiological emissions. DOE
assessed construction impacts qualitatively in relation to the land area to be disturbed under each
alternative.
Nonradiological Emissions. DOE analyzed the potential incremental impacts of only those
substances for which it expects releases to the atmosphere during the normal operation of spent nuclear
fuel facilities. The nonradiological releases evaluated for each alternative include seven criteria
pollutants and 23 toxic pollutants. DOE selected the toxic substances for analysis by comparing the
anticipated chemical usage at the proposed spent nuclear fuel facilities to the list of 257 toxic air
pollutants in the South Carolina Air Pollution Regulations (SCDHEC 1976). The SRS modeled
potential emissions of the listed toxic chemicals that DOE anticipates would be used during spent
nuclear fuel activities. The following subsections discuss the results for both criteria and toxic
pollutants. Tables 5-3 and 5-4 list the estimated maximum incremental concentrations of these
pollutants at the Site boundary, while Tables 5-5 and 5-6 contain the incremental rates of release.
Radiological Emissions. DOE evaluated the potential radiological releases to the atmosphere
from spent fuel management at the SRS using existing Site historical operations information. Based
on the actual 1993 emissions data from the Receiving Basin for Offsite Fuels (WSRC 1994d), DOE
estimates that emissions from any of the wet storage options under Alternatives 1 through 4 would
Table 5-3. Estimated incremental air quality impacts at the Savannah River Site boundary from operations of spent nuclear fuel alternatives -
criteria pollutants (-g/m3).
Incremental Concentrations from Alternatives
Maximum
Averaging Regulatory Potential Actual
Pollutantb Time Standardc Concentration Concentratione
No
Action Decentralization 1992/1993 Planning Basis
1 2a 2b 2c 3a 3b 3c
CRITERIA POLLUTANTS (-g/m3)
Carbon monoxide 8-hour 10,000 818 23 <0.01 0.1 0.1 4.3 0.1 0.1 4.3
1-hour 40,000 3,553 180 <0.01 0.8 0.8 32 0.8 0.8 32
Ozone (as VOC) 1-hour 245 N/Ad N/Ad 1.6 0.3 0.3 2.6 0.3 0.3 2.6
Nitrogen oxides Annual 100 30 4 <0.01 0.01 <0.01 11.00 <0.01 <0.01 11.0
geometric
mean
Particulate matter Annual 50 9 3 - - - <0.01 - - 0.01
(<10-m) 24-hour 150 93 56 - - - 0.40 - - 0.40
Total suspended Annual 75 20 11 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
particulates (TSP)
Sulfur dioxide Annual 80 18 10 - <0.01 <0.01 0.01 <0.01 <0.01 0.01
24-hour 365 356 185 - 0.01 0.01 0.43 0.01 0.01 0.43
3-hour 1,300 1,210 634 - 0.05 0.05 3.2 0.05 0.05 3.2
Lead Calendar 1.5 <0.01 <0.01 - - - - - - -
quarter mean
Gaseous Fluorides (as 1-month 0.8 0.11 0.03 - - - 0.02 - - 0.02
HF) 1-week 1.6 0.6 0.15 - - - 0.10 - - 0.10
24-hour 2.9 1.20 0.31 - - - 0.20 - - 0.20
12-hour 3.7 2.40 0.62 - - - 0.40 - - 0.40
Table 5-3. (continued).
Incremental Concentrations from Alternatives
Maximum
Averaging Regulatory Potential Actual
Pollutantb Time Standardc Concentration Concentratione
Regionalization A Regionalization B
4a 4b 4c 4d 4e 4f 4g
CRITERIA POLLUTANTS (-g/m3)
Carbon monoxide 8-hour 10,000 818 23 0.2 0.2 4.3 0.2 0.2 5.5 -
1-hour 40,000 3,553 180 1.2 1.2 32 1.5 1.5 41 -
Ozone (as VOC) 1-hour 245 N/Ad N/Ad 0.5 0.5 2.6 0.6 0.6 3.3 1.4
Nitrogen oxides Annual 100 30 4 <0.01 <0.01 11 <0.01 <0.01 14 -
geometric
mean
Particulate matter Annual 50 9 3 - - 0.01 - - 0.01 -
(<10-m) 24-hour 150 93 56 - - 0.4 - - 0.5 -
Total suspended Annual 75 20 11 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 -
particulates (TSP)
Sulfur dioxide Annual 80 18 10 <0.01 <0.01 0.01 <0.01 <0.01 0.01 -
24-hour 365 356 185 0.02 0.02 0.43 0.02 0.02 0.55 -
3-hour 1,300 1,210 634 0.09 0.09 3.2 0.11 0.11 4.1 -
Lead Calendar 1.5 <0.01 <0.01 - - - - - - -
quarter mean
Gaseous Fluorides 1-month 0.8 0.11 0.03 - - 0.02 - - 0.02 -
(as HF) 1-week 1.6 0.6 0.15 - - 0.10 - - 0.13 -
24-hour 2.9 1.20 0.31 - - 0.20 - - 0.25 -
12-hour 3.7 2.40 0.62 - - 0.40 - - 0.51 -
Table 5-3. (continued).
Incremental Concentrations from Alternatives
Averaging Regulatory Maximum Actual Centralization
Time Standardc Potential Concentratione
Concentration
5a 5b 5c 5d
CRITERIA POLLUTANTS (-g/m3)
Carbon monoxide 8-hour 10,000 818 23 1.0 1.0 5.1 -
1-hour 40,000 3,553 180 6.7 6.7 37 -
Ozone (as VOC) 1-hour 245 N/Ad N/Ad 1.4 1.4 3.1 1.4
Nitrogen oxides Annual 100 30 4 0.04 0.04 11.1 -
geometric
mean
Particulate matter Annual 50 9 3 - - 0.01 -
(<10-m) 24-hour 150 93 56 - - 0.40 -
Total suspended particulates (TSP) Annual 75 20 11 <0.01 <0.01 <0.01 -
Sulfur dioxide Annual 80 18 10 <0.01 <0.01 0.02 -
24-hour 365 356 185 0.09 0.09 0.49 -
3-hour 1,300 1,210 634 0.50 0.50 3.5 -
Lead Calendar 1.5 <0.01 <0.01 - - - -
quarter mean
Gaseous Fluorides (as HF) 1-month 0.8 0.11 0.03 - - 0.02 -
1-week 1.6 0.6 0.15 - - 0.10 -
24-hour 2.9 1.20 0.31 - - 0.10 -
12-hour 3.7 2.40 0.62 - - 0.40 -
- = No impact.
a. Maximum modeled ground-level concentration at SRS perimeter unless higher offsite concentrations are otherwise specified.
b. Major pollutants of concern regarding spent nuclear fuel management activities.
c. Most stringent Federal and state regulatory standards (CFR 1991a), (SCDHEC 1976).
d. Measurement data currently unavailable.
e. Maximum operational air pollutant emissions projected for baseline year 1995. Concentration estimates based on actual emissions from all SRS sources for calendar year 1990
plus maximum potential emissions for sources permitted through December 1992.
Table 5-4. Estimated incremental air quality impacts at the Savannah River Site boundary from operations of spent nuclear fuel alternatives -
toxic pollutants (-g/m3).
Incremental Concentrations from Alternatives
Maximum
Averaging Regulatory Potential Actual
Pollutantb Time Standardc Concentration Concentrationd
No
Action Decentralization 1992/1993 Planning Basis
1 2a 2b 2c 3a 3b 3c
TOXIC POLLUTANTS (-g/m3)
Nitric acid 24-hour 125 51 6.7 - - - <0.01 - - <0.01
1,1,1,- Trichloroethane 24-hour 9,550 81 22 <0.01 <0.01 <0.01 0.01 <0.01 <0.01 0.01
Benzene 24-hour 150 32 31 - - - 0.04 - - 0.04
Ethanolamine 24-hour 200 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Ethyl benzene 24-hour 4,350 0.58 0.12 - - - <0.01 - - <0.01
Ethylene glycol 24-hour 650 0.20 0.08 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Formaldehyde 24-hour 7.5 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Glycol ethers 24-hour + <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Hexachloronapthalene 24-hour 1.0 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Hexane 24-hour 200 0.21 0.07 <0.01 <0.01 <0.01 0.04 <0.01 <0.01 0.04
Manganese 24-hour 25 0.82 0.10 - - - <0.01 - - <0.01
Methyl alcohol 24-hour 1,310 2.9 0.51 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Methyl ethyl ketone 24-hour 14,750 6.0 0.99 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Methyl isobutyl ketone 24-hour 2,050 3.0 0.51 - - - <0.01 - - <0.01
Methylene chloride 24-hour 515 10.5 1.8 - - - 0.02 - - 0.02
Naphthalene 24-hour 1,250 0.01 0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Phenol 24-hour 190 0.03 0.03 - - - <0.01 - - <0.01
Phosphorus 24-hour 0.5 <0.001 <0.001 - - - <0.001 - - <0.001
Sodium hydroxide 24-hour 20 0.01 0.01 - - - <0.01 - - <0.01
Toluene 24-hour 2,000 9.3 1.6 <0.01 <0.01 <0.01 0.04 <0.01 <0.01 0.04
Trichloroethylene 24-hour 6,750 4.8 1.0 - - - <0.01 - - <0.01
Vinyl acetate 24-hour 176 0.06 0.02 - - - <0.01 - - <0.01
Xylene 24-hour 4,350 39 3.8 0.01 0.01 0.01 0.05 0.01 0.01 0.05
Table 5-4. (continued).
Pollutantb Averaging Regulatory Maximum Actual Incremental Concentrations from Alternatives
Time Standardc Potential Concentrationd
Concentration
Regionalization A Regionalization B
4a 4b 4c 4d 4e 4f 4g
TOXIC POLLUTANTS (-g/m3)
Nitric acid 24-hour 125 51 6.7 - - 1.0 - - 1.3 -
1,1,1,- Trichloroethane 24-hour 9,550 81 22 <0.01 <0.01 0.01 <0.01 <0.01 0.01 <0.01
Benzene 24-hour 150 32 31 - - 0.04 - - 0.05 -
Ethanolamine 24-hour 200 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Ethyl benzene 24-hour 4,350 0.58 0.12 - - <0.01 - - <0.01 -
Ethylene glycol 24-hour 650 0.20 0.08 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Formaldehyde 24-hour 7.5 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Glycol ethers 24-hour + <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Hexachloronapthalene 24-hour 1.0 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Hexane 24-hour 200 0.21 0.07 <0.01 <0.01 0.04 <0.01 <0.01 0.05 <0.01
Manganese 24-hour 25 0.82 0.10 - - <0.01 - - <0.01 -
Methyl alcohol 24-hour 1,310 2.9 0.51 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Methyl ethyl ketone 24-hour 14,750 6.0 0.99 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Methyl isobutyl ketone 24-hour 2,050 3.0 0.51 - - <0.01 - - <0.01 -
Methylene chloride 24-hour 515 10.5 1.8 - - 0.02 - - 0.02 -
Naphthalene 24-hour 1,250 0.01 0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Phenol 24-hour 190 0.03 0.03 - - <0.01 - - <0.01 -
Phosphorus 24-hour 0.5 <0.001 <0.001 - - <0.001 - - <0.001 -
Sodium hydroxide 24-hour 20 0.01 0.01 - - <0.01 - - <0.01 -
Toluene 24-hour 2,000 9.3 1.6 <0.01 <0.01 0.04 <0.01 <0.01 <0.05 <0.01
Trichloroethylene 24-hour 6,750 4.8 1.0 - - <0.01 - - <0.01 -
Vinyl acetate 24-hour 176 0.06 0.02 - - <0.01 - - <0.01 -
Xylene 24-hour 4,350 39 3.8 0.01 0.01 0.05 0.01 0.01 0.06 0.01
Table 5-4. (continued).
Incremental Concentrations from Alternatives
Pollutantb Averaging Regulatory Maximum Actual Centralization
Time Standardc Potential Concentrationd
Concentration
5a 5b 5c 5d
TOXIC POLLUTANTS (-g/m3)
Nitric acid 24-hour 125 51 6.7 - - 1.0 -
1,1,1,- Trichloroethane 24-hour 9,550 81 22 <0.01 <0.01 0.01 <0.01
Benzene 24-hour 150 32 31 - - 0.04 -
Ethanolamine 24-hour 200 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Ethyl benzene 24-hour 4,350 0.58 0.12 - - <0.01 -
Ethylene glycol 24-hour 650 0.20 0.08 <0.01 <0.01 <0.01 <0.01
Formaldehyde 24-hour 7.5 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Glycol ethers 24-hour + <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Hexachloronapthalene 24-hour 1.0 <0.01 <0.01 <0.01 <0.01 <0.01 <0.01
Hexane 24-hour 200 0.21 0.07 <0.01 <0.01 0.04 <0.01
Manganese 24-hour 25 0.82 0.10 - - <0.01 -
Methyl alcohol 24-hour 1,310 2.9 0.51 <0.01 <0.01 <0.01 <0.01
Methyl ethyl ketone 24-hour 14,750 6.0 0.99 <0.01 <0.01 <0.01 <0.01
Methyl isobutyl ketone 24-hour 2,050 3.0 0.51 - - <0.01 -
Methylene chloride 24-hour 515 10.5 1.8 - - 0.02 -
Naphthalene 24-hour 1,250 0.01 0.01 <0.01 <0.01 <0.01 <0.01
Phenol 24-hour 190 0.03 0.03 - - <0.01 -
Phosphorus 24-hour 0.5 <0.001 <0.001 - - <0.001 -
Sodium hydroxide 24-hour 20 0.01 0.01 - - <0.01 -
Toluene 24-hour 2,000 9.3 1.6 <0.01 <0.01 0.04 <0.01
Trichloroethylene 24-hour 6,750 4.8 1.0 - - <0.01 -
Vinyl acetate 24-hour 176 0.06 0.02 - - <0.01 -
Xylene 24-hour 4,350 39 3.8 0.01 0.01 0.05 0.01
- No impact.
+ Not available.
a. Maximum modeled ground-level concentration at SRS perimeter unless higher offsite concentrations are otherwise specified.
b. Major pollutants of concern regarding spent nuclear fuel.
c. Most stringent Federal and state regulatory standards (CFR 1991a), (SCDHEC 1976).
d. Maximum operational air pollutant emissions projected for baseline year 1995. Concentration estimates based on actual emissions from all SRS sources for calendar year
1990 plus maximum potential emissions for sources permitted through December 1992.
Table 5-5. Incremental air quality pollutant emission rates related to spent nuclear fuel alternatives - criteria pollutants.
Baseline Alternatives
Pollutant
Maximum No
Design Action Decentralization 1992/1993 Planning Basis
Capacity Actualb
1 2a 2b 2c 3a 3b 3c
CRITERIA POLLUTANTS (TONS PER YEAR)
NOx 2.22x104 2.62x103 - 6.0x100 6.0x100 2.0x104 6.0x100 6.0x100 2.0x104
Particulates
TSP 3.62x103 9.80x102 - 4.0x10-1 4.0x10-1 1.5x101 4.0x10-1 4.0x10-1 1.5x101
PM10 2.66x103 4.97x102 - 2.6x10-1 2.6x10-1 9.3x100 2.6x10-1 2.6x10-1 9.3x100
CO 6.77x103 1.99x102 - 1.5x100 1.5x100 3.8x101 1.5x100 1.5x100 3.8x101
SO2 6.42x104 6.68x103 1.6x10-3 4.0x10-1 4.0x10-1 1.2x101 4.0x10-1 4.0x10-1 1.2x101
Gaseous Fluorides 2.14x10-2 1.07x10-2 - - - 2.4x101 - - 2.4x101
Ozone (as VOC) N/Ac N/Ac - 6.0x10-1 6.0x10-1 1.8x10-1 6.0x10-1 6.0x10-1 1.8x10-1
CRITERIA POLLUTANTS (TONS PER YEAR) Regionalization A Regionalization B
4a 4b 4c 4d 4e 4f 4g
NOx 2.22x104 2.62x103 8.5x100 8.5x100 2.0x104 1.1x101 1.1x101 2.5x104 -
Particulates
TSP 3.62x103 9.80x102 6.0x10-2 6.0x10-2 1.5x101 7.6x10-2 7.6x10-2 1.5x101 -
PM10 2.66x103 4.97x102 1.45x101 1.45x101 9.3x100 1.8x101 1.8x101 9.3x100 -
CO 6.77x103 1.99x102 2.0x100 2.0x100 3.8x101 2.5x100 2.5x100 5.2x101 -
SO2 6.42x104 6.68x103 5.5x10-2 5.5x10-2 1.3x101 7.6x10-2 7.6x10-2 1.7x101 -
Gaseous Fluorides 2.14x10-2 1.07x10-2 - - 2.4x101 - - 3.0x101 -
Ozone (as VOC) N/Ac N/Ac 8.5x10-1 8.5x10-1 1.8x10-1 1.1x100 1.1x100 2.3x10-1 -
Table 5-5. (continued).
Pollutant Maximum Actualb Alternatives
Design
Capacity
Centralization
CRITERIA POLLUTANTS (TONS PER YEAR) 5a 5b 5c 5d
NOx 2.2x104 2.6x103 5.6x101 5.6x101 2.0x104 -
Particulates
TSP 3.62x103 9.8x102 2.1x100 2.1x100 1.8x101 -
PM10 2.66x103 4.97x102 1.4x100 1.4x100 9.3x100 -
CO 6.77x103 1.99x102 2.7x101 2.7x101 6.9x101 -
SO2 6.42x104 6.68x103 8.1x100 8.1x100 2.0x101 -
Gaseous Fluorides 2.14x10-2 1.07x10-2 2.4x101 -
Ozone (as VOC) N/Ac N/Ac 4.6x100 4.6x100 2.4x101 -
a. Source: WSRC (1994a).
b. Maximum operational air pollutant emissions projected for baseline year 1995. Concentration estimates based on actual emissions from all SRS sources for calendar
year 1990 plus maximum potential emissions for sources permitted through December 1992.
c. Emissions data currently unavailable.
- No proposed incremental emissions.
Table 5-6. Incremental air quality pollutant emission rates related to spent nuclear fuel alternatives - toxic pollutants.
Baseline Alternatives
Pollutant
Maximum No
Design Action Decentralization 1992/1993 Planning Basis
Capacity Actualb
1 2a 2b 2c 3a 3b 3c
TOXIC POLLUTANTS (TONS PER YEAR)
Nitric Acid 1.13x103 2.56x100 5.1x10-2 5.1x10-2 5.1x10-2 1.24x102 5.1x10-2 5.1x10-2 1.24x102
1,1,1-Trichloroethane 8.0x101 NAc - - - 7.02x10-1 - - 7.02x10-1
Benzene 2.9x101 4.48x100 - - - 8.02x10-1 - - 8.02x10-1
Ethanolamine 2.21x10-2 5.35x10-3 1.46x10-3 1.46x10-3 1.46x10-3 1.46x10-3 1.46x10-3 1.46x10-3 1.46x10-3
Ethyl Benzene 2.56x100 1.07x100 - - - 8.02x10-4 - - 8.02x10-4
Ethylene Glycol 6.83x10-1 4.17x10-1 2.25x10-2 2.25x10-2 2.25x10-2 4.27x10-2 2.25x10-2 2.25x10-2 4.27x10-2
Formaldehyde 4.55x10-2 4.8x10-4 3.6x10-6 3.6x10-6 3.6x10-6 3.6x10-6 3.6x10-6 3.6x10-6 3.6x10-6
Glycol Ethers 4.36x10-3 1.99x10-4 4.06x10-3 4.06x10-3 4.06x10-3 4.06x10-3 4.06x10-3 4.06x10-3 4.06x10-3
Hexachloronaphthalene <0.01 NAc 3.65x10-5 3.65x10-5 3.65x10-5 3.6x10-5 3.65x10-5 3.65x10-5 3.6x10-5
Hexane 3.54x100 2.22x10-1 3.28x10-3 3.28x10-3 3.28x10-3 8.13x10-1 3.28x10-3 3.28x10-3 8.13x10-1
Manganese 2.84x10-1 3.43x10-1 - - - 1.51x10-2 - - 1.51x10-2
Methyl Alcohol 6.62x10-1 3.46x10-1 6.84x10-2 6.84x10-2 6.84x10-2 8.68x10-2 6.84x10-2 6.84x10-2 8.68x10-2
Methyl Ethyl Ketone 6.41x100 3.17x100 2.19x10-3 2.19x10-3 2.19x10-3 3.47x10-2 2.19x10-3 2.19x10-3 3.47x10-2
Methyl Isobutyl Ketone 8.25x100 2.25x100 - - - 1.27x10-2 - - 1.27x10-2
Methylene Chloride 1.53x100 1.19x100 - - - 8.23x10-1 - - 8.23x10-1
Naphthalene 7.22x10-2 3.08x10-2 5.84x10-4 5.84x10-4 5.84x10-4 6.08x10-4 5.84x10-4 5.84x10-4 6.08x10-4
Phenol 8.07x10-2 1.37x10-2 - - - 6.01x10-5 - - 6.01x10-5
Phosphorus 2.97x10-3 1.65x10-4 - - - 1.6x10-6 - - 1.6x10-6
Sodium Hydroxide 1.26x10-1 1.26x10-1 - - - 5.97x10-2 - - 5.97x10-2
Toluene 3.91x100 7.66x10-1 5.0x10-2 5.0x10-2 5.0x10-2 9.2x10-1 5.0x10-2 5.0x10-2 9.2x10-1
Trichloroethylene 2.52x101 9.8x100 - - - 5.52x10-4 - - 5.52x10-4
Vinyl Acetate 4.38x10-2 5.9x10-3 - - - 5.0x10-5 - - 5.0x10-5
Xylene 1.46x103 1.22x101 1.58x10-1 1.58x10-1 1.58x10-1 1.4x100 1.58x10-1 1.58x10-1 1.4x100
Table 5-6. (continued).
Baseline Alternatives
Pollutant
Maximum Actualb Regionalization A Regionalization B
Design
Capacity
4a 4b 4c 4d 4e 4f 4g
TOXIC POLLUTANTS (TONS PER YEAR)
Nitric Acid 1.1x103 2.6x100 5.1x10-2 5.1x10-2 1.2x102 6.5x10-2 6.5x10-2 1.5x102 -
1,1,1-Trichloroethane 8.0x101 NAc - - 7.0x10-1 - - 8.9x10-1 -
Benzene 2.9x101 4.5x100 - - 8.0x10-1 - - 1.0x100 -
Ethanolamine 2.2x10-2 5.4x10-3 1.5x10-3 1.5x10-3 1.5x10-3 1.9x10-3 1.9x10-3 1.9x10-3 -
Ethyl Benzene 2.6x100 1.1x100 - - 8.0x10-4 - - 1.0x10-3 -
Ethylene Glycol 6.8x10-1 4.2x10-1 2.3x10-2 2.3x10-2 4.3x10-2 2.9x10-2 2.9x10-2 5.5x10-2 -
Formaldehyde 4.6x10-2 4.8x10-4 3.6x10-6 3.6x10-6 3.6x10-5 4.6x10-6 4.6x10-6 4.6x10-6 -
Glycol Ethers 4.4x10-3 2.0x10-4 4.1x10-3 4.1x10-3 4.1x10-3 5.2x10-3 5.2x10-3 5.2x10-3 -
Hexachloronapthalene <0.01 NAc 3.7x10-5 3.7x10-5 3.6x10-5 4.7x10-5 4.7x10-5 4.6x10-5 -
Hexane 3.5x100 2.2x10-1 3.3x10-3 3.3x10-3 8.1x10-1 4.2x10-3 4.2x10-3 1.0x100 -
Manganese 2.8x10-1 3.4x10-1 - - 1.5x10-2 - - 1.9x10-2 -
Methyl Alcohol 6.6x10-1 3.5x10-1 6.8x10-2 6.8x10-2 8.7x10-2 8.6x10-2 8.6x10-2 1.1x10-1 -
Methyl Ethyl Ketone 6.4x100 3.2x100 2.2x10-3 2.2x10-3 3.5x10-2 2.8x10-3 2.8x10-3 4.4x10-2 -
Methyl Isobutyl Ketone 8.3x100 2.3x100 - - 1.3x10-2 - - 1.7x10-2 -
Methylene Chloride 1.5x100 1.2x100 - - 8.2x10-1 - - 1.0x100 -
Naphthalene 7.2x10-2 3.1x10-2 5.8x10-4 5.8x10-4 6.1x10-4 7.4x10-4 7.4x10-4 7.7x10-4 -
Phenol 8.1x10-2 1.4x10-2 - - 6.0x10-5 - - 7.6x10-5 -
Phosphorus 3.0x10-3 1.7x10-4 - - 1.6x10-6 - - 2.0x10-6 -
Sodium Hydroxide 1.3x10-1 1.3x10-1 - - 6.0x10-2 - - 7.6x10-2 -
Toluene 3.9x100 7.7x10-1 5.0x10-2 5.0x10-2 9.2x10-1 6.4x10-2 6.4x10-2 1.2x100 -
Trichloroethylene 2.5x101 9.8x100 - - 5.5x10-4 - - 7.0x10-4 -
Vinyl Acetate 4.4x10-2 5.9x10-3 - - 5.0x10-5 - - 6.4x10-5 -
Xylene 1.5x103 1.2x101 1.6x10-1 1.6x10-1 1.4x100 2.0x10-1 2.0x10-1 1.8x100 -
Table 5-6. (continued).
Pollutant Maximum Actualb Alternatives
Design
Capacity
Centralization
5a 5b 5c 5d
TOXIC POLLUTANTS (TONS PER YEAR)
Nitric Acid 1.1x103 2.6x100 5.1x10-2 5.1x10-2 1.2x102 -
1,1,1-Trichloroethane 8.0x101 NAc - - 7.0x10-1 -
Benzene 2.9x101 4.5x100 - - 8.0x10-1 -
Ethanolamine 2.2x10-2 5.4x10-3 1.5x10-3 1.5x10-3 1.5x10-3 -
Ethyl Benzene 2.6x100 1.1x100 - - 8.0x10-4 -
Ethylene Glycol 6.8x10-1 4.2x10-1 2.3x10-2 2.3x10-2 4.3x10-2 -
Formaldehyde 4.6x10-2 4.8x10-4 3.6x10-6 3.6x10-6 3.6x10-6 -
Glycol Ethers 4.4x10-3 2.0x10-4 4.1x10-3 4.1x10-3 4.1x10-3 -
Hexachloronapthalene <0.01 NAc 3.7x10-5 3.7x10-5 3.6x10-5 -
Hexane 3.5x100 2.2x10-1 3.3x10-3 3.3x10-3 8.1x10-1 -
Manganese 2.8x10-1 3.4x10-1 - - 1.5x10-2 -
Methyl Alcohol 6.6x10-1 3.5x10-1 6.8x10-2 6.8x10-2 8.7x10-2 -
Methyl Ethyl Ketone 6.4x100 3.2x100 2.2x10-3 2.2x10-3 3.5x10-2 -
Methyl Isobutyl Ketone 8.3x100 2.3x100 - - 1.3x10-2 -
Methylene Chloride 1.5x100 1.2x100 - - 8.2x10-1 -
Naphthalene 7.2x10-2 3.1x10-2 5.8x10-4 5.8x10-4 6.1x10-4 -
Phenol 8.1x10-2 1.4x10-2 - - 6.0x10-5 -
Phosphorus 3.0x10-3 1.7x10-4 - - 1.6x10-6 -
Sodium Hydroxide 1.3x10-1 1.3x10-1 - - 6.0x10-2 -
Toluene 3.9x100 7.7x10-1 5.0x10-2 5.0x10-2 9.2x10-1 -
Trichloroethylene 2.5x101 9.8x100 - - 5.5x10-4 -
Vinyl Acetate 4.4x10-2 5.9x10-3 - - 5.0x10-5 -
Xylene 1.5x103 1.2x101 1.6x10-1 1.6x10-1 1.4x100 -
a. Source: WSRC (1994a).
b. Maximum operational air pollutant emissions projected for baseline year 1995. Concentration estimates based on actual emissions from all SRS sources for calendar
year 1990 plus maximum potential emissions for sources permitted through December 1992.
c. NA= Emissions data currently unavailable.
- No proposed incremental emissions.
consist of about 2 y 10-7 curies per year of cesium-137. Releases from dry storage activities under
these alternatives would be somewhat less. For Alternative 5 where SRS would manage about 2,740
MTHM (3,020 tons) of spent fuel (versus about 206 to 257 MTHM [227 to 283 tons] for the other
alternatives), the atmospheric releases of cesium-137 would be proportionally higher.
DOE used actual emissions from F- and H-Areas during 1985 and 1986, a period when the SRS
was processing material through the separations facilities at close to maximum capacity to evaluate
potential releases from spent nuclear fuel management activities. DOE believes that the isotopes
released during this period, and their emission rates, represent maximum emissions that could occur
under any of the alternatives (Table 5-7). The results of the analyses are presented in this section and
the human health consequences are discussed in Section 5.12. Section 5.15 presents the analysis of
the consequences of accidents.
Construction Emissions. Potential impacts to air quality from construction activities would
include fugitive dust from the clearing of land, as well as exhaust emissions from support equipment
(e.g., earth-moving vehicles, diesel generators). The amount of dust produced would be proportional
to the land area disturbed for the new facilities, all of which would be located near the center of the
Site. The areas affected by each alternative would be as follows:
- No Action - 0 acres
- Decentralization, 1992/1993 Planning Basis and Regionalization A (by fuel type) - 6 to
9 acres
- Regionalization B (by location) - 7 to 11 acres
- Centralization - 70 to 100 acres
- Shipping fuel offsite - 1 acre
DOE anticipates that overall construction impacts to air quality would be minimal and of a short
duration (6 months to 3 years). The SRS sitewide compliance with state and Federal ambient air
quality standards would not be affected by any construction-related activities associated with spent fuel
management.
Table 5-7. Estimated maximum annual emissions (in curies) of radionuclides to the atmosphere from
spent nuclear fuel management activities.
Radionuclide Annual Emissionsa,b
Tritium (elemental) 1.88x105,c
Cesium-134 3.60x10-4
Cesium-137 4.07x10-3
Curium-244 2.00x10-4
Cerium-141 1.83x10-3
Cerium-144 3.11x10-2
Americium-241 2.27x10-4
Cobalt-60 4.00x10-6
Plutonium-238 1.28x10-3
Plutonium-239 4.01x10-4
Strontium-90 1.39x10-2
Rubidium-103 7.25x10-3
Uranium-235 2.00x10-3
Osmium-185 3.60x10-4
Nibium-95 2.89x10-2
Selenium-75 1.52x10-5
Zirconium-95 1.68x10-2
Rubidium-106 5.12x10-3
Krypton-85 6.80x105
Carbon-14 2.80x101
a. Source: Hamby (1993).
b. Source terms are taken from 1985/86 F-/H-Area releases.
c. Historically, less than 10 percent of the atmospheric tritium releases have been from processing
operations in the F-/H-Area Canyons.
5.7.1 Alternative 1 - No Action
The SRS would not process any spent nuclear fuel under the No Action alternative. Normal site
baseline emissions would continue (Tables 5-3, 5-4, 5-5, 5-6 and 5-7). DOE would not construct any
new facilities under this alternative.
5.7.2 Alternative 2 - Decentralization
Atmospheric emissions under two of the Decentralization options (dry storage and wet storage)
would be similar to those for No Action. Those from the processing of the spent fuel (Option 2c)
would be of somewhat higher concentrations (Tables 5-3, 5-4, 5-5, 5-6 and 5-7). The emissions would
originate from existing facilities involved in the management of spent fuel under this alternative as
well as new ones that DOE would construct (Figure 3-2).
5.7.3 Alternative 3 - 1992/1993 Planning Basis
Emissions to the atmosphere would be similar to those for Alternative 2 because the amount of
fuel managed would be similar [223 and 220 MTHM (246 and 243 tons), Alternative 3 and
Alternative 2 respectively] and the facilities required would be the same (Figure 3-2).
5.7.4 Alternative 4 - Regionalization
Regionalization A (by fuel type). Atmospheric emissions would be similar to the releases from
Alternative 2 because of the similarity in volumes of fuel managed [213 and 220 MTHM (235 and
243 tons), respectively] and in the facilities involved (Figure 3-2).
Regionalization B (by location). Emissions would be somewhat higher than for
Regionalization A for both dry and wet storage options if the SRS receives all the spent fuel in the
eastern portion of the country, because the Site would manage about 20 percent more fuel.
Atmospheric emissions from processing would not change from those under other alternatives because
the amount of aluminum-clad fuel involved would be the same. Facility requirements would also be
similar (Figure 3-2).
Shipping all of the current SRS inventory off the Site (Option 4g) would result in the lowest
emissions to the atmosphere of any of the options under this alternative. These releases would result
from the characterization and canning of the fuel prior to shipment.
5.7.5 Alternative 5 - Centralization
The atmospheric emissions resulting from centralizing all the spent nuclear fuel at the SRS would
be the greatest of all the alternatives. The Site would manage about 2,740 MTHM (3,020 tons) of
fuel. Releases from storage activities for centralization would be proportionally higher than for the
other alternatives where the SRS would manage about 206 to 257 MTHM (227 to 283 tons) of spent
fuel. However, emissions from processing under Alternative 5 would be similar to those under the
other alternatives because the same amount of aluminum-clad fuel would be processed in each case.
The facilities required under all three options would be similar in function (Figure 3-2) but of much
larger capacity than for other alternatives.
Shipping all the SRS fuel to another site (Option 5d) would result in the lowest level of
atmospheric releases of any alternative, similar to those under Regionalization B, Option 4g.
5.8 Water Quality and Related Consequences
SRS use of surface-water and groundwater resources under any of the alternatives would not
substantially increase the volumes currently used for process, cooling, and domestic water on the Site.
Table 5-8 summarizes the groundwater and surface water usage requirements for each alternative and
option, and compares them to current SRS usages.
The Centralization Alternative (Option 5c), under which DOE would transfer all spent nuclear
fuel to the SRS, would result in the largest amount of water use [approximately 378.5 million liters
(100 million gallons) per year], which is a small amount compared to current SRS water requirements
of approximately 89.7 billion liters (23.7 billion gallons) per year. This represents an increase of
approximately 0.4 percent above current usage. Therefore, DOE anticipates that water use under any
of the alternatives would have minimal impact on the water resources of the Site.
The impact on water quality of the operation of any of the alternatives would also be minimal.
Existing SRS treatment facilities could accommodate all new spent fuel-related domestic and process
wastewater streams. The expected total SRS flow volumes would still be well within the design
capacities of the Site treatment systems. Because these plants would continue to meet National
Pollutant Discharge Elimination System limits and reporting requirements, DOE expects no impact on
the water quality of the receiving streams. The increased cooling water flows would also meet all
discharge permit limits and would have minimal impacts on the receiving water.
Each of the alternatives would contribute to the very small releases of radionuclides that normal
SRS operations discharge to the surface water through federally permitted wastewater outfalls.
Table 5-8. Annual groundwater and surface water usage requirements for each alternative. ,b
Groundwater Surface Water
Alternative Usage per Year Usage per Year Total Annual
Current SRS Usage 14.0 billion liters 75.7 billion liters 89.7 billion liters
No Action
Option 1 - Wet Storage 35.1 million liters None 35.1 million liters
Decentralization
Option 2a - Dry Storage 48.7 million liters 6.1 million liters 54.8 million liters
Option 2b - Wet Storage 50.6 million liters 7.2 million liters 57.8 million liters
Option 2c - Processingc 48.7 million liters 310.8 million liters 359.5 million liters
Planning Basis
Option 3a - Dry Storage 48.7 million liters 6.1 million liters 54.8 million liters
Option 3b - Wet Storage 50.6 million liters 7.2 million liters 57.8 million liters
Option 3c - Processingc 48.7 million liters 310.8 million liters 359.5 million liters
Regionalization - A
Option 4a - Dry Storage 48.7 million liters 6.1 million liters 54.8 million liters
Option 4b - Wet Storage 50.6 million liters 7.2 million liters 57.8 million liters
Option 4c - Processingc 47.6 million liters 308.8 million liters 356.5 million liters
Regionalization - B
Option 4d - Dry Storage 48.7 million liters 6.1 million liters 54.8 million liters
Option 4e - Wet Storage 50.6 million liters 7.2 million liters 57.8 million liters
Option 4f - Processingc 48.7 million liters 310.8 million liters 356.5 million liters
Option 4g - Ship Outc 38.1 million liters 3.0 million liters 41.1 million liters
Centralization
Case 5a - Dry Storage 67.7 million liters 6.1 million liters 73.8 million liters
Case 5b - Wet Storage 69.6 million liters 7.2 million liters 76.8 million liters
Case 5c - Processingc 67.7 million liters 310.8 million liters 378.5 million liters
Case 5d - Ship Outc 38.1 million liters 3.0 million liters 41.1 million liters
a. Source: WSRC (1994b).
b. To convert liters to gallons, multiply by 0.26418.
c. First 10 years only.
Table 5-9 summarizes the estimated maximum amounts of radioactivity that could be released to the
Savannah River in liquid effluents from normal spent nuclear fuel management activities. DOE used
actual liquid releases from F- and H-Area during 1985 and 1986 to estimate potential releases that
could occur during spent fuel management activities. DOE believes the isotopes and amounts released
during this period are representative of releases that could occur during processing under any of the
alternatives. This is because 1985 and 1986 represent periods when the F- and H-Area separations
facilities operated at or near peak capacity to process spent nuclear fuel. Estimated releases from wet
or dry storage would be less than these amounts. Consequently, the estimated releases given in
Table 5-9 represent the upper limit of liquid radiological releases that DOE expects as a result of spent
Table 5-9. Estimated maximum liquid radiological releases (in curies) to the Savannah River from
spent nuclear fuel management activities.
Radionuclide Annual Releasea,b
Tritium 1.3x104,c
Strontium-90 2.4x10-1
Iodine-129 2.2x10-2
Cesium-137 1.1x10-1
Plutonium-239 7.0x10-3
a. Source: Hamby (1993).
b. Source terms are taken from 1985/86 F-/H-Area releases.
c. Less than 1 percent of this quantity was from processing operations in F-/H-Area.
nuclear fuel management activities. The consequences to human health due to these releases are
discussed in Section 5.12, Occupational and Public Health and Safety.
Construction of new facilities under any alternative would require amounts of water that would
be only a very small percentage of the current daily water use at the SRS. Good engineering practice
measures would prevent sediment runoff or spills of fuel or chemicals. Therefore, construction
activities should have no impact on surface or groundwater quality at the Site.
DOE also analyzed the potential impacts of accidents in F- and H-Areas on surface and
groundwater quality. The analysis evaluated two types of accidental releases: one to the ground
surface (e.g., overflow of a wet storage pool) and another directly to the subsurface (e.g., failure of a
pool liner). Because pool water could contain some radionuclides, but would not contain any toxic or
harmful chemicals, the following evaluation addresses only the consequences of radionuclide releases.
A release of pool water onto the ground from the Receiving Basin for Offsite Fuels, in H-Area,
would not flow directly into any stream or other surface-water body. The building is in a graded,
gravel-covered area among other buildings and alongside a railroad spur and access road. A tank farm
surrounded by an earthen berm is immediately to the south. A channelized drainage ditch begins
approximately 244 meters (800 feet) west of the basin building and passes through culverts under a
railroad line and Road E before emptying into a tributary of Fourmile Branch about 500 meters
(1,650 feet) from the Receiving Basin. The grading at the Site would contain a small volume of water
overflowing the basin in the immediate area of the building. In the unlikely event that a larger spill
reached the drainage ditch to the west, DOE could contain the water by blocking either of the two
culverts through which the drainage ditch passes. After containing the spilled water, DOE could
remove and properly dispose of it. DOE would design and construct new facilities containing storage
pools in a manner that would confine any overflow or other surface release of pool water. Therefore,
DOE believes that there will be no direct release to surface water from spills of pool water at an
existing or potential facility.
An overflow from a pool could reach the groundwater by slowly flowing downward from the
surface through the unsaturated zone until it reached the water table, which is 9 to 15 meters (30 to
50 feet) below the grade in the F- and H-Areas. Overflow water would take several years to reach the
water table, based on a vertical velocity of between 0.9 and 2.1 meters (3 to 7 feet) per year (DOE
1987). As discussed in the following paragraphs, once in the groundwater, a plume would take many
years to reach either of the closest surface-water bodies, Fourmile Branch to the south or Upper Three
Runs Creek to the north.
DOE has calculated the travel times of groundwater in the F- and H-Areas based on specific
information on the hydraulic conductivity, the hydraulic gradient, and the effective porosity of aquifers
in this area (WSRC 1993a) and on the use of Darcy's Law. Water would take between 16 and 500
years to travel 1.6 kilometers (1 mile) toward Fourmile Branch or Upper Three Runs Creek. These
estimates of travel time agree with values obtained from the results of DOE modeling studies
performed on the F- and H-Areas (Geotrans 1993; appended to WSRC 1993a). The reason for this
wide range of potential travel time is that the hydraulic conductivity of aquifer materials is highly
variable and can vary in the same aquifer by several orders of magnitude. This slow movement
through the subsurface, either vertically through the unsaturated zone or horizontally within the
aquifer, would facilitate the removal of radionuclides from the spill plume through a number of
processes. These include radioactive decay, trapping of particulates in the soil, and ion exchange and
adsorption by the soil (Hem 1989). DOE believes that travel time of a contaminant plume through the
subsurface in the F- or H-Area or in the adjacent representative host site would be such that no
radionuclides would reach Fourmile Branch, Upper Three Runs Creek, or any other surface-water body
by this route. For the same reasons, no radioactive contaminants introduced into the subsurface in
these areas would move off the Site in groundwater.
DOE does not believe that releases of radionuclides such as those described above would reach
SRS drinking-water sources that lie in deep aquifers under the Site. These aquifers are several
hundred feet below the ground surface, and a number of thick aquifers and aquitards separate them
from the water table aquifer (see Section 4.8). In addition to the distances and the presence of
confining layers, vertical flow in the intervening stratified sedimentary aquifers is slow in comparison
to horizontal flow. Radionuclide contamination of offsite drinking water sources is even more
unlikely given the depth of their source aquifers, the distances involved, and the attenuation of
contaminants in the soils, as described above.
DOE also evaluated a second kind of unintentional release in the F- or H-Area, a direct leak to
the subsurface from a breach in a storage pool during routine operations. The analysis assumed a
19-liter (5-gallon)-per-day leak as a result of secondary containment or piping failure at a new state-of-
the-art wet storage and fuel transfer facility (Creed 1994). The analysis assumed further that the leak
would go undetected for 1 month, a conservative assumption given the sensitivity of the leak detection
equipment that these new facilities would require. The reliability and sensitivity of the leak detection
devices would be equal to or superior to those required by the U.S. Nuclear Regulatory Commission
(NRC 1975) for spent nuclear fuel storage facilities in commercial nuclear power plants. DOE would
require spent nuclear fuel storage pools (whether fuel unloading pools or storage basins) to have leak
detection monitoring devices, pool water level monitors, and radiation monitors designed to alarm both
locally and in a continuously staffed central location. Constant process monitoring, mass balance, and
facility design (including double-walled containment of vessels and piping) would also be used by
DOE to limit operational releases from new wet storage facilities, including fuel unloading pools and
storage basins, to near zero.
To provide a common basis for analysis of spent nuclear fuel alternatives at its various sites,
DOE developed a generic infrastructure design for a hypothetical spent nuclear fuel complex (Hale
1994). This design includes proposed criteria for temporary wet storage basins, fuel loading and
unloading pools, and transfer canals.
Based on the design criteria in Hale (1994), a leak from one of these basins if constructed in
F- or H-Area could result in the introduction of radionuclide-contaminated water into the ground at
depths as much as 13.4 meters (44 feet) below grade. Such a release would go directly to the water
table aquifer or to the unsaturated zone above it, depending on the depth of the water table. In either
case, the processes governing the slow plume movement (i.e., the hydraulic conductivity, hydraulic
gradient, and effective porosity of aquifers in the F- and H-Areas) and the processes resulting in the
attenuation of contaminants and radionuclides (i.e., radioactive decay, trapping of particulates in the
soil, ion exchange in the soil, and adsorption to soil particles) described in the previous paragraphs
would also prevent or mitigate impacts to surface-or groundwater resources from releases of this type.
There could be localized contamination of groundwater in the surface aquifer in the immediate vicinity
of the storage facilities. This aquifer is not used as a source of drinking water. DOE believes that no
radionuclide contamination of deeper confined aquifers that are sources of onsite or offsite drinking
water could occur from a release of this type. And, as noted earlier, these wet storage facilities would
be equipped with state-of-the-art leak detection devices, pool level monitors, and radiation monitors
that would limit and mitigate any subsurface releases.
5.8.1 Alternative 1 - No Action
5.8.1.1 Option 1 - Wet Storage. During operations under this alternative, current levels of
water usage would not change. Nor would changes occur in thermal discharges from cooling water or
the quantity or quality of radioactive and nonradioactive wastewater effluents.
The viable accidents under this alternative would be a release of pool water onto the ground
surface or a breach of the liner of the wet storage basins in which the spent nuclear fuel would be
stored. As discussed above, radionuclides in the released water would enter the water table aquifer but
would not reach any surface-water or any drinking water aquifer on or off the SRS. Basin water
contains no toxic or hazardous chemicals. Therefore, accidental releases from the basins would have
minimal impacts on surface- and groundwater resources.
Spills of chemicals would not reach surface- or groundwater due to existing proper engineering
design and environmental controls, and to rapid containment and cleanup.
5.8.2 Alternative 2 - Decentralization
Operations under either the dry or wet storage option for the Decentralization alternative would
increase Site water usage by less than 0.1 percent above current levels. Processing would increase use
by about 0.4 percent. Release of nonradioactive and radioactive materials to surface waters would
increase only slightly and would be well within discharge permit limits and DOE dose limits. There
would be no releases to groundwater during normal operations. Overall impacts to water quantity and
water quality would be minimal.
Impacts to water resources due to accidental releases onto the ground or into the subsurface
would also be minimal as explained above. Potential contamination would be limited to the surface
aquifer.
5.8.3 Alternative 3 - 1992/1993 Planning Basis
DOE expects that the impacts to water resources under the dry storage, wet storage, and
processing cases for this alternative would be similar to those described for the same options under
Alternative 2, Decentralization. Overall impacts would be minimal.
5.8.4 Alternative 4 - Regionalization
DOE expects that the impacts to water resources under the three options for regionalization by
fuel type (Regionalization A) would be similar to those described for the same options under
Alternative 2, Decentralization. Regionalization B (by geographic location) would result in impacts
somewhat greater than those for Alternative 2 because the SRS would have to manage an additional 37
MTHM (41 tons) of spent fuel. In either case, overall impacts would be minimal. For Option 4g,
shipping all SRS fuel to Oak Ridge Reservation, impacts to water resources would be the smallest of
any alternative, similar to those for Option 5d - Centralization.
5.8.5 Alternative 5 - Centralization
The first three options for this alternative - dry storage (Option 5a), wet storage (Option 5b), and
processing (Option 5c) - assume that DOE would transfer all spent nuclear fuel to the SRS for
management. The impacts of operations to water resources under these options would be similar in
nature to the impacts for the same options under Alternative 2, Decentralization, as described in
Section 5.8.2. However, the extent of the impacts would be greater because the number and size of
facilities that DOE would construct and operate and the quantities of fuel it would manage would be
larger than those for any other alternative. Even so, DOE expects the overall impacts of construction
and operation to be minor. For example, the total volume of water that the SRS would withdraw for
construction, cooling, processing, and domestic use under any of these three options would not exceed
approximately 378.5 million liters (100 million gallons) per year. This requirement would be
approximately 0.4 percent of the 89.7 billion liters (23.7 billion gallons) that the SRS currently uses
annually.
Similarly, DOE believes that the overall impacts of accidents under any of these three options
would be minor, even though the number and size of the facilities would be greater under this
alternative than for any other. Radionuclides released during an accident would not affect any
surface-water or any drinking water aquifer. However, surface aquifer resources would receive
contamination in the area of any release.
For Option 5d (shipping the spent nuclear fuel off the Site), impacts to water resources would be
smaller than those for any other alternative or option. DOE would have to build only one new facility
(for fuel characterization) and the spent fuel would remain at SRS only for the first part of the 40-year
management period. Overall impacts would be minimal.
5.9 Ecology
DOE expects that construction impacts, which would include loss of some wildlife habitat due to
land clearing, would be greatest under the Centralization Alternative, Dry Storage option.
Representative impacts from operations would include disturbance and displacement of animals caused
by movement and noise of personnel, equipment, and vehicles; however, these impacts would be
minor under all the proposed alternatives. Construction and operation would not disturb any critical or
sensitive habitat, nor would they affect any wetland areas. Releases of radionuclides to the
environment from any of the proposed alternatives would be small and would not be expected to
accumulate in aquatic or terrestrial ecosystems or measurably affect the health or viability of plant and
animal communities.
5.9.1 Alternative 1 - No Action
Under this alternative, DOE could refurbish or modify existing wet storage facilities and would
confine any activity to these facilities. As a consequence, DOE expects no impacts to ecological
resources. Impacts of operations under this alternative would be minimal, limited to some minor
disturbance of animals by vehicular traffic.
5.9.2 Alternative 2 - Decentralization
5.9.2.1 Option 2a - Dry Storage. This option would require some new construction, but any
construction activity would occur either within the boundaries of F- and H-Areas, which are already
heavily developed, or adjacent to them. As a result, this construction would have little or no impact
on ecological resources. There would be no impacts to wetlands, threatened or endangered species,
socially or commercially important species (such as the eastern wild turkey), or disturbance-sensitive
species (such as wood warblers and vireos). Impacts of operations under this option would be limited
to some minor disturbance of animals by slight increases in vehicular traffic. No threatened,
endangered, or candidate species occur in the area of operations. Species likely to be disturbed or
killed by vehicles (e.g., cotton rat, gray squirrel, opossum, and white-tailed deer) are common to
ubiquitous in the area. Overall impact to ecological resources would be minimal.
5.9.2.2 Option 2b - Wet Storage. Construction impacts would be similar to those described
for dry storage (Option 2a). Impacts of operations under this option would also be similar to those
described for dry storage (Option 2a). Overall impacts to ecological resources would be minimal.
5.9.2.3 Option 2c - Processing and Storage. Construction and operations impacts for this
option would also be similar to those for dry storage (Option 2a). Overall impacts would still be
minimal.
5.9.3 Alternative 3 - 1992/1993 Planning Basis
Both construction and operational impacts for the three options under this alternative would be
similar to those described for Alternative 2 - Decentralization. Overall impacts would be minimal.
5.9.4 Alternative 4 - Regionalization
Under the Regionalization A alternative, impacts to ecological resources would be minimal as
described for Alternative 2. Impacts due to the Regionalization B options would be somewhat greater
due to the larger volume of spent fuel that the SRS would manage. Overall impacts would still be
minimal, however.
The smallest impacts would occur under Option 4g because DOE would ship all spent fuel off
the Site.
5.9.5 Alternative 5 - Centralization
5.9.5.1 Option 5a - Dry Storage. The discussion that follows assumes that any facility
development would take place in an area that does not contain any pristine wetlands, old growth
timber, threatened and endangered species, or designated critical habitat. More specifically, because
the upland areas south and east of H-Area are dominated by planted pine (primarily loblolly and slash)
stands, the discussion of impacts assumes that any facility development in support of spent nuclear
fuel management would take place in an area of 5- to 40-year-old pines. Finally, the analysis assumes
that any facility development would require a site-specific National Environmental Policy Act (NEPA)
review as required under 10 CFR Part 1021 and in accordance with the Council on Environmental
Quality's NEPA implementing regulations (CFR 1991b).
The proposed interim dry storage facility and support facilities, requiring approximately
0.28 square kilometers (70 acres) to 0.4 square kilometer (100 acres) of land, would be built
somewhere within the largely wooded roughly 2.8 square kilometer (700-acre) area south and east of
H-Area west of F-Road, and north of Fourmile Branch. This area has a number of advantages; among
them: it would be relatively easy to connect with existing utilities (gas, water, sewer); it would
minimize the amount of supporting infrastructure (e.g., railroad spurs, access roads, and transmission
lines) that would have to be built; and it would enable DOE to consolidate spent nuclear fuel
management activities in an area that has been altered many times over the years by farming (before
1951) and timber management activities (after 1951).
Construction activities would result in the clearing of as much as approximately 0.4 square
kilometer (100 acres) of planted 5- to 40-year-old loblolly or slash pine for new facilities on the
undeveloped representative host site south and east of H-Area. This land clearing would involve a
relatively small number of loggers and heavy equipment operators, but probably would drive most
birds and larger, more mobile animals from the area. Some smaller, less mobile animals, such as
turtles, toads, lizards, mice, and voles, probably would be killed. Aside from the loss of 0.28 to
0.4 square kilometer (70-100 acres) of planted pines that provide habitat for a limited number of
reptiles, birds, and mammals, construction impacts would be minor.
Any land clearing and timber harvesting conducted on the undeveloped host site would be
carefully planned and conducted according to widely accepted Best Management Practices to minimize
erosion and soil loss and to prevent impacts to downgradient wetlands and streams. DOE and SRS
policy is to achieve "no net loss" of wetlands. DOE has issued a guidance document, Information for
Mitigation of Wetlands Impacts at the Savannah River Site (DOE 1992), for project planners that puts
forth a practical approach to wetlands protection that begins with avoidance of impacts (if possible),
moves to minimization of impacts (if avoidance is impossible), and requires compensatory measures
(wetlands restoration, creation, enhancement, or acquisition) in the event that impacts cannot be
avoided.
In the event that new facility development was required, DOE would perform predevelopment
surveys to ensure that its activities would not affect threatened and endangered species or sensitive
habitats. To the extent practicable, land clearing and timber harvesting would be restricted to times of
the year when songbirds and game birds were not nesting or rearing young. In South Carolina, most
songbirds nest, rear, and fledge young from March to September (Sprunt and Chamberlain 1970).
Quail, dove, and wild turkey in the region normally nest and fledge young during the spring and
summer (Sprunt and Chamberlain 1970).
No threatened or endangered plants or animals are known to be present in the area under
consideration for development. Construction activities probably would not affect two small wetlands
(Carolina bays) lying in the east-central portion of the undeveloped host site. Construction activities
would not affect plant and animal diversity locally or regionally, because the managed loblolly and
slash pine stands that would be removed are not unique, nor do they provide habitat for any protected,
sensitive, unusual, or Federally listed plant or animal species.
Impacts of operations under this option would be similar to, but slightly greater than, those
described for Option 2a. Overall impacts to ecological resources would be minor.
5.9.5.2 Option 5b - Wet Storage. Construction impacts under this option would be less than
those described for Option 5a because less land area would be required for new facilities. Impacts of
operations under this case would be similar to those described for Option 5a. Overall impacts to
ecological resources would be minor.
5.9.5.3 Option 5c - Processing and Storage. Construction impacts under this case would
be similar to those described for Option 5a. This case would require the largest number of workers of
all the cases under consideration. It would result in more noise, more traffic, and a generally higher
level of disturbance to terrestrial wildlife (specifically reptiles, songbirds, and small and large
mammals) accustomed to feeding, foraging, perching, hunting, nesting, or denning in the area. Some
animals would be driven from the area permanently, while others probably would become accustomed
to the increased noise and activity levels, and would return to the area. Overall impacts to ecological
resources would be minor.
5.9.5.4 Option 5d - Shipment off the Site. Construction impacts under this case would be
smaller than those for any other alternative, excluding Alternative 1 - No Action. Impacts of operation
under this case would also be minimal, limited to some minor disturbances of animals by vehicular
traffic. Overall impacts to ecological resources would be minimal.
5.10 Noise
As described in Section 4.10, noises generated on the SRS do not travel off the Site at levels that
affect the general population. Therefore, SRS noise impacts for each alternative would be limited to
noise resulting from the transportation of personnel and materials to and from the Site that could affect
nearby communities and from onsite sources that could affect some wildlife near these sources. DOE
would address the effects of noise on wildlife near spent nuclear fuel management facilities under any
alternative in a project-specific NEPA evaluation.
Transportation noises would be a function of the size of the workforce (i.e., an increased
workforce would produce increased employee traffic and corresponding increases in deliveries by truck
and rail and a decreased workforce would produce decreased employee traffic and corresponding
decreases in deliveries). The analysis of traffic noise took into account railroad noise and noise from
the major roadways that provide access to the SRS. DOE does not expect the number of freight trains
per day in the region and through the Site to change as a result of any of the alternatives, although
some trains could be dedicated to the transport of spent nuclear fuel. Rail shipments of spent nuclear
fuel, regardless of the alternative, would not substantially increase the rail traffic on the CSX line
through the SRS. Therefore, vehicles used to transport employees and personnel on roadways would
be the principal sources of community noise impacts. This analysis used the day-night average sound
level (DNL) to assess community noise, as suggested by the Environmental Protection Agency
(EPA 1974; 1982) and the Federal Interagency Committee on Noise (FICON 1992). The analysis
based its estimate of the change in day-night average sound level from the baseline noise level for
each alternative on the projected changes in employment and traffic levels. The baseline levels are
those for 1995. The analysis also considered the combination of construction and operation
employment. The traffic noise analysis considered SC 125 and SC 19, both of which are used to
access the SRS. Changes in noise level below 3 decibels would not be likely to result in a change in
community reaction (FICON 1992).
DOE projects no new employment due to operations for any of the alternatives. Some additional
construction jobs may be required but overall SRS employment would not exceed the 1995 baseline
levels, except for Alternatives 5a, 5b, and 5c. The maximum Site employment of about 20,000 jobs
would occur in 1995 for all alternatives except 5a, 5b, and 5c for which the peak would occur in about
2002 due to a peak in construction employment. The general decrease in employment after 1995
could result in some decrease in vehicle trips to and from the Site. There would be at most a few
truck trips per day to and from the Site carrying spent nuclear fuel under any of the alternatives. This
increase in truck trips would not result in a perceptible increase in traffic noise levels along the routes
to the SRS. The day-night average sound level along SC 125 and SC 19 and other access routes
would probably decrease slightly except in the peak construction years under Alternatives 5a, 5b, and
5c, as a result of the overall decrease in employment levels at the SRS after 1995. DOE expects no
change in the community reaction to noise along these routes. Consequently, no mitigation efforts are
necessary.
5.11 Traffic and Transportation
This section discusses the consequences of both the onsite transportation of spent nuclear fuel
and the increased traffic patterns due to construction activities at the SRS. Traffic due to operations of
spent nuclear fuel facilities will remain at or below current Site levels because workers for the new
activities will be drawn from the existing SRS workforce. The consequences of the transportation of
spent fuel between the SRS and other DOE sites are described in Appendix I of Volume 1 of this
Environmental Impact Statement (EIS).
5.11.1 Traffic
Traffic impacts would be bound by Alternative 5b (Centralization - Wet Storage) which would
result in the greatest number of additional construction workers (and vehicles) onsite. Level of
service, a measure of traffic flow, was estimated for each road to and from the SRS. Traffic delays
could be experienced at SC 19 and SC 230 intersections during peak hours. However, the number of
construction vehicles in support of spent nuclear fuel construction activities would contribute less than
17 percent (HNUS 1994) to the total traffic flow. Therefore, the change in level of service due to
Alternative 5b would be minimal.
5.11.2 Transportation
This section discusses the potential radiological consequences due to incident free transportation
and accidents during transport. All SRS onsite shipments are carried out by rail.
5.11.2.1 Onsite Spent Nuclear Fuel Shipments. DOE based the number of fuel
shipments on the amount and type of spent nuclear fuel stored at various SRS locations and the final
storage location or disposition specified in the spent nuclear fuel alternatives. The number of
shipments from each location was determined by dividing the amount of spent nuclear fuel at each
location by the capacity of the shipping cask. Individual shipments from the various facilities were
summed to obtain the total number of shipments for each alternative (HNUS 1994).
Onsite shipments are those that originate and terminate at the SRS. Movements of spent nuclear
fuel within functional areas (e.g., H-Area or F-Area) are operational transfers, not onsite shipments;
therefore, this analysis does not consider them.
5.11.2.2 Incident-Free Transportation Analysis. Under each alternative, DOE analyzed
incident-free (normal transport) radiological impacts to transport vehicle crews and members of the
general public from onsite rail shipments. The analysis calculated occupational radiation doses to the
transport vehicle crew members (four locomotive operators). Because the general public does not have
immediate access to areas where the SRS would transport spent nuclear fuel, the analysis assumed that
any general public dose is to escorted individuals on the Site waiting at any of several train crossings
at the time a fuel shipment passed. The analysis calculated radiological doses to the general public
using the Riskind (Yuan et al. 1993) computer code. The results are presented in Table 5-10.
The magnitude of incident-free consequence depends on the dose rate on the external surface of
the transport vehicle, the exposure time, and the number of people exposed. For each receptor, the
analysis assumed the external dose rate 2 meters (6.6 feet) from the shipping cask was 100 millirem
per hour (HNUS 1994), which is the SRS procedurally-allowed maximum dose rate during onsite fuel
shipments. Actual receptor dose rates would depend on receptor distance from the shipping cask
[5 meters (16.4 feet) for the general public]. The duration of exposure would depend on the transport
vehicle speed and the number of shipments. In addition, occupational exposure time would depend on
the distance of each shipment.
The analysis calculated health effects measured as the number of latent cancer fatalities (LCFs)
by multiplying the resultant occupational and general public doses by risk factors of 4 x 10-4 and
5 x 10-4 latent cancer fatalities per person-rem (DOE 1993a), respectively.
Table 5-10 summarizes the collective doses (person-rem) and health effects (latent cancer
fatalities) associated with the incident-free onsite shipment of spent nuclear fuel at the SRS. Collective
Table 5-10. Collective doses and health effects for onsite, incident-free spent nuclear fuel shipments
by alternative.
Option Occupational General Public Number of LCFsa
(person-rem) (person-rem)
Occupational General Public
No Action
Option 1b -Wet Storage 1.5x100 1.4x10-1 6.0x10-4 7.0x10-5
Decentralization
Option 2a - Dry Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 2b - Wet Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 2c - Processing 5.3x10-1 3.7x10-2 2.1x10-4 1.9x10-5
Planning Basis
Option 3a - Dry Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 3b - Wet Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 3c - Processing 5.3x10-1 3.7x10-2 2.1x10-4 1.9x10-5
Regionalization
Option 4a - Dry Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 4b - Wet Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 4c - Processing 5.3x10-1 3.7x10-2 2.1x10-4 1.9x10-5
Option 4d - Dry Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 4e - Wet Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 4f - Processing 5.3x10-1 3.7x10-2 2.1x10-4 1.9x10-5
Option 4g - Ship Out NAb NAb NAb NAb
Centralization
Option 5a - Dry Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 5b - Wet Storage 2.5x100 2.3x10-1 1.0x10-3 1.2x10-4
Option 5c - Processing 5.3x10-1 3.7x10-2 2.1x10-4 1.9x10-5
Option 5d - Ship Out NAb NAb NAb NAb
a. LCF = latent cancer fatality.
b. NA = not applicable.
doses and latent cancer fatalities for members of the public would be approximately a factor of 10 less
than those for the occupational worker. The data indicate that the lowest collective doses and lowest
latent cancer fatality would be associated with the Processing option under the Decentralization,
Planning Basis, Regionalization, and Centralization alternatives.
5.11.2.3 Transportation Accident Analysis. DOE analyzed radiological impacts from
potential accidents to both the onsite maximally exposed individual (MEI), and offsite members of the
general public from onsite rail shipments. The analysis calculated doses using the Riskind (Yuan
et al. 1993) computer code with site-specific meteorology, demographics, and spent fuel activity. Risk
was calculated using site-specific rail accident rates and accident probabilities (HNUS 1994).
The magnitude of accident consequence would depend on the amount of radioactive material to
which the individual(s) was exposed, the exposure time, and the number of people exposed. The
analysis assumed that the maximum reasonably foreseeable amount of radioactive material for the type
of spent fuel shipped on the SRS was released (HNUS 1994). The assumed duration of exposure for
each receptor was 2 hours. The assumed maximally exposed individual was an SRS worker
downwind of the accident at distances of 50 and 100 meters (164 and 330 feet).
The analysis calculated offsite exposure using both rural and suburban population density-specific
census data. The rural and suburban population densities have an average of 6 persons per square
kilometer and 244 persons per square kilometer, respectively. The west-northwest sector has the
highest population density within 80 kilometers (50 miles) of the SRS.
The analysis used site-specific meteorology at the 50th and 95th percentile to determine dose
consequences. Joint probability includes both the event frequency and the probability of the maximum
reasonably foreseeable type of accident occurring.
The analysis calculated health effects measured as the number of latent cancer fatalities by
multiplying the resultant occupational and general public doses by the risk factors of 4 x 10-4 and
5 x 10-4 latent cancer fatalities per person-rem (DOE 1993a), respectively. Risk was calculated by
multiplying the resultant doses by the joint probability of 1 x 10-4 (HNUS 1994).
Tables 5-11 and 5-12 summarize the collective doses and associated latent cancer fatalities for
postulated onsite rail accidents with subsequent releases of radioactive material to the environment.
The dose consequences of an accidental release of radioactive material was assessed for the 95th and
typical 50th percentile meteorological conditions (i.e., those that would result in lower doses 95 and 50
percent of the time, respectively). In all cases the estimated number of latent cancer fatalities would
be low.
5.11.3 Onsite Mitigation and Preventative Measures
All onsite shipments must be in compliance with DOE Savannah River Directive Implementation
Instruction 5480.3, "Safety Requirements for the Packaging and Transportation of Hazardous Materials,
Hazardous Substances, and Hazardous Wastes." DOE, DOE-SR, or the Nuclear Regulatory
Commission (NRC) must approve packages used for onsite shipments with a certificate of
Table 5-11. Impacts on maximally exposed individual from spent nuclear fuel transportation accident
on the Savannah River Site.
Dose Percentile Distance Dose to Number of Risk
(meters) MEIa (rem) LCFsb per year
50 percent 100 0.16 6.4x10-5 1.6x10-5
95 percent 50 0.37 1.5x10-4 3.7x10-5
a. MEI = maximally exposed individual.
b. LCF = latent cancer fatality.
Table 5-12. Impacts on offsite population from spent nuclear fuel transportation accident on the
Savannah River Site.
Population Dose Offsite Population Number of LCFsa Risk
Density Category Percentile Dose (person-rem) per year
Rural 50th 1.7 8.7x10-4 1.7x10-4
Rural 95th 7.1 3.6x10-3 3.6x10-3
Suburban 50th 5.2 2.6x10-3 2.6x10-3
Suburban 95th 21.3 1.1x10-2 1.1x10-2
a. LCF = latent cancer fatality.
compliance. If DOE or NRC has not certified an onsite package as Type B, the shipper must establish
administrative controls and site-mitigating circumstances that will ensure package integrity. The
administrative and emergency response considerations must provide sufficient control so that accidents
would not result in loss of containment, shielding, or criticality; or the uncontrolled release of
radioactive material would not create a hazard to the health and safety of the public or workers.
In the event of an accident, SRS has established an emergency management program. This
program incorporates activities associated with emergency planning, preparedness, and response.
5.12 Occupational and Public Health and Safety
5.12.1 Radiological Health
This human health effects analysis relied principally on data on F- and H-Area emissions
documented for the 1985, 1986, and 1993 operating years (Marter 1986; 1987; WSRC 1994d). During
the 1985-1986 period, F- and H-Areas processing facilities operated at high capacity; DOE believes,
therefore, that these emissions represent conservative estimates as to the emissions that could result
from spent nuclear fuel management activities at the SRS. This air and surface-water emissions
information defined the source terms for the baseline evaluation (No Action alternative) of health
effects discussed in this section. To estimate health effects, this analysis defined six human receptor
groups:
- The F- and H-Area workers assigned to F- and H-Area operations involving nuclear
materials
- The F- and H-Area workers assigned to the Receiving Basin for Offsite Fuels for storage
operations
- The maximally exposed individual residing at the SRS boundary
- The projected 1994 offsite population of 628,200 persons residing within an 80-kilometer
(50-mile) radius of F- and H-Areas
- The maximally exposed individual potentially affected by SRS surface-water emissions
- The approximate offsite population of 65,000 persons whom SRS surface-water emissions
could affect.
With the exception of the worker group, this analysis calculated exposures for the remaining four
receptor groups using the baseline source terms as input data to automated atmospheric and surface-
water transport, human intake, and human dosimetry models configured for routine use at SRS
(Hamby 1994). The analysis estimated worker exposures using averaged dosimetry data recorded for
F- and H-Area workers from 1983 through 1987 and Receiving Basin for Offsite Fuels workers for
1993 (Matheny 1994), corrected for an assumed occupancy factor of 0.25 (i.e., a worker could be
potentially exposed during one-quarter of his/her shift). This correction was applied to the 1983-1987
data only. At the SRS, the waterborne exposure pathway does not exist for the worker receptor group
because Site drinking water is drawn from deep aquifers unaffected by any radiological releases.
The analysis developed incremental receptor group exposure estimates (millirem per year, person-
rem per year; effective dose equivalent) based on spent fuel quantities for each of the nonbaseline
alternatives (i.e., Alternatives 2 through 5) and their options by applying calculated ratios of metric
tons of heavy metal (MTHM) for each alternative and option compared to the No Action alternative.
DOE used these ratios as incremental scaling factors to estimate exposures under each option. The
calculation of the MTHM ratios used the data presented in Table 3-1. Table 5-13 lists the results of
the exposure estimate calculations. Since these incremental exposures include contributions to the
effective dose equivalent from existing (No Action) spent fuel management at the SRS, the change in
health effects for each alternative can be estimated as the difference between the alternatives presented.
The analysis calculated the potential health effects expressed in the exposed receptor groups
consistent with risk determination guidance issued by the DOE Office of NEPA Oversight (DOE
1993a) and International Commission on Radiological Protection Publication 60 (ICRP 1991). For
exposed individuals and populations, the potential health effect (detriment) of interest is latent fatal
cancer. For exposed individuals, this analysis presents the health effect as the maximum incremental
probability for detriment expression; for exposed populations, it presents the annual incremental
detriment incidence. For completeness, it also provides the "project life" (i.e., 40 years) detriment
incidence as the annual incidence multiplied by 40. Table 5-14 (worker) and Table 5-15 (maximally
exposed individual and offsite population) summarize the health effects calculations.
The Centers for Disease Control and Prevention is conducting a comprehensive reconstruction of
historic offsite doses associated with SRS operations. The results of this investigation are not yet
available.
5.12.2 Nonradiological Health
DOE used the operations air quality data listed in Tables 5-3, 5-4, 5-5 and 5-6 (and Table 8 of
WSRC 1994a) to evaluate health impacts associated with potential exposure to the following two
compound classes: criteria pollutants and toxic pollutants. The analysis evaluated two hypothetical
receptor locations: (1) a worker in S-Area and (2) a maximally exposed individual at the SRS
boundary. However, it was unnecessary to postulate an intake of criteria pollutant or toxic compounds
by these receptors because airborne concentration standards are available for these compounds.
Tables 5-3 and 5-4 list 8 criteria pollutants and 23 toxic compounds. The toxic compounds were
classified as carcinogens and noncarcinogens consistent with Environmental Protection Agency
carcinogenicity group (weight of evidence) designations published in the Integrated Risk Information
Table 5-13. Incremental radioactive contaminant annual exposure summary.
Onsite Workersa MEI Offsitea,b,d Offsite
(mrem/year) Populationa,d
(person-rem/
year
(person-
Alternative (mrem/ rem/ Air Water Air Water
year)c year)
No Action - Wet Storage (Option 1) 100 0.2 9x10-8 3x10-8 4x10-6 6x10-7
Decentralization - Dry Storage 83 0.2 8x10-8 2x10-8 3x10-6 5x10-7
(Option 2a)
Decentralization - Wet Storage 104 0.2 9x10-8 3x10-8 4x10-6 6x10-7
(Option 2b)
Decentralization - Processing 145 70 0.4 0.1 14 2.2
(Option 2c)
Planning Basis - Dry Storage 84 0.2 8x10-8 2x10-8 3x10-6 5x10-7
(Option 3a)
Planning Basis - Wet Storage 105 0.2 1x10-7 3x10-8 4x10-6 6x10-7
(Option 3b)
Planning Basis - Processing 147 71 0.4 0.1 15 2.2
(Option 3c)
Regionalization A - Dry Storage 83 0.2 8x10-8 2x10-8 3x10-6 5x10-7
(Option 4a)
Regionalization A - Wet Storage 103 0.2 9x10-8 3x10-8 4x10-6 6x10-7
(Option 4b)
Regionalization A - Processing 148 76 0.4 0.1 16 2.4
(Option 4c)
Regionalization B - Dry Storage 105 0.2 1x10-7 3x10-8 4x10-6 6x10-7
(Option 4d)
Regionalization B - Wet Storage 131 0.3 1x10-7 4x10-8 5x10-6 7x10-7
(Option 4e)
Regionalization B - Processing 175 74 0.4 0.1 15 2.3
(Option 4f)
Regionalization B - Ship Out <100 <0.2 <9x10-8 <3x10-8 <4x10-6 <6x10-7
(Option 4g)
Centralization - Dry Storage 1,102 2.2 1x10-6 3x10-7 4x10-5 6x10-6
(Option 5a)
Centralization - Wet Storage 1,377 2.8 1x10-6 4x10-7 5x10-5 8x10-6
(Option 5b)
Centralization - Processing (Option 5c) 1,422 79 0.4 0.1 16 2.4
Centralization - Ship Out (Option 5d) <100 <0.2 <9x10-8 <3x10-8 <4x10-6 <6x10-7
a. Insignificant digits are displayed for comparison purposes only.
b. MEI = maximally exposed individual.
c. The DOE administrative dose limit is 2,000 mrem (DOE 1994a).
d. Data is provided separately for the air and water exposure pathways because the receptors are not
co-located.
Table 5-14. Incremental fatal cancer incidence and maximum probability for workers.
Annual 40-Year Maximum
Alternative Incidencea Incidence Probability
No Action - Wet Storage (Option 1) 8x10-5 3x10-3 4x10-5
Decentralization - Dry Storage (Option 2a) 7x10-5 3x10-3 3x10-5
Decentralization - Wet Storage (Option 2b) 8x10-5 3x10- 4x10-5
-3
Decentralization - Processing (Option 2c) 3x10-2 1 6x10-5
Planning Basis - Dry Storage (Option 3a) 7x10-5 3x10-3 3x10-5
Planning Basis - Wet Storage (Option 3b) 8x10-5 3x10- 4x10-5
-3
Planning Basis - Processing (Option 3c) 3x10-2 1 6x10-5
Regionalization A - Dry Storage (Option 4a) 7x10-5 3x10-3 3x10-5
Regionalization A - Wet Storage (Option 4b) 8x10-5 3x10- 4x10-5
-3
Regionalization A - Processing (Option 4c) 3x10-2 1 6x10-5
Regionalization B - Dry Storage (Option 4d) 8x10-5 3x10-3 4x10-5
Regionalization B - Wet Storage (Option 4e) 1x10-4 4x10- 5x10-5
-3
Regionalization B - Processing (Option 4f) 3x10-2 1 7x10-5
Regionalization B - Ship Out (Option 4g) <8x10-5 <3x10-3 <4x10-5
Centralization - Dry Storage (Option 5a) 9x10-4 4x10-2 4x10-4
Centralization - Wet Storage (Option 5b) 1x10-3 4x10- 5x10-4
-2
Centralization - Processing (Option 5c) 3x10-2 1 6x10-4
Centralization - Ship Out (Option 5d) <8x10-5 <3x10-3 <4x10-5
a. Number of latent fatal cancers over a lifetime which could be attributed to one year of spent
nuclear fuel management activities.
System (IRIS) data base (DOE 1994b). For purposes of health effects analysis, carcinogens are those
compounds designated Group A (human carcinogens), Group B1 (probable human carcinogen, limited
evidence in human studies), Group B2 (probable human carcinogen, inadequate evidence or no data
from human studies), and Group C (possible human carcinogen). Using this designation, three of the
23 toxic compounds are carcinogens: benzene (Group A), formaldehyde (Group B1), and methylene
chloride (Group B2).
Carcinogen health effects are expressed as the incremental probability of an individual
developing cancer, assuming a lifetime (70 years) of exposure to the carcinogen. DOE used cancer
risk (slope) factors published in IRIS (Integrated Risk Information System) to obtain unit risk factors
(risk per concentration) needed to calculate incremental probability. Carcinogens with insufficient (i.e.,
incomplete or unavailable carcinogen assessment data) information listed in the Integrated Risk
Information System data base precluded a quantitative risk assessment; this analysis evaluated them as
noncarcinogens.
Table 5-15. Incremental fatal cancer incidence and maximum probability for the maximally exposed
individual and offsite population (air and water pathways).
Population Population MEI
Alternative Annual 40-Year Maximum
Incidencea Incidence Probability
No Action - Wet Storage (Option 1)
Air 2x10-9 7x10-8 4x10-14
Water 3x10-10 1x10-8 1x10-14
Decentralization - Dry Storage (Option 2a)
Air 2x10-9 6x10-8 4x10-14
Water 2x10-10 9x10-9 1x10-14
Decentralization - Wet Storage (Option 2b)
Air 2x10-9 8x10-8 5x10-14
Water 3x10-10 1x10-8 2x10-14
Decentralization - Processing (Option 2c)
Air 7x10-3 0.3 2x10-7
Water 1x10-3 4x10-2 6x10-8
Planning Basis - Dry Storage (Option 3a)
Air 2x10-9 6x10-8 4x10-14
Water 2x10-10 9x10-9 1x10-14
Planning Basis - Wet Storage (Option 3b)
Air 2x10-9 8x10-8 5x10-14
Water 3x10-10 1x10-8 2x10-14
Planning Basis - Processing (Option 3c)
Air 7x10-3 0.3 2x10-7
Water 1x10-3 4x10-2 6x10-8
Regionalization A - Dry Storage (Option 4a)
Air 2x10-9 6x10-8 4x10-14
Water 2x10-10 9x10-9 1x10-14
Regionalization A - Wet Storage (Option 4b)
Air 2x10-9 8x10-8 5x10-14
Water 3x10-10 1x10-8 2x10-14
Regionalization A - Processing (Option 4c)
Air 8x10-3 0.3 2x10-7
Water 1x10-3 5x10-2 6x10-8
Regionalization B - Dry Storage (Option 4d)
Air 2x10-9 8x10-8 5x10-14
Water 3x10-10 1x10-8 2x10-14
Regionalization B - Wet Storage (Option 4e)
Air 2x10-9 1x10-7 6x10-14
Water 4x10-10 1x10-8 2x10-14
Regionalization B - Processing (Option 4f)
Air 8x10-3 0.3 2x10-7
Water 1x10-3 5x10-2 6x10-8
Regionalization B - Ship Out (Option 4g)
Air <2x10-9 <7x10-8 <4x10-14
Water <3x10-10 <1x10-8 <1x10-14
Table 5-15. (continued).
Population Population MEI
Alternative Annual 40-Year Maximum
Incidencea Incidence Probability
Centralization - Dry Storage (Option 5a)
Air 2x10-8 8x10-7 5x10-13
Water 3x10-9 1x10-7 2x10-13
Centralization - Wet Storage (Option 5b)
Air 3x10-8 1x10-6 6x10-13
Water 4x10-9 2x10-7 2x10-13
Centralization - Processing (Option 5c)
Air 8x10-3 0.3 2x10-7
Water 1x10-3 5x10-2 6x10-8
Centralization - Ship Out (Option 5d)
Air <2x10-9 <7x10-8 <4x10-14
Water <3x10-10 <1x10-8 <1x10-14
a. Number of latent fatal cancers over a lifetime that could be attributed to one year of spent nuclear fuel
management activities.
This analysis evaluated noncarcinogenic and priority pollutant compound health effects by adding
hazard quotients to obtain a hazard index. The hazard quotient is the ratio of compound concentration
or dose to a Reference Concentration (RfC) or Dose (RfD) (EPA 1989). The regulatory standard used
in this analysis was the more stringent of the following: (1) Occupational Safety and Health
Administration (OSHA) 8-hour permissible exposure limit (PEL), (2) American Conference of
Governmental Industrial Hygienists (ACGIH) threshold limit value (TLV), or (3) State of South
Carolina air quality standards. The use of the noncancer hazard index assumed a level of exposure
(i.e., RfC) below which adverse health effects are unlikely. The hazard index is not a statistical
probability; therefore it cannot be interpreted as such.
Table 5-16 summarizes nonradiological health effects attributable to atmospheric emissions of
toxic and criteria pollutant compounds. Because no hazard index value would exceed unity (1.0),
adverse health effects are unlikely under any alternative.
5.12.3 Industrial Safety
This section describes the following measures of impact for workplace hazards: (1) total
reportable injuries and illnesses and (2) fatalities in the work force. This analysis considers
injury/illness and fatality incidence rates for construction workers separately because of the relatively
Table 5-16. Nonradiological annual incremental health effects summary.
Alternative Worker Cancer Worker Hazard MEI Cancer MEI Hazard Index
Probabilitya Index Probabilitya,b
No Action - Wet Storage Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 1)
Decentralization - Dry Storage Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 2a)
Decentralization - Wet Storage Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 2b)
Decentralization - Processing Insufficient data 6x10-3 Insufficient data 5x10-4
(Option 2c)
Planning Basis - Dry Storage Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 3a)
Planning Basis - Wet Storage Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 3b)
Planning Basis - Processing Insufficient data 6x10-3 Insufficient data 5x10-4
(Option 3c)
Regionalization A - Dry Insufficient data 2x10-6 Insufficient data 2x10-7
Storage (Option 4a)
Regionalization A - Wet Insufficient data 2x10-6 Insufficient data 2x10-7
Storage (Option 4b)
Regionalization A - Processing Insufficient data 6x10-3 Insufficient data 5x10-4
(Option 4c)
Regionalization B - Dry Insufficient data 2x10-6 Insufficient data 3x10-7
Storage (Option 4d)
Regionalization B - Wet Insufficient data 2x10-6 Insufficient data 3x10-7
Storage (Option 4e)
Regionalization B - Processing Insufficient data 8x10-3 Insufficient data 6x10-4
(Option 4f)
Regionalization B - Ship Out Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 4g)
Centralization - Dry Storage Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 5a)
Centralization - Wet Storage Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 5b)
Centralization - Processing Insufficient data 6x10-3 Insufficient data 5x10-4
(Option 5c)
Centralization - Ship Out Insufficient data 2x10-6 Insufficient data 2x10-7
(Option 5d)
a. Insufficient data exists in the IRIS data base to perform a quantitative inhalation cancer risk assessment.
b. MEI = maximally exposed individual.
more hazardous nature of construction work. Table 5-17 lists the incidence of injuries/illnesses and
fatalities for construction and non-construction workers. These data are for the highest employment
year (i.e., maximum hours worked in any year from 1994 through 2035, assuming 2,000 hours per
worker) (WSRC 1994b). This analysis used the average occupational injury/illness and fatality
incidence rates experienced by DOE and its contractors from 1988 through 1992 to calculate the
incidence of industrial hazards listed in Table 5-17 (DOE 1993b).
Table 5-17. Incremental industrial hazard maximum annual incidence summary.
Alternative Construction Construction Nonconstruction Nonconstruction
Injuries and Fatalities Injuries and Fatalities
Illnesses Illnesses
No Action - Wet Storage 92 <1 159 <1
(Option 1)
Decentralization - Dry Storage 71 <1 159 <1
(Option 2a)
Decentralization - Wet Storage 71 <1 159 <1
(Option 2b)
Decentralization - Processing 66 <1 159 <1
(Option 2c)
Planning Basis - Dry Storage 71 <1 159 <1
(Option 3a)
Planning Basis - Wet Storage 82 <1 159 <1
(Option 3b)
Planning Basis - Processing 66 <1 159 <1
(Option 3c)
Regionalization A - Dry 82 <1 159 <1
Storage (Option 4a)
Regionalization A - Wet 82 <1 159 <1
Storage (Option 4b)
Regionalization A - Processing 66 <1 159 <1
(Option 4c)
Regionalization B - Dry 89 <1 199 <1
Storage (Option 4d)
Regionalization B - Wet 102 <1 199 <1
Storage (Option 4e)
Regionalization B - Processing 82 <1 199 <1
(Option 4f)
Regionalization B - Ship Out 22 <1 159 <1
(Option 4g)
Centralization - Dry Storage 316 1 159 <1
(Option 5a)
Centralization - Wet Storage 337 1 159 <1
(Option 5b)
Centralization - Processing 316 1 159 <1
(Option 5c)
Centralization - Ship Out 22 <1 159 <1
(Option 5d)
5.13 Utilities and Energy
The existing capacities and distribution systems at the SRS for electricity, steam, water, and
domestic wastewater treatment are adequate to support any of the five alternatives. Table 5-18
summarizes estimates of the annual requirements for electricity, steam, and domestic wastewater
treatment for each alternative and case, and compares them to current SRS usage of these resources.
Table 5-8 lists information on water usage by alternative. The utility and energy requirements for all
Table 5-18. Estimates of annual electricity, steam, and domestic wastewater treatment requirements
for each alternative. ,b
Domestic Wastewater
Electricity Usage Steam Usage Treatment
Alternative (megawatt hours per year) (kilograms per year)c (liters per year)d
Current SRS Usage 659,000 1.7 billion 690 million
1. No Action
Option 1 - Wet 1,400 11.3 million 35.1 million
Storage
2. Decentralization
Option 2a - Dry 19,400 16.7 million 48.7 million
Storage
Option 2b - Wet 22,400 14.4 million 50.6 million
Storage
Option 2c - Processing 56,400 19.1 million 48.7 million
3. 1992/1993 Planning Basis
Option 3a - Dry 19,400 16.7 million 48.7 million
Storage
Option 3b - Wet 22,400 14.4 million 50.6 million
Storage
Option 3c - Processing 56,400 19.1 million 48.7 million
4. Regionalization - A
Option 4a - Dry 24,400 16.7 million 48.7 million
Storage
Option 4b - Wet 27,400 14.4 million 50.6 million
Storage
Option 4c - Processing 67,400 16.5 million 47.6 million
Regionalization - B
Option 4d - Dry 24,400 16.7 million 48.7 million
Storage
Option 4e - Wet 27,400 14.4 million 50.6 million
Storage
Option 4f - Processing 56,400 19.1 million 48.7 million
Option 4g - Ship Out 11,400 11.7 million 38.1 million
5. Centralization
Option 5a - Dry 44,400 16.7 million 67.7 million
Storage
Option 5b - Wet 47,400 14.4 million 69.6 million
Storage
Option 5c - Processing 110,400 19.1 million 67.7 million
Option 5d - Ship Out 11,400 11.7 million 38.1 million
a. Source: WSRC (1994b).
b. Water requirements are shown in Table 5-8.
c. To convert kilograms to pounds, multiply by 2.2046.
d. To convert liters to gallons, multiply by 0.26418.
the alternatives represent a small percentage of current requirements. No new generation or treatment
facilities would be necessary; connections to existing networks would require only short tie-in lines.
Increases in SRS fuel consumption would be minimal because overall activity on the Site would not
increase due to changes in the SRS mission and the general reduction in employment levels. The
overall impacts of any of the alternatives on the SRS utilities and energy resources would be minimal.
The smallest increase in demand would result from the No Action alternative, which would be
similar to current spent nuclear fuel-related requirements at the SRS. The largest increases would be
due to the centralization of spent nuclear fuel at the SRS (Alternative 5). Alternative 5 would result in
a maximum additional electrical demand of about 110,400 megawatt-hours annually (Option 5c), and
an increased steam consumption of about 19.1 million kilograms (42.1 million pounds) per year
(Option 5c). Water requirements would also be greatest under this Alternative (Table 5-8). Annual
withdrawals of Savannah River water for cooling purposes would reach about 310.8 million liters
(82.1 million gallons) and groundwater usage for domestic and processing purposes would total
approximately 69.6 million liters (18.4 million gallons). The volume of domestic wastewater requiring
treatment would range from approximately 35 to 70 million liters (9 to 18 million gallons) per year.
This additional water usage amounts to an increase of about 10 percent over current SRS water
requirements.
Among the three management options, processing would result in the greatest increase in demand
on utilities and energy in comparison to either the dry or wet storage options. In general, dry and wet
storage would be similar in their requirements of these resources.
5.14 Materials and Waste Management
This section discusses potential impacts of the management of materials and wastes associated
with the implementation of alternatives identified for spent nuclear fuel management. Sections 5.7 and
5.12 (Air Quality and Occupational and Public Health and Safety, respectively) discuss the impacts of
hazardous and toxic materials as they relate to routine operations and accidents.
DOE has projected rates and volumes of waste and impacts of waste generation at SRS for low-
level, transuranic, and high-level wastes for each of the alternatives for spent nuclear fuel management.
Table 5-19 summarizes the estimated annual average and total volume of these three waste types that
each alternative would produce during a 40-year management period. The discussion
Table 5-19. Annual average and total volume (cubic meters)d of radioactive wastes produced under
each alternative during the 40-year interim management period.
Low-level wasteb Transuranic waste High-level wastec
Alternative Average Total Average Total Average Total
1. No Action
Option 1 - Wet Storage 400 16,000 17 700 0.4 4
2. Decentralization
Option 2a - Dry Storage 400 16,000 18 720 0.4 4
Option 2b - Wet Storage 400 16,000 18 720 0.4 4
Option 2c - Processing 800 32,000 19 760 2.3 23
3. 1992/1993 Planning Basis
Option 3a - Dry Storage 400 16,000 18 720 0.4 4
Option 3b - Wet Storage 400 16,000 18 720 0.4 4
Option 3c - Processing 750 30,000 19 760 1.7 17
4. Regionalization - A
Option 4a - Dry Storage 400 16,000 17 700 0.4 4
Option 4b - Wet Storage 400 16,000 17 700 0.4 4
Option 4c - Processing 790 31,600 18 720 2.3 23
4. Regionalization - B
Option 4d - Dry Storage 400 16,000 17 700 0.4 4
Option 4e - Wet Storage 400 16,000 17 700 0.4 4
Option 4f - Processing 790 31,600 18 720 2.3 23
Option 4g - Ship Out 400 4,000 18 180 0 0
5. Centralization
Option 5a - Dry Storage 400 16,000 16 640 0 0
Option 5b - Wet Storage 400 16,000 20 800 2.3 23
Option 5c - Processing 800 32,000 20 800 2.3 23
Option 5d - Ship Out 400 4,000 18 180 0 0
a. Based on WSRC (1994b).
b. Source: WSRC (1994c).
c. Figures are for the initial 10-year period when most processing would be completed.
d. To convert cubic meters to cubic yards multiply by 1.307.
below also identifies the impacts that the waste produced by spent nuclear fuel activities would have
on the existing SRS capacity to manage each waste type.
DOE has not developed estimates of low-level mixed, hazardous, or solid sanitary wastes that
spent nuclear fuel management activities at the SRS could generate, although it is anticipated that
these activities would produce these waste types only in limited quantities. Further, the discussions in
Section 5.14.2 related to the impacts of spent fuel management wastes on the SRS waste capacities do
not include considerations of wastes that will result from Site cleanup because assessments for these
activities are still underway and will undergo NEPA review as part of the SRS Waste Management
Environmental Impact Statement (DOE 1995).
Volume 1 of this spent nuclear fuel EIS provides information concerning the major Federal
environmental laws and regulations, Executive Orders, and DOE Orders that apply to pollution
prevention at the Savannah River Site. The DOE views source reduction as the first priority in its
pollution prevention program, followed by an increased emphasis on recycling. Source reduction will
reduce the waste management burden while eliminating the potential for future liability and cleanup.
Recycling and using recycled materials will conserve resources and landfill space. Waste treatment
and disposal are considered only when prevention or recycling is not possible or practical. Since
creating a Savannah River Site waste minimization program (the precursor of the SRS pollution
prevention program) in 1990, the amounts of wastes of all types (excluding low-level wastes, which
are a by-product of environmental restoration activities) generated have decreased, with greatest
reductions in hazardous and mixed wastes (Hoganson and Miles 1994).
5.14.1 Alternative Comparison
The first four alternatives would generate similar amounts of radioactive waste because the
activities that produce the wastes would be similar under each of the alternatives. Most of the low-
level and transuranic wastes would be generated during the first part of the 40-year management
period while DOE was transferring existing inventory and renovating the Receiving Basin for Offsite
Fuels and a reactor basin. The characterization and canning of the current inventory prior to
placement into storage would also result in some waste generation. Once in storage, management
activities would produce only small amounts of radioactive waste for the rest of the 40-year period.
The dry- and wet-storage options would both produce about 16,000 cubic meters (20,912 cubic
yards) of low-level waste and between 640 cubic meters (836 cubic yards) and 800 cubic meters
(1,046 cubic yards) of transuranic waste during the 40-year management period. Both options would
generate small amounts of high-level waste. The processing of the existing aluminum-clad fuels and
storage of the others (the third option under each alternative) would generate all three types of waste:
low-level and high-level wastes in appreciably greater volumes, and transuranic waste in slightly-
greater volumes.
Alternative 5 (excluding the Ship Out option) could result in somewhat larger volumes of
radioactive waste than the other four alternatives. However, any increase in waste would not be
directly proportional to the larger amounts of fuel that would be managed on the Site, because most of
the originating sites would characterize and can their fuel prior to shipment so that it could be placed
directly into storage at the SRS. Therefore, the radioactive wastes produced during centralization at
the Site would come from the initial fuel transfer and pool renovations and from characterizing and
canning small amounts of new fuel. The processing of existing aluminum-clad fuels would produce
the same types and volumes of waste as for the other alternatives.
The option for shipping the SRS inventory off the Site for regionalization or centralization
elsewhere would also result in the production of some radioactive waste. This would occur during
characterization and canning prior to shipment and would generate the smallest volumes of waste of
any alternative action: 4,000 cubic meters (5,228 cubic yards) of low-level waste and 180 cubic
meters (235 cubic yards) of transuranic waste. This waste would be produced only during the initial
10 years of the management period.
5.14.2 Impact on the SRS Waste Management Capacity
The impact of spent nuclear fuel activities on SRS waste management capacities would be
minimal because the Site could accommodate the waste with existing and planned radioactive waste
storage and disposal facilities. DOE would transfer high-level waste to the F/H Tank Farms for
volume reduction and then to the Defense Waste Processing Facility (DWPF) for conversion into a
borosilicate glass form suitable for prolonged storage. The SRS would use the Consolidated
Incineration Facility, once operational, to treat the low-level waste. This facility has sufficient
permitted capacity [105,500 cubic meters (137,889 cubic yards) per year] to treat the anticipated
volume of these materials. However, actual through-put volume is dependent upon operational
variables and waste characteristics. The F/H Effluent Treatment Facility would treat liquid low-level
waste. This facility has sufficient design process capacity [598 million liters (158 million gallons) per
year] to treat the anticipated volumes of these materials. DOE would manage the transuranic wastes
with existing and planned storage capacity.
5.15 Accident Analysis
Operations involving the receipt, handling, processing, or storing of spent nuclear fuel would
involve radioactive materials or toxic chemicals. These materials would be received, treated, stored,
transferred between facilities, disposed of on the Site, and shipped off the Site. Under certain
circumstances, these materials could be involved in an accident.
An accident is a series of unexpected or undesirable events initiated by equipment failure, human
error, or a natural phenomenon such as severe weather, earthquake, or volcanism. These events can
cause the release of either radioactive or chemically toxic materials inside a facility or to the
environment.
This section summarizes analyses of possible accidents involving spent nuclear fuel operations at
the SRS. To provide a perspective on potential accidents, this section summarizes various accidents
associated with spent nuclear fuel activities that have occurred at the SRS (historic accidents) and
reviews previous accident analyses for Site operations. This section uses the results of previous
analyses as a baseline for determining the impacts for the alternatives that involve new facilities. For
each alternative, this section discusses the accidents with the largest point estimates of risk
(radiological impacts in terms of potential fatal cancers x frequency of the initiating event).
The facilities considered for each alternative are either existing facilities for which the approved
safety analyses were used, or new facilities (WSRC 1994b) for which existing safety analysis results
were substituted by evaluating the type of accident(s) that could be postulated to occur based on the
projected function of the facility. Two facilities that contain very small amounts of contact-handled
spent nuclear fuel, Buildings 331-M and 773-A, were not included in this analysis because accidents
analyzed for the major facilities would bound the consequences of possible accidents in these two
locations.
This section addresses historic accidents, facility radiological accidents, chemical hazard
accidents, and secondary impacts. Section 5.11 addresses onsite transportation accidents.
5.15.1 Historic Accidents at the Savannah River Site
Impacts from accidents can involve fatalities, injuries, or illness. Fatalities can be prompt
(immediate) such as in construction accidents or latent (delayed) such as an increase in latent fatal
cancers due to radiation exposure. Section 5.12 addresses worker injuries, illnesses, and the potential
for increased cancer risk anticipated from normal operations of the facilities. Nonradiation accidents
have dominated impacts to workers at the SRS (Durant et al. 1987); impacts to the public from
historic SRS accidents have been negligible.
The SRS has maintained an operational event data base on its facilities since the 1950s. This
data base currently contains approximately 450,000 entries including data on the Receiving Basin for
Offsite Fuel, the principal wet storage pool facility at the SRS; and both F-and H-Area Canyons. For
this EIS, DOE reviewed the data base to identify historic spent nuclear fuel-related accidents at these
facilities. Fuel cutting events, fuel handling events, and various liquid releases related to spent nuclear
fuel management over the 40-year operating history of the SRS were examined. The purpose of the
data base review was to provide an historic perspective on the types of accidents that have occurred at
the SRS. Events representative of fuel failures include higher than expected contamination levels in
fuel storage basin water and evidence of fuel canister cracking at a weld. Fuel handling incidents were
due in large part to crane operator errors or crane and handling equipment failures. The data base also
includes reports of incorrect fuel cropping, where the active region of fuel was exposed under water.
These historical events provided a basis for the selection of representative accidents covering the
spectrum of spent nuclear fuel management activities. No significant offsite impacts have resulted
from these historic occurrences.
5.15.2 Potential Facility Accidents
The SRS spent nuclear fuel alternatives have the potential for radiological accidents (see
Attachment A, Table A-2) that could affect the health and safety of workers and the public. The
concerns and characteristics that are common to these accidents would be common regardless of
whether the cause were a natural phenomenon or human error. For health effects to occur, an accident
must allow a release of hazardous material to, or an increase in radiation levels in, the facility or the
environment. The released material must be transported to locations frequented by humans. The
quantities of hazardous materials that reach locations where people are and the ways they interact with
people are important factors in the determination of health effects.
A number of studies have investigated the ways in which radioactivity reaches humans, how the
body absorbs and retains it, and the resulting health effects. The International Commission on
Radiological Protection has made specific recommendations for estimating these health effects
(ICRP 1991). This organization is the recognized body for establishing standards for the protection of
workers and the public from the effects of radiation exposure. Health effects include acute damage
(up to and including death) and latent effects, including cancers and genetic damage. An
SRS-developed computer code, AXAIR89Q, estimates potential radiation doses to maximally exposed
individuals or population groups from accidental releases of radionuclides.
The AXAIR89Q code is a highly automated site-specific environmental dispersion and dosimetry
code for postulated airborne releases. The environmental dispersion models used are based on NRC
Regulatory Guide 1.145 (NRC 1983). The exposure pathways considered in the AXAIR89Q code
include inhalation of radionuclides and gamma irradiation from the radioactive plume.
Doses from the inhalation of radionuclides in air depend on the amount of radionuclides released;
the dispersion factor; the physical, chemical, and radiological characteristics of the radionuclides; and
various biological parameters such as breathing rate and biological half-life. The AXAIR89Q code
uses a conservative breathing rate of 12,000 cubic meters (424,000 cubic feet) per year for adults. The
dose commitment factors used in the environmental dosimetry code, as described in the following
section, are from Internal Dose Conversion Factors for Calculation of Dose to the Public (DOE 1988).
External gamma radiation doses from the traveling plume depend on the spatial distribution of
the radionuclides in the air, the energy of the radiation, and the extent of shielding. The AXAIR89Q
code takes no credit for shielding in calculating doses. The code calculates gamma doses using a
nonuniform Gaussian model, which has more realistic modeling than doses from the conventional
uniform semi-infinite plume model.
In addition to using the worst sector, 99.5 percentile meteorology, conservative breathing rates,
and taking no credit for shielding, the AXAIR89Q code also takes no credit for the probable plume
rise from stack releases. Therefore, the offsite maximum individual doses calculated by AXAIR89Q
provide conservative bounding estimates of radiological consequences to exposed individuals and
populations from postulated accidental atmospheric releases.
AXAIR89Q has been validated for compliance to accepted standards for such software.
Attachment A, Accident Analysis, discusses AXAIR89Q and its predecessor, AXAIR. When used in
conjunction with models for predicting health effects, the results from AXAIR89Q can be compared
with other site-specific codes such as RSAC-5, because both codes provide relative radionuclide
concentrations based on the guidance provided in NRC Regulatory Guide 1.145.
This section summarizes the potential for radiological accidents and their consequences for the
cases under each alternative. Attachment A describes the methodology and assumptions used in the
assessment; describes radiological accident scenarios in more detail; provides source terms and
references used to estimate the doses and impacts for each alternative and case; and includes scaling
factors that the DOE decisionmaker can apply to the source term or dose for each facility associated
with a case.
DOE assessed the potential impacts from a selected spectrum of radiological release accidents,
ranging from low (1 x 10-6 event per year) to high (more than 1 event per year) frequencies of
occurrence, along with the associated impacts (doses and potential latent fatal cancers) that could
result. The accidents used as references are attributed to individual facilities based on their functions
and processes (see Attachment A, Table A-3), not to specific cases or alternatives. This enables a
comparison of alternatives depending on which facilities support a specific case or alternative.
Figure 5-1 is a flowchart for the preparation of accident analysis information. No new analyses
occurred because existing documentation adequately supports a quantitative or qualitative estimation of
potential impacts, as required by the National Environmental Policy Act of 1969. The assessment of
postulated radiological accidents associated with spent nuclear fuel at the SRS indicates that the
highest point estimate of risk to the public within 80 kilometers (50 miles) of the Site would be
1.4 x 10-3 latent fatal cancer per year. The estimated dose to the same population from all causes,
including natural background sources, would be about 19,000 person-rem per year (DOE 1990), which
could cause about nine latent fatal cancers per year in the same population. For perspective, natural
background radiation sources would result in approximately 6,000 times the risk associated with the
largest consequence accident postulated in this EIS for the various spent nuclear fuel management
alternatives.
DOE did not quantitatively analyze the potential health effects for SRS workers less than 100
meters (328 feet) from radiological accidents. Computer codes used to calculate radiological doses can
experience potentially large errors as a source disperses throughout a building. However, DOE did
carry out a qualitative evaluation of the potential radiological effects to SRS workers in the immediate
vicinity of an accident related to spent fuel management. DOE estimates that the consequences of an
accident for the most part would result in higher than normal radiation doses. However, no fatalities
would occur except in the event of an inadvertent criticality in FB-Line, where up to four fatalities
may result. This evaluation is discussed in more detail in Section A.2.6.2 of Attachment A.
5.15.2.1 Alternative 1 - No Action. This alternative identifies the minimum actions deemed
necessary for continued safe and secure management of spent nuclear fuel at the SRS. As explained in
Chapter 3, this is not a status quo condition. Spent nuclear fuel would be maintained close to
defueling or current storage locations with minimal facility upgrade or equipment replacement. Only
local transport would occur. SRS activities required to safely store spent nuclear fuel would continue.
This alternative would require SRS to place corroded and pitted fuel elements in cans to minimize
spread of material into the pool. DOE estimated potential radiological accident impacts that could
occur under this alternative using existing DOE-approved safety analyses for the interim wet storage of
Figure 5-1. Accident analysis process. spent nuclear fuel at SRS facilities. As indicated in Attachment A, Table A-3, the facilities required
under this alternative would consist of existing facilities, including necessary upgrades to support safe
interim wet storage. In addition, Attachment A, Table A-4, provides a reference accident spectrum
associated with these facilities for this alternative. Attachment A, Table A-2, lists the references for
the source terms considered in analyzing potential accidents under this alternative, as well as their
estimated frequencies. Table 5-20 lists the accident scenario with the highest point estimates of risk to
the general public. Table 5-21 compares the potential radiological accidents and health effects of the
interim wet storage (Option 1) of spent nuclear fuel for the No Action alternative.
Table 5-20. Highest point estimates of risk among receptor groups (Option 1).
Receptor Groups
Maximally Exposed Population to 80 kilometers
Offsite Individual
Overall Point Estimate of Riska 1.6x10-7 (Fuel Assembly Breach) 1.4x10-3 (Fuel Assembly Breach)
a. Units of latent fatal cancers per year.
5.15.2.2 Alternative 2 - Decentralization. Accident assessments considered for this
alternative include those considered for the No Action alternative for wet storage (Option 2b) plus
assessments for the dry storage (Option 2a) of spent nuclear fuel and for the processing of spent fuel
(Option 2c). Option 2c (processing) assumes the use of existing facilities to dissolve, separate, and
further stabilize spent nuclear fuel. For cases that include some treatment (e.g., canning) of spent
nuclear fuel, such treatment is referred to as "stabilization," not processing. The amount of fuel of
various types to be considered would include those quantities from the production reactors, existing
research fuel, foreign research reactor fuel, and fuel transported for safety or research activities.
5.15.2.2.1 Option 2a - Dry Storage - DOE estimated potential radiological accident
impacts that could occur in this case using existing DOE-approved safety analysis reports submitted to
DOE by Westinghouse Savannah River Company for vault storage of special nuclear material from
existing facilities.
DOE has not incorporated the technology to support interim dry storage of spent
nuclear fuel at the SRS. To provide a basis for evaluating the potential impacts from this alternative
case, this assessment used data from existing safety analyses for special nuclear material storage
facilities and extrapolated these data to apply to spent nuclear fuel. DOE also considered radiological
accidents associated with wet storage, at least in the near term, because the spent nuclear fuel is
currently in wet storage. Similarly, this assessment includes fuel handling accidents throughout the
transition phase (i.e., until fuel is in interim dry storage). As indicated in Attachment A, Table A-4,
Table 5-21. Radioactive release accidents and health effects for spent nuclear fuel alternatives. ,b
Potential Fatal Cancers Point Estimate of Riskc
Frequency
Alternative (by case) Accident Scenario (per year)
Maximally Maximally
exposed Population to Colocated exposed Population to Colocated
offsite 80 kilometersd Workere Workere offsite 80 kilometersf Worker Worker
individuald individual
1. No Action
Option 1 Wet Storage A1 Fuel Assembly 1.6x10-1 1.0x10-6 8.5x10-3 (a) 4.8x10-6 1.6x10-7 1.4x10-3 (a) 7.7x10-7
Breach
A4 Material Release 2.4x10-3 3.0x10-6 2.5x10-2 (a) 2.0x10-5 7.2x10-9 6.0x10-5 (a) 4.8x10-8
(Adjacent Facility)
A5 Criticality in Water 3.1x10-3 1.5x10-6 4.4x10-3 (a) 5.6x10-5 4.7x10-9 1.4x10-5 (a) 1.7x10-7
A7 Spill/Liquid 2.0x10-4 2.7x10-6 9.0x10-3 (a) 1.1x10-6 5.4x10-10 1.8x10-6 (a) 2.2x10-10
Discharge (external)
A8 Spill/Liquid 1.1x10-1 1.2x10-13 1.0x10-9 (a) 8.0x10-15 1.3x10-14 1.1x10-10 (a) 8.8x10-16
Discharge (internal)
2. Decentralization
Option 2a Dry A1 Fuel Assembly 1.6x10-1 1.0x10-6 8.5x10-3 (a) 4.8x10-6 1.6x10-7 1.4x10-3 (a) 7.7x10-7
Storage Breach
A3 Material Release 1.4x10-3 1.1x10-9 3.5x10-6 (a) (b) 1.5x10-12 4.9x10-9 (a) (b)
(Dry Vault)
A4 Material Release 2.4x10-3 3.0x10-6 2.5x10-2 (a) 2.0x10-5 7.2x10-9 6.0x10-5 (a) 4.8x10-8
(Adjacent Facility)
A5 Criticality in Water 3.1x10-3 1.5x10-6 4.4x10-3 (a) 5.6x10-5 4.7x10-9 1.4x10-5 (a) 1.7x10-7
A7 Spill/Liquid 2.0x10-4 2.7x10-6 9.0x10-3 (a) 1.1x10-6 5.4x10-10 1.8x10-6 (a) 2.2x10-10
Discharge (external)
A8 Spill/Liquid 1.1x10-1 1.2x10-13 1.0x10-9 (a) 8.0x10-15 1.3x10-14 1.1x10-10 (a) 8.8x10-16
Discharge (internal)
Option 2b Wet A1 Fuel Assembly 1.6x10-1 1.0x10-6 8.5x10-3 (a) 4.8x10-6 1.6x10-7 1.4x10-3 (a) 7.7x10-7
Storage Breach
A4 Material Release 2.4x10-3 3.0x10-6 2.5x10-2 (a) 2.0x10-5 7.2x10-9 6.0x10-5 (a) 4.8x10-8
(Adjacent Facility)
A5 Criticality in Water 3.1x10-3 1.5x10-6 4.4x10-3 (a) 5.6x10-5 4.7x10-9 1.4x10-5 (a) 1.7x10-7
A7 Spill/Liquid 2.0x10-4 2.7x10-6 9.0x10-3 (a) 1.1x10-6 5.4x10-10 1.8x10-6 (a) 2.2x10-10
Discharge (external)
A8 Spill/Liquid 1.1x10-1 8.2x10-13 1.0x10-9 (a) 8.0x10-15 1.3x10-14 1.1x10-10 (a) 8.8x10-16
Discharge (internal)
Option 2c Processing A1 Fuel Assembly 1.6x10-1 1.0x10-6 8.5x10-3 (a) 4.8x10-6 1.6x10-7 1.4x10-3 (a) 7.7x10-7
Breach
A2 Material Release 2.6x10-1 3.4x10-8 2.6x10-4 (a) 3.6x10-8 8.9x10-9 6.8x10-5 (a) 9.4x10-9
(Processing)
Option 2c A3 Material Release 1.4x10-3 1.1x10-9 3.5x10-6 (a) (b) 1.5x10-12 4.9x10-9 (a) (b)
(continued) (Dry Vault)
A4 Material Release 2.4x10-3 3.0x10-6 2.5x10-2 (a) 2.0x10-5 7.2x10-9 6.0x10-5 (a) 4.8x10-8
(Adjacent Facility)
A5 Criticality in Water 3.1x10-3 1.5x10-6 4.4x10-3 (a) 5.6x10-5 4.7x10-9 1.4x10-5 (a) 1.7x10-7
A6 Criticality in 1.4x10-4 3.5x10-6 4.3x10-3 (a) 1.0x10-4 4.9x10-10 6.0x10-7 (a) 1.4x10-8
Processing
A7 Spill/Liquid 2.0x10-4 2.7x10-6 9.0x10-3 (a) 1.1x10-6 5.4x10-10 1.8x10-6 (a) 2.2x10-10
Discharge (external)
A8 Spill/Liquid 1.1x10-3 1.2x10-13 1.0x10-9 (a) 8.0x10-15 1.3x10-14 1.1x10-10 (a) 8.8x10-16
Discharge (internal)
3. 1992/1993 Planning Basis
Option 3a Dry Same as Option 2a for Decentralization
Storage
Option 3b Wet Same as Option 2b for Decentralization
Storage
Option 3c Processing Same as Option 2c for Decentralization
4. Regionalization - A
Option 4a Dry Same as Option 2a for Decentralization
Storage
Option 4b Wet Same as Option 2b for Decentralization
Storage
Option 4c Processing Same as Option 2c for Decentralization
4. Regionalization - B
Option 4d Dry Same as Option 2a for Decentralization
Storage
Option 4e Wet Same as Option 2b for Decentralization
Storage
Option 4f Processing Same as Option 2c for Decentralization
Option 4g Shipping Same as Option 1 for No Action
Out
5. Centralization
Option 5a Dry Same as Option 2a for Decentralization
Storage
Option 5b Wet Same as Option 2b for Decentralization
Storage
Option 5c Processing Same as Option 2c for Decentralization
Option 5d Shipping Same as Option 1 No Action
Out
a. The safety analysis reports from which information was extracted for these accidents were written before the issuance of DOE Order 5480.23; previous Orders did not require the inclusion of workers.
b. The safety analysis reports from which information was extracted for these accidents were written before the issuance of DOE Order 5480.23; previous Orders did not require the inclusion of
colocated workers.
c. Units for point estimates of risk are given in potential latent fatal cancers per year.
d. ICRP 60 risk factor for the general public (5.0 x 10-4 fatal cancer per year) was used to determine potential latent fatal cancers.
e. ICRP 60 risk factor for workers (4.0 x 10-4 fatal cancer per year) was used to determine potential latent fatal cancers.
the facilities required under this alternative would consist of existing and new facilities necessary to
support the safe handling, stabilization, and dry storage of spent nuclear fuel. In addition, Table A-4
identifies a potential accident spectrum associated with these facilities for this case. Attachment A,
Table A-2, lists the references for the source terms considered in analyzing potential accidents under
this alternative case, as well as the estimated frequency of occurrence for each accident. Table 5-21
lists the potential radiological accidents and health effects associated with dry storage of spent nuclear
fuel for the Decentralization alternative. For the transition period of wet to dry storage, Table 5-22
lists the accident scenario with the highest overall point estimate of risk to the general public.
Table 5-22 lists the accident scenario with the highest point estimate of risk (after transition) to the
general public when the fuel had been moved from wet storage (after approximately 15 years) and
placed in interim dry storage. This indicates a substantial reduction in risk (more than six orders of
magnitude) when fuel handling events are no longer potential accident initiators.
Table 5-22. Highest point estimates of risk among receptor groups (Option 2a).
Receptor Groups
Maximally Exposed Population to 80 kilometers
Offsite Individual
Overall Point Estimate of Riska 1.6x10-7 (Fuel Assembly 1.4x10-3 (Fuel Assembly
Breach) Breach)
Transitioned to Dry Storage 1.5x10-12 (Dry Vault Material 4.9x10-9 (Dry Vault Material
Point Estimate of Riska Release) Release)
a. Units of latent fatal cancers per year.
5.15.2.2.2 Option 2b - Wet Storage - DOE estimated potential radiological accident
impacts that could occur under this case using existing DOE-approved safety analysis reports and
amendments submitted to DOE by Westinghouse Savannah River Company for existing wet storage
facilities.
As indicated in Attachment A, Table A-4, the facilities (modules as defined in the WSRC
1994b and Figure 3-2) would consist of existing facilities and specific upgrades necessary to support
safe interim wet storage. In addition, Table A-4 identifies the reference accident spectrum associated
with these facilities for this option. Attachment A, Table A-2, lists the references for the source terms
considered in analyzing potential accidents under this alternative option, as well as the estimated
frequency of occurrence for each accident. Table 5-21 lists the radiological accidents and
consequences of the wet storage (Option 2b) of spent nuclear fuel for the Decentralization alternative.
Table 5-23 lists the accident scenario with the highest point estimate of risk to the general public. For
wet pool storage options, there are no transition phases.
Table 5-23. Highest point estimates of risk among receptor groups (Option 2b).
Receptor Groups
Maximally Exposed
Offsite Individual Population to 80 kilometers
Overall Point Estimate of Riska 1.6x10-7 (Fuel Assembly 1.4x10-3 (Fuel Assembly
Breach) Breach)
a. Units of latent fatal cancers per year.
5.15.2.2.3 Option 2c - Processing and Storage - Processing for the SRS is defined
as the operation of the separations facilities in F- or H-Areas.
The H-Area facilities were designed to
recover uranium and plutonium from spent production reactor fuel, and the F-Area facilities were
designed to recover plutonium.
DOE estimated potential radiological accident impacts that could occur under this option using
existing DOE-approved safety analysis reports submitted to DOE by Westinghouse Savannah River
Company for processes and for vault storage of special nuclear material from existing facilities. DOE
also considered radiological accidents associated with wet storage, because the spent nuclear fuel is
currently in wet storage. Similarly, it included fuel handling accidents throughout the processing
phase (i.e., until special nuclear material is in interim dry storage). As indicated in Attachment A,
Table A-4, the facilities required under this option would consist of existing and new facilities
necessary to support safe handling and processing of spent nuclear fuel into special nuclear material
for dry storage. In addition, Table A-4 identifies the reference accident spectrum associated with these
facilities for this case. Attachment A, Table A-2, lists the references for the source terms considered
in analyzing potential accidents under this alternative case, as well as the estimated frequency of
occurrence for each accident. Table 5-21 lists the radiological release accidents and health effects for
the processing of spent nuclear fuel to special nuclear material for the Decentralization alternative.
Table 5-24 lists the accident scenario with the highest overall point estimate of risk to the general
public from the transition period of wet spent fuel storage into processing for special nuclear material.
When the fuel had been processed from wet storage to special nuclear material and placed in its
interim dry storage, Table 5-24 lists the accident scenario with the highest point estimate of risk after
transition to the general public. This indicates a substantial reduction in risk (more than six orders of
magnitude) when fuel handling events and processing events are no longer potential accident initiators.
Table 5-24. Highest point estimates of risk among receptor groups (Option 2c).
Receptor Groups
Maximally Exposed
Offsite Individual Population to 80 kilometers
Overall Point Estimate of Riska 1.6x10-7 (Fuel Assembly 1.4x10-3 (Fuel Assembly
Breach) Breach)
Transitioned to Dry Storage 1.5x10-12 (Dry Vault Material 4.9x10-9 (Dry Vault Material
Point Estimate of Riska Release) Release)
a. Units of latent fatal cancers per year.
For this option, DOE assumes it could not process some fuel clad in stainless steel or zirconium
into special nuclear material and, therefore, would dry-store it as fuel. The technology for dry storage
of nonaluminum-clad fuel has been demonstrated and is assumed to pose no greater risk than
monitored dry storage of special nuclear material.
5.15.2.3 Alternative 3 - 1992/1993 Planning Basis. Because this alternative would be
consistent with the status quo at the SRS, existing documents contain sufficient information to
examine its accident analysis impacts. The SRS would continue to receive the spent nuclear fuel
designated for the Site, and DOE would complete facilities already planned to accommodate the
existing inventory and the spent nuclear fuel receipts. This alternative would require the same
facilities already used to support the cases discussed in the Section 5.15.2.2. The major difference
would be the amount of fuel ultimately stored because this alternative assumes the continued receipt of
fuel beyond that shipped to the SRS under the Decentralization alternative.
5.15.2.3.1 Option 3a - Dry Storage - DOE estimated potential radiological accident
impacts that could occur under this case using existing DOE-approved safety analysis reports for vault
storage from existing facilities and the study discussed for Option 2a.
DOE also considered
radiological accidents associated with wet storage, at least in the near term, because the spent nuclear
fuel is currently in wet storage. Similarly, it included fuel handling accidents throughout the transition
phase (i.e., until the fuel is in interim dry storage). As indicated in Attachment A, Table A-4, the
facilities required under this option would consist of existing and new facilities necessary to support
the safe handling and stabilization of spent nuclear fuel for dry storage. In addition, Table A-4
identifies the reference accident spectrum associated with these facilities for this case. Attachment A,
Table A-2, lists the authorization basis references for the source terms considered in analyzing
potential accidents under this option, as well as the estimated frequency of occurrence for each
accident. Table 5-21 lists the radiological release accidents and health effects for the dry storage of
spent nuclear fuel for the 1992/1993 Planning Basis alternative. For the entire period, the accident
scenarios with the highest point estimates of risk to the general public would be the same as those for
Option 2a, as listed in Table 5-22.
5.15.2.3.2 Option 3b - Wet Storage - DOE estimated potential radiological accident
impacts that could occur under this case using existing DOE-approved safety analysis reports and from
amendments submitted to DOE by Westinghouse Savannah River Company for wet storage for
existing facilities.
As indicated in Attachment A, Table A-4, the facilities required under this option
would consist of existing facilities and upgrades necessary to support safe interim wet storage. In
addition, Table A-4 identifies the reference accident spectrum associated with these facilities for this
option. Attachment A, Table A-2, lists the references for the source terms considered in analyzing
potential accidents under this option, as well as the estimated frequency of occurrence for each
accident. Table 5-21 lists the radiological release accidents and health effects of the wet storage
(Option 3b) of spent nuclear fuel for the 1992/1993 Planning Basis alternative. The accident scenario
with the highest point estimate of risk to the general public would be the same as that for Option 2b,
as listed in Table 5-23.
5.15.2.3.3 Option 3c - Processing and Storage.
Table 5-21 lists the radioactive
release accidents and health effects for the processing of spent nuclear fuel for this option. After
processing is complete, the accident scenario with the highest point estimate of risk would be
associated with the storage of special nuclear materials, as discussed for Option 2c and listed in
Table 5-24.
5.15.2.4 Alternative 4 - Regionalization. This alternative comprises Regionalization A and
Regionalization B subalternatives. Under the Regionalization A subalternative (Options 4a, 4b, and
4c), the SRS would receive all aluminum-clad fuel from the other sites considered in this EIS and
would transfer its existing inventory of stainless steel- and Zircaloy-clad fuel to other DOE sites, as
appropriate. These proposed activities would reflect current and past activities, so sufficient
information and analyses are available to enable the scaling or other extrapolation of radiological
accident impacts. The total amount of spent nuclear fuel to be managed under Regionalization A
would be slightly less than that for Alternatives 2 and 3; the decisionmaker could use this amount to
adjust the estimated point estimate of risk by the use of an appropriate adjustment (scaling) factor, as
discussed in Attachment A, Section A.2.9.
Under the Regionalization B subalternative (Options 4d, 4e, 4f, and 4g), the SRS would receive
all existing and new spent nuclear fuel east of the Mississippi River. The decisionmaker could use the
change in spent nuclear fuel inventories to adjust the estimated point estimate of risk by the use of an
appropriate adjustment (scaling) factor, as discussed in Attachment A, Section A.2.9. For the purposes
of this evaluation, Option 4g (Section 5.15.2.4.7) assumes that DOE would ship all fuel off the Site to
the Oak Ridge Reservation.
5.15.2.4.1 Option 4a - Dry Storage - This case is similar to Option 2a, with the
exception of the quantity and type of fuel to be stored.
As with Option 2a, this assessment evaluated
existing analyses; the point estimates of risk are the same as those for Option 2a.
5.15.2.4.2 Option 4b - Wet Storage - This case is similar to Option 2b, with the
exception of a slightly smaller quantity of fuel to be stored.
As with Option 2b, this assessment
evaluated existing analyses, and the point estimates of risk are the same as those for Option 2b.
5.15.2.4.3 Option 4c - Processing and Storage - For this option, the accident
analysis evaluation is similar to Option 2c.
DOE assumes that it could process spent nuclear fuel
associated with regionalization at SRS with existing facilities, because they are designed to process
aluminum-clad fuel. However, the small amount of aluminum-clad fuel received after major
processing options are completed would be placed in wet storage.
5.15.2.4.4 Option 4d - Dry Storage - The accident analysis evaluation for this option
is similar to that for Option 2a, with the exception of the increased inventories and types of fuel to be
stored.
5.15.2.4.5 Option 4e - Wet Storage - The accident analysis evaluation for this option
is similar to that for Option 2b, with the exception of the increased inventories and types of fuel to be
stored.
5.15.2.4.6 Option 4f - Processing and Storage - For this option, the accident
analysis evaluation is similar to Option 2c.
DOE assumes that it could process all the current SRS
aluminum-clad spent nuclear fuel with existing facilities. However, all receipts of spent nuclear fuel
will be placed in dry storage as discussed for Option 4d.
5.15.2.4.7 Option 4g - Shipping Off Site - This option assumes that DOE would
characterize the fuel and ship it all off the Site.
Thus, the potential radiological accidents considered
are the same as those for Alternative 1.
5.15.2.5 Alternative 5 - Centralization. This alternative for the SRS would involve fuel
types and new facilities beyond those considered for any other alternative. For instance, under this
alternative, the SRS would receive spent nuclear fuel from the U.S. Navy. One of the new facilities
that would be necessary to support this type of spent nuclear fuel is the Expended Core Facility (ECF).
Volume 1, Appendix D, includes a detailed accident analyses for this proposed facility using
SRS-specific parameters.
This alternative would bound the maximum number of spent nuclear fuel-related accident
scenarios that DOE could expect at the SRS, due to the number of new facilities at the Site that would
have to accommodate the diversity and the increased amount of the fuel to be managed. The
decisionmaker could use this maximum amount of spent nuclear fuel to adjust the estimated risk by
the use of an appropriate scaling factor, as discussed in Attachment A, Section A.2.9. For the
purposes of this evaluation, Option 5d (Section 5.15.2.5.4) assumes that DOE would ship all fuel off
the Site to another DOE facility.
5.15.2.5.1 Option 5a - Dry Storage - The major difference in dry storage facilities
between this alternative and the others would be the addition of a facility for Naval spent nuclear fuels
and the large quantity of spent fuel shipped to the SRS from the Hanford Site.
DOE estimated
potential radiological accident impacts that could occur under this option using DOE-approved safety
analysis reports submitted to DOE by Westinghouse Savannah River Company for vault storage in
existing facilities at the SRS and the study discussed for Option 2a. In addition, DOE considered
radiological accidents associated with wet storage, at least in the near term, because the SRS spent
nuclear fuel is currently in wet storage. Similarly, it included fuel handling accidents throughout the
transition phase (i.e., until fuel is in interim dry storage). As indicated in Attachment A, Table A-4,
the facilities required under this option would consist of existing and new facilities necessary to
support the safe handling and stabilization of spent nuclear fuel for dry storage. In addition,
Table A-4 identifies the reference accident spectrum associated with these facilities for this case.
Attachment A, Table A-2, lists the references for the source terms considered in analyzing potential
accidents under this option, as well as the estimated frequency of occurrence for each accident.
Table 5-21 compares the radiological release accidents and health effects for the dry storage of spent
nuclear fuel for the Centralization alternative. From the transition period of wet to dry storage, the
accident scenario with the highest point estimate of risk to the general public would be the same as
that for Option 2a, as listed in Table 5-22. When the fuel had been moved from wet storage (after
approximately 25 years) and placed in interim dry storage, the accident scenario with the highest point
estimate of risk to the population would be the same as the Option 2a dry storage phase.
5.15.2.5.2 Option 5b - Wet Storage - The accident analysis evaluation for this option
is similar to that for Option 2b, with the exception of the amount and type of fuel to be stored.
5.15.2.5.3 Option 5c - Processing and Storage - For this option, the accident
analysis evaluation is similar to Option 2c.
DOE assumes that it could process the current SRS
aluminum-clad spent nuclear fuel with existing facilities. However, the SRS would place all receipts
of fuel in dry storage, as discussed for Option 5a.
5.15.2.5.4 Option 5d - Shipping Off Site - This option assumes that DOE would
perform the characterization of the fuel at the SRS, and then would ship all fuel off the Site.
Thus,
the potential radiological accidents considered are the same as those for the No Action alternative.
5.15.3 Chemical Hazard Evaluation
For toxic chemicals, several government agencies recommend the quantification of health effects
as threshold values of concentrations in air or water that cause short-term effects. The long-term
health consequences of human exposure to toxic chemicals are not as well understood as those for
radiation. Thus, the potential health effects from toxic chemicals are more subjective than those from
radioactive materials.
This section provides a quantitative discussion for an analyzed chemical accident at the
Receiving Basin for Offsite Fuel facility and qualitative discussions addressing chemical hazards for
each of the other existing SRS facilities involved in the receipt, processing, transport, or storage of
spent nuclear fuel.
5.15.3.1 Receiving Basin for Offsite Fuel. The maximum reasonably foreseeable chemical
hazard accident for the Receiving Basin for Offsite Fuel would involve the release of nitrogen dioxide
vapor following the complete reaction of a drum of target cleaning solution (13.4 percent nitric acid)
with sodium nitrite (WSRC 1993b). The initiator for this accident is a leak from a storage tank into
the target cleaning solution and involves multiple failures or maloperations with an accident
probability comparable to that of a natural phenomena accident. Table 5-25 shows the concentration
of nitrogen dioxide vapor that an individual at the SRS boundary and a maximally exposed colocated
worker could receive.
Table 5-25. Results of analyzed chemical accident.
Receptor Group Frequency NO2 Concentration