APPENDIX B Idaho National Engineering Laboratory 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 I
Appendix B
idaho National Engineering Laboratory
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 Overview 2-1
2.1.1 History of Spent Nuclear Fuel Activities 2-1
2.1.2 Current Activities at Spent Nuclear Fuel-Related Facilities 2-3
2.1.3 Spent Nuclear Fuel Mission 2-8
2.2 Regulatory Framework for Spent Nuclear Fuel Management 2-9
2.3 Spent Nuclear Fuel Management Program at the INEL 2-10
3. Spent Nuclear Fuel Management Alternatives 3-1
3.1 Description of Alternatives 3-1
3.1.1 Alternative I: No Action 3-6
3.1.2 Alternative 2: Decentralization 3-9
3.1.3 Alternative 3: 199211993 Planning Basis 3-10
3.1.4 Alternative 4: Regionalization 3-13
3.1.5 Alternative 5: Centralization 3-15
3.2 Comparison of Alternatives 3-18
4. Affected Environment 4.1-1
4.1 Overview 4.1-1
4.2 Land Use 4.2-1
4.2.1 Existing and Planned Land Uses at the INEL 4.2-1
4.2.2 Existing and Planned Land Use in Surrounding Areas 4.2-3
4.3 Socioeconomics 4.3-1
4.3.1 Employment 4.3-1
4.3.2 Population and Housing 4.3-2
4.3.3 Community Services 4.3-4
4.3.4 Public Finance 4.3-8
4.4 Cultural Resources 4.4-1
4.4.1 Archeological Sites and Historic Structures 4.4-1
4.4.2 Native American Cultural Resources 4.4-3
4.4.3 Paleontological Resources 4.4-3
4.5 Aesthetic and Scenic Resources 4.5-1
4.5.1 Visual Character of the INEL Site 4.5-1
4.5.2 Scenic Areas 4.5-1
4.6 Geology 4.6-1
4.6.1 General Geology 4.6-1
4.6.2 Natural Resources 4.6-4
4.6.3 Seismic Hazards 4.6-4
4.6.4 Volcanic Hazards 4.6-7
4.7 Air Quality 4.7-1
4.7.1 Climatology and Meteorology 4.7-1
4.7.2 Air Quality 4.7-2
4.8 Water Resources 4.8-1
4.8.1 Surface Water 4.8-1
4.8.2 Subsurface Water 4.8-4
4.8.3 Water Use and Rights 4.8-13
4.9 Ecological Resources 4.9-1
4.9.1 Flora 4.9-1
4.9.2 Fauna 4.9-2
4.9.3 Threatened, Endangered, and Sensitive Species 4.9-3
4.9.4 Wetlands 4.9-3
4.10 Noise 4.10-1
4.11 Traffic and Transportation 4.11-1
4.11.1 Roadways 4.11-1
4.11.2 Raikoads 4.114
4.11.3 Airports and Air Traffic 4.114
4.11.4 Accidents 4.11-5
4.11.5 Transportation of Waste, Materials, and Spent Nuclear Fuel 4.11-5
4.12 Occupational and Public Health and Safety 4.12-1
4.12.1 Radiological Health and Safety 4.12-1
4.12.2 Nonradiological Exposure and Health Effects 4.12-2
4.12.3 Occupational Health and Safety 4.12-2
4.13 Idaho National Engineering Laboratory Services 4.13-1
4.13.1 Water Consumption 4.13-1
4.13.2 Electricity Consumption 4.13-1
4.13.3 Fuel Consumption 4.13-2
4.13.4 Wastewater Disposal 4.13-2
4.13.5 Security and Emergency Protection 4.13-3
4.14 Materials and Waste Management 4.14-1
4.14.1 High-Level Waste 4.14-1
4.14.2 Transuranic Waste 4.14.2
4.14.3 Mixed Low-Level Waste 4.14.2
4.14.4 Low-Level Waste 4.14.2
4.14.5 Hazardous Waste 4.14.3
4.14.6 Industdal/Commerciai Solid Waste 4.14.3
4.14.7 Hazardous Materials 4.14.3
5. Environmental Consequences 5.1-1
5.1 Overview 5.1-1
5.2 Land Use 5.2-1
5.3 Socioeconomics 5.3-1
5.3.1 Methodology 5.3-1
5.3.2 Alternatives 1 and 2 - No Action and Decentralization 5.3-2
5.3.3 Alternatives 3, 4a, 4b(1), and Sb - 1992/1993 Planning Basis,
Regionalization by Fuel Type, Regionalization by Geography
(INEL), and Centralization at the INEL 5.3-3
5.3.4 Alternatives 4b(2) and Sa - Regionalization by Geography (Elsewhere)
and Centralization at Other DOE Sites 5.3-3
5.4 Cultural Resources 5.4-1
5.5 Aesthetic and Scenic Resources 5.5-1
5.6 Geology 5.6-1
5.7 Air Quality and Related Consequences 5.7-1
5.7.1 Alternative 1 - No Action 5.7-1
5.7.2 Alternative 2 - Decentralization 5.7-3
5.7.3 Alternative 3 - 1992/1993 Planning Basis 5.7-5
5.7.4 Alternative 4a - Regionalization by Fuel Type 5.7-6
5.7.5 Alternative 4b(l) - Regionalization by Geography (INEL)5.7-6
5.7.6 Alternative 4b(2) - Regionalization by Geography
(Elsewhere) 5.7-7
5.7.7 Alternative 5a - Centralization at Other DOE Sites 5.7-8
5.7.8 Alternative Sb - Centralization at the INEL . . 5.7-9
5.8 Water Resources and Related Consequences 5.8-1
5.9 Ecology 5.9-1
5.10 Noise 5.10-1
5.11 Traffic and Transportation 5.11-1
5.11.1 Introduction 5.11-1
5.11.2 Methodology 5.11-1
5.11.3 Onsite Spent Nuclear Fuel Shipments 5.11-2
5.11.4 Incident-Free Impacts 5.11-3
5.11.5 Accident Impacts 5.11-4
5.11.6 Onsite Mitigative and Preventative Measures 5.11-6
5.12 Occupational and Public Health and Safety 5.12-1
5.12.1 Radiological Exposure and Health Effects 5.12-1
5.12.2 Nonradiological Exposure and Health Effects 5.12-4
5.12.3 Industrial Safety 5.12-5
5.13 Idaho National Engineering Laboratory Services 5.13-1
5.13.1 Construction 5.13-1
5.13.2 Operations 5.13-2
5.14 Materials and Waste Management 5.14-1
5.14.1 Alternative 1 - No Action 5.14-1
5.14.2 Alternative 2 - Decentralization 5.14-1
5.14.3 Alternative 3 - 199211993 Planning Basis 5.14-1
5.14.4 Alternative 4a - Regionalization by Fuel Type 5.14-5
5.14.5 Alternative 4b(1) - Regionalization by Geography (INEL) 5.14-5
5.14.6 Alternative 4b(2) - Regionalization by Geography (Elsewhere)5.14-6
5.14.7 Alternative 5a - Centralization at Other DOE Sites 5.14-6
5.14.8 Alternative Sb - Centralization at the INEL 5.14-6
5.15 Accidents 5.15-1
5.15.1 Introduction 5.15-1
5.15.2 Historic Perspective 5.15-2
5.15.3 Methodology for Determining the Maximum Reasonably Foreseeable
Radiological Accidents 5.15-13
5.15.4 Impacts from Postulated Maximum Reasonably Foreseeable
Radiological Accidents 5.15-24
5.15.5 Impacts from Postulated Maximum Reasonably Foreseeable
Toxic Material Accidents 5.15-40
5.15.6 Maximum Reasonably Foreseeable Radiological Accident Scenario
Descriptions 5.15-50
5.16 Cumulative Impacts and Impacts from Connected or Similar Actions 5.16-1
5.16.1 Land Use 5.16-1
5.16.2 Socioeconomics 5.16-5
5.16.3 Cultural Resources 5.16-5
5.16.4 Air Quality 5.l6-6
5.16.5 Occupational and Public Health and Safety 5.16-6
5.16.6 Materials and Waste Management 5.16-7
5.17 Adverse Environmental Effects That Cannot be Avoided 5.17-1
5.18 Relationship Between Short-Term Use of the Environment and the
Maintenance of Long-Term Productivity 5.18-1
5.19 Irreversible and Irretrievable Commitment of Resources 5.19-1
5.20 Potential Mitigation Measures 5.20-1
5.20.1 Pollution Prevention 5.20-1
5.20.2 Cultural Resources 5.20-1
5.20.3 Traffic and Transportation 5.20-2
5.20.4 Accidents 5.20-3
6. References 6-1
TABLES
2-1. INEL spent nuclear fuel history 2-2
2-2. Major INEL spent-nuclear fuel storage facilities 2-6
3-1. Summary of spent nuclear fuel management alternatives at the Idaho National
Engineering Laboratory 3-2
3-2. Potential spent nuclear fuel projects required for each alternative 3-5
3-3. Spent nuclear fuel inventory for each alternative by 2035 (metric tons of heavy metal) 3-7
3-4. Comparison of impacts from construction 3-20
3-5. Comparison of impacts from normal operations 3-23
3-6. Comparison of impacts from accidents 3-26
4.3-1. Projected labor force, employment, and population for the INEL region of
influence, 1995-2004 4.3-2
4.3-2. Number of housing units, vacancy rates, median house value, and niedian
monthly rent by county and region of influence 4.3-6
4.3-3. Summary of public services available in the region of influence 4.3-7
4.3-4. Total revenues and expenditures by county, Fiscal Year 1991 4.3-8
4.4-1. Plants used by the Shoshone-Bannock Tribes that are located on or near the INEL 4.4-4
4.7-1. Baseline annual average and maximum hourly emission rates of nonradiological
air pollutants at the INEL 4.7-5
4.7-2. Comparison of baseline ambient air concentrations with most stringent applicable
regulations and guidelines at the INEL 4.7-7
4.7-3. Summary of airborne radionuclide emissions from INEL facility areas
(curies per year) 4.7-9
4.8-1. Highest detected contaminant concentrations in groundwater at the Idaho National
Engineering Laboratory (1987 to 1992) 4.8-10
4.9-1. Threatened and endangered species, special species of concern, and sensitive
species that may be found on the INEL 4.9-4
4.11-1. Baseline traffic for selected highway segments 4.11-3
4.11-2. Baseline annual vehicle miles traveled for Idaho National Engineering
Laboratory-related traffic 4.11-3
4.11-3. Loaded rail shipments to and from the Idaho National Engineering
Laboratory (1988-1992) 4.11-4
4.11-4. Cumulative doses and cancer fatalities from incident-free onsite shipments
of nonnaval spent nuclear fuel at the Idaho National Engineering Laboratory
for 1995 through 2035 4.11-6
5.3-1. Estimated changes in employment and population for Alternatives 3, 4a, 4b(1),
and 5b, 1995 - 2004 5.3-2
5.3-2. Estimated changes in employment and population for Alternatives 4b(2)
and 5a, 1995 - 2004 5.3-2
5.6-1. Estimated INEL gravel/borrow use (cubic meters) 5.6-1
5.7-1. Maximum impacts to nonradiological air quality from spent nuclear fuel-criteria
pollutants 5.7-2
5.7-2. Maximum impacts to nonradiological air quality from spent nuclear fuel-toxic
air pollutants 5.7-2
5.7-3. Annual dose increments by alternative in comparison to the baseline 5.7-3
5.7-4. Radionuclide emissions by alternative for spent nuclear fuel projects 5.7-4
5.11-1. Impacts from maximum reasonably foreseeable spent nuclear fuel transportation
accident on INEL (using generic rural and suburban population densities) 5.11-5
5.12-1. Annual Occupational radiation exposure and employment summary 5.12-2
5.12-2. Annual nonoccupational radiation exposure summary 5.12-2
5.12-3. Annual fatal cancer incidence and probability summary from radiological exposure 5.12-2
5.12-4. 40-year fatal cancer incidence summary from radiological exposure 5.12-3
5.12-5. Annual industrial safety health effects incidence summary 5.12-5
5.13-1. Estimated increase in annual electricity, water, wastewater treatment, and fuel
requirements for construction activities associated with each alternative 5.13-1
5.13-2. Estimated increase in annual electricity, water, wastewater treatment, and fuel
requirements for operations activities associated with each alternative 5.13-2
5.14-1. Average annual waste generation projections for selected SNF management
alternatives at INEL 5.14-2
5.14-2. Peak waste generation highlights for selected SNF management alternatives
at INEL 5.14-3
5.15-1. Summary of radiological accidents for worker located 100 meters downwind
from the point of release 5.15-3
5.15-2. Summary of radiological accidents for individual located at the nearest
point of public access within the site boundary 5.15-5
5.15-3. Summary of radiological accidents for maximally exposed hypothetical
individual located at the nearest site boundary 5.15-7
5.15-4. Summary of radiological accidents for offsite population within
80 kilometers (50 miles) from the point of release 5.15-9
5.15-5. Accident frequency categones 5.15-17
5.15-6. Determination of accident frequency adjustment factors for Alternatives 2
through 5 based on estimated number of annual spent nuclear fuel shipments
under each alternative 5.15-23
5.15-7. Impacts from selected maximum reasonably foreseeable radiological accidents
Alternative 1, No Action (50 and 95 percentile meteorological conditions) 5.15-26
5.15-8. Estimated secondary impacts resulting from the maximum reasonably foreseeable
accidents postulated under Alternative 1, No Action, assuming conservative
(95 percentile) meteorological conditions 5.15-28
5.15-9. Impacts from selected maximum reasonably foreseeable accidents - Alternative 2,
Decentralization (50 and 95 percentile meteorological conditions) 5.15-29
5.15-10. Impacts from selected maximum reasonably foreseeable accidents - Alternative 3,
Planning Basis (50 and 95 percentile meteorological conditions) 5.15-31
5.15-11. Impacts from selected maximum reasonably foreseeable accidents - Alternative 4a,
Regionalization by Fuel Type (50 and 95 percentile meteorological conditions) 5.15-33
5.15-12. Impacts from selected maximum reasonably foreseeable accidents - Alternative 4b(l),
Regionalization by Geography (INEL) (50 and 95 percentile meteorological
conditions) 5.15-35
5.15-13. Impacts from selected maximum reasonably foreseeable accidents - Alternative 4b(2),
Regionalization by Geography (Elsewhere) (50 and 95 percentile meteorological
conditions) 5.15-36
5.15-14. Impacts from selected maximum reasonably foreseeable accidents - Alternative Sa,
Centralization at Other DOE Sites (50 and 95 percentile meteorological
conditions) 5.15-38
5.15-15. Impacts from selected maximum reasonably foreseeable accidents - Alternative Sb,
Centralization at the INEL (50 and 95 percentile meteorological conditions) 5.15-39
5.15-16. Summary of chemical concentrations for postulated nonprocessing-related accidental
releases at the Idaho Chemical Processing Plant under Alternatives 1 through 5 5.15-45
5.15-17. Summary of chemical concentrations for postulated processing-related accidental
releases at the Idaho Chemical Processing Plant under Alternatives 4b(l) and Sb 5.l5-48
5.16-1. Nonhealth-related cumulative impacts 5.16-2
5.16-2. Health-related cumulative impacts 5.16-3
FIGURES
2-1. Major facility areas located at the Idaho National Engineering Laboratory site 2-4
2-2. Existing (1995) distribution of INEL SNF 2-7
4.2-1. Selected land uses at the INEL and in the surrounding region 4.2-2
4.3-1. Historic and projected baseline employment at the Idaho National Engineering Laboratory,
1990-2004 4.3-3
4.3-2. Historic and projected total population for the counties of the region of influence,
1940 through 2004 4.3-5
4.6-1. Location of INEL in context of regional geologic features 4.6-2
4.6-2. Lithologic logs of deep drill holes in the INEL area 4.6-3
4.6-3. Earthquakes with magnitudes greater than 2.5 from 1884 to 1989 4.6-5
4.6-4. Contribution of the seismic sources to the mean peak acceleration at the
Idaho Chemical Processing Plant 4.6-8
4.6-5. Map of the INEL showing locations of volcanic rift zones and lava flow hazard
zones 4.6-10
4.7-1. Depiction of annual average wind direction and speed at INEL meteorological
monitoring stations 4.7-3
4.7-2. Comparison of dose to maximally exposed individual to the National Emission
Standard for Hazardous Air Pollutants dose limit and the dose from background
sources 4.7-11
4.8-1. Selected facilities and predicted inundation map for probable maximum flood-induced
overtopping failure of Mackay Dam at the INEL 4.8-2
4.8-2. Location of the INEL, Snake River Plain, and generalized groundwater flow
direction of the Snake River Plain Aquifer 4.8-5
4.8-3. Hydrostratigraphy across the INEL and water table surface 4.8-7
4.11-1. Transportation routes in the vicinity of the INEL 4.11-2
5.3-1. INEL employment by SNF alternative relative to site employment projections 5.3-4
5.15-1. Comparison of fatality rates among workers in various industry groups 5.15-11
1. INTRODUCTION
The U.S. Department of Energy (DOE) has prepared the Department of Energy Programmatic
Spent Nuclear Fuel Management and Idaho National Engineering Laboratory Environmental
Restoration and Waste Management Programs Environmental Impact Statement (SNF and INEL EIS)
to assist its management in making two decisions. The first decision, which is programmatic, is to
determine the management program for DOE spent nuclear fuel. The second decision is on the future
direction of environmental restoration, waste management, and spent nuclear fuel management
activities at the Idaho National Engineering Laboratory.
Volume 1 of the EIS, which supports the programmatic decision, considers the effects of spent
nuclear fuel management on the quality of the human and natural environment for planning years 1995
through 2035. DOE has derived the information and analysis results in Volume 1 from several site-
specific appendixes. Volume 2 of the EIS, which supports the INEL-specific decision, describes
environmental impacts for various environmental restoration, waste management, and spent nuclear
fuel management alternatives for planning years 1995 through 2005.
This Appendix B to Volume 1 considers the impacts on the INEL environment of the
implementation of various DOE-wide spent nuclear fuel management alternatives. The Naval Nuclear
Propulsion Program, which is a joint Navy/DOE program, is responsible for spent naval nuclear fuel
examination at the INEL. For this appendix, naval fuel that has been examined at the Naval Reactors
Facility and turned over to DOE for storage is termed naval-type fuel. This appendix evaluates the
management of DOE spent nuclear fuel including naval-type fuel. Naval spent nuclear fuel
examination is addressed in Appendix D; Section 5.16 of this appendix includes relevant
environmental consequences from Appendix D.
In addition to this introduction, Appendix B contains the following chapters:
- Chapter 2 - Background: Describes INEL spent nuclear fuel facilities, the regulatory
framework for spent nuclear fuel management at the INEL, and the INEL spent nuclear fuel
management program.
- Chapter 3 - Spent Nuclear Fuel Management Alternatives: Describes the DOE-wide spent
nuclear fuel management alternatives as the INEL would implement them, and provides a
summary comparison of potential environmental consequences for each alternative, as
described in Chapter 5.
- Chapter 4 - Affected Environment: Describes the INEL site and the surrounding
environment that DOE spent nuclear fuel management actions could affect.
- Chapter 5 - Environmental Consequences: Provides the results of environmental
consequence analyses for each spent nuclear fuel management alternative.
- Chapter 6 - References
Volume 1 contains a list of acronyms and abbreviations and a glossary that is applicable to this
appendix.
2. BACKGROUND
This chapter contains an overview of the Idaho National Engineering Laboratory (INEL) facilities
and historic events related to spent nuclear fuel, a description of the regulatory framework for the
actions evaluated in this document, and an overview of the current spent nuclear fuel management
program at the INEL.
2.1 Overview
The following sections provide a general overview of the INEL including its history, current
activities, and mission as they relate to spent nuclear fuel management and future decisions.
2.1.1 History of Spent Nuclear Fuel Activities
The U.S. Atomic Energy Commission, a predecessor of the U.S. Department of Energy (DOE),
established the INEL, formerly the National Reactor Testing Station, to build, test, and operate various
types of nuclear reactors, support plants, and associated equipment. Since its establishment in 1949
(see Table 2-1), DOE and its predecessor agencies have built 52 reactors at the INEL. The major
DOE programs at the site have included test irradiation services, uranium recovery from highly
enriched spent fuels, calcination of liquid radioactive waste, light-water-cooled reactor safety testing
and research, operation of research reactors, environmental restoration, and storage and surveillance of
solid transuranic wastes. In support of the DOE reactor research program and as part of the spent
nuclear fuel reprocessing program, the INEL has received spent nuclear fuel from more than 30 offsite
sources, including naval reactors, university reactors, commercial reactors, and DOE research reactors,
as well as fuels fabricated in the United States and irradiated in foreign reactors (DOE 1993).
The Experimental Breeder Reactor-I, now a National Historic Landmark, maintains a key place
in the history of nuclear power in the United States. In December 1951, this reactor generated the first
usable electricity from a nuclear reactor. The Experimental Breeder Reactor-I also demonstrated that a
nuclear reactor could actually produce more fuel than it consumes.
Of special significance to spent nuclear fuel is the history of the Idaho Chemical Processing
Plant. From 1953 to 1992, this plant recovered usable uranium from spent nuclear fuel from United
States government reactors. The plant operated for 39 years as a full-scale production facility. But in
Table 2-1. INEL spent nuclear fuel history.
Year Event
1949 National Reactor Testing Station established
1951 Site reactor first to generate electricity from nuclear fission
1953 ICPPa began operation
1953 Test of first submarine nuclear reactor
1957 Expended Core Facility constructed
1965 DOE contract with Public Service Company of Colorado (Fort
St. Vrain)
1974 Site became Idaho National Engineering Laboratory
1980 DOE contracted to receive Public Service Company of Colorado
(Fort St. Vrain) spent nuclear fuel
1992 Decision to discontinue reprocessing of spent nuclear fuel at ICPPa
announced
1992 DOE creates Office of Spent Fuel Management
1993 Court order of June 28, 1993 issued
a. ICPP = Idaho Chemical Processing Plant.
April 1992, DOE decided to phase out reprocessing for material recovery, resulting in the shutdown of
the reprocessing operation.
Spent naval nuclear fuel handling at the Naval Reactors Facility originated in 1957 with the
construction of the Expended Core Facility. The original building contained a water pit and shielded
cells, which are connected to the water pit by transfer tunnels. The Expended Core Facility examines
spent nuclear fuel from operating naval ships and from prototype naval reactors. The examinations
support research and development for naval fuel quality improvement. Over the years, the Navy made
additions and improvements at the Naval Reactors Facility site, including the construction and
operation of three prototype reactors and facilities for training naval nuclear powerplant operators.
The Naval Nuclear Propulsion Program is placing the prototype reactors, which have reached the ends
of their useful lives, in layup. All training is expected to end before DOE issues the Record of
Decision for this Environmental Impact Statement (EIS). Expended Core Facility activities are
continuing. Appendix D describes the Naval Reactors Facility in more detail.
In 1965 the United States entered into a contract with Public Service Company of Colorado, with
which the United States agreed to lease special nuclear material to Public Service Company of
Colorado for fuel at the Fort St. Vrain Nuclear Power Plant. In 1980, the United States and Public
Service Company of Colorado modified the 1965 contract, requiring DOE to accept returned Fort St.
Vrain spent nuclear fuel at the INEL. From 1980 to 1986, Public Service Company of Colorado made
approximately 120 shipments of Fort St. Vrain spent nuclear fuel to the INEL.
In 1974 the National Reactor Testing Station became the Idaho National Engineering Laboratory.
The INEL mission broadened to include research and engineering for nonnuclear programs and
environmental restoration and waste management activities.
In the early 1980s, pursuant to the West Valley Demonstration Project Act (42 USC 2021a) and
a court order, DOE agreed to accept 125 special case commercial reactor spent nuclear fuel assemblies
located at the state-owned Western New York Nuclear Service Center. DOE began a project to
demonstrate the viability of a transportable spent nuclear fuel storage cask, with the intention of
shipping the fuel to the INEL. Based on this, New York State Energy Research and Development
Authority, which has jurisdiction over the center, has allowed continued storage until DOE obtained
U.S. Nuclear Regulatory Commission Certificates of Compliance, which have been issued. The fuel
remains at West Valley awaiting the Record of Decision for this EIS.
In addition to the naval and INEL-generated fuel on the site, some special-case spent nuclear
fuel, such as fuel from university reactors, has been shipped directly to the Idaho Chemical Processing
Plant for storage. Damaged fuel from the 1979 Three Mile Island accident was shipped directly to
Test Area North for examination and storage as part of a research mission.
In 1990, DOE issued an Environmental Assessment and Finding of No Significant Impact for
Public Service Company of Colorado shipments of Fort St. Vrain spent nuclear fuel to the INEL. The
State of Idaho challenged the adequacy of the Environmental Assessment and, in June 1993, the
United States District Court for the District of Idaho found for the State and ordered DOE to prepare
this EIS. A DOE appeal of the order resulted in a December 1993 amendment that governs the DOE
schedule and obligation for preparing the EIS.
2.1.2 Current Activities at Spent Nuclear Fuel-Related Facilities
Six major facility areas at the INEL (Figure 2-1) store spent nuclear fuel: Argonne National
Laboratory - West, Idaho Chemical Processing Plant, Naval Reactors Facility, Power Burst Facility,
Figure 2-1. Major facility areas located at the Idaho National Engineering Laboratory site. Test Area North, and Test Reactor Area. Spent fuel at the INEL is kept in a variety of dry and wet
configurations. The total amount of spent nuclear fuel at the INEL accounts for about 10 percent (by
weight of heavy metal) of the spent nuclear fuel in the DOE complex (DOE 1993).
Table 2-2 lists the primary INEL spent nuclear fuel storage facilities, the types of fuel in storage,
and the storage configurations. Figure 2-2 indicates the relative proportion of fuel at these facilities.
The number and variety of wet and dry storage configurations currently in use at the INEL is largely
the result of the different purposes for the facilities (e.g., at-reactor storage, storage research and
development, reprocessing, and fuel research and development). The condition of the spent nuclear
fuel in storage is generally good with the notable exception of the fuel in the Underwater Fuel Storage
Facility (CPP-603). The following paragraphs briefly describe each primary facility area that manages
spent nuclear fuel.
The Argonne National Laboratory - West generates spent nuclear fuel as a result of research and
development activities related to advanced reactor design. DOE has brought small quantities of spent
nuclear fuel from other reactors to this facility to support these activities. Reactors at Argonne
National Laboratory - West are the Experimental Breeder Reactor II, the Transient Reactor Test
Facility, the Zero Power Physics Reactor, and the Neutron Radiography Reactor. Storage facilities
include both wet (including molten sodium) and dry configurations.
The Idaho Chemical Processing Plant historically received spent nuclear fuel from many onsite
and offsite reactors for reprocessing (i.e., the recovery of uranium for reuse). However, DOE decided
to phase out reprocessing activities in 1992. The new mission for this facility area is receipt and
storage, plus research and development of technologies in support of the disposition of spent nuclear
fuel. The Idaho Chemical Processing Plant stores virtually all types of spent nuclear fuel except
production reactor fuel [i.e., fuel from Hanford Site and Savannah River Site (SRS) production
reactors]. It stores nonproduction aluminum-based spent nuclear fuel. This facility uses both wet and
dry storage configurations.
The Naval Reactors Facility includes the Expended Core Facility, which receives and examines
naval spent nuclear fuel to support fuel development and performance analyses. In addition, the
Expended Core Facility removes structural support material from fuel assemblies before the transfer of
the fuel portion to the Idaho Chemical Processing Plant for interim storage.
Table 2-2. Major INEL spent nuclear fuel storage facilities.
Fuel Type(c)
Facility(a) Storage Type(b) 1 2 3 4 5 6a 6b 6c
Argonne National Laboratory - West
Experimental Breeder Reactor II Liquid sodium -
Hot Fuel Examination Facility Dry -
Neutron Radiography Reactor Wet -
Radioactive Scrap and Waste Facility Dry -
Transient Reactor Test Facility Dry -
Idaho Chemical Processing Plant
Underwater Fuel Storage Facilityd Wet - - - -
Irradiated Fuel Storage Facility Dry -
Fuel Storage Area/Fluorinel Dissolution Wet - - - -
Process Cell
Underground Storage Facility Dry -
Naval Reactors Facility
Expended Core Facility Wet - -
Expended Core Facility Rail Siding Dry -
Power Burst Facility
Power Burst Facility Storage Canal Wet -
Test Reactor Area
Materials Test Reactor Canal Wet - -
Advanced Reactivity Measurement Wet -
Facility
Coupled Fast Reactivity Measurement Wet -
Facility
Advanced Test Reactor Canal Wet -
Test Area North
Test Area North Pool Wet -
Test Area North Pad Dry -
a. This table lists the major spent fuel storage facilities. Other facilities (e.g., laboratories) might periodically
contain small quantities of spent nuclear fuel.
b. Wet storage involves water-filled pools. Dry storage involves a variety of configurations (e.g., casks, wells,
buildings).
c. The spent fuel types are as follows:
1. Naval-type fuel
2. Savannah River Site production fuels and other aluminum-clad fuels
3. Hanford Site production fuels
4. Graphite fuels
5. Special case commercial fuels
6a. Experimental reactors - stainless steel-clad fuels
6b. Experimental reactors - zirconium-clad fuels
6c. Experimental reactors - other fuel configurations
d. Spent nuclear fuel storage at this facility will cease by December 31, 2000, as part of an agreement between
DOE and the State of Idaho.
Figure 2-2. Distribution of INEL SNF. The Power Burst Facility reactor was placed in operational standby in 1992. A limited amount
of spent nuclear fuel from this facility remains in wet storage, in a storage pool that is in good
condition, but it is small and uneconomical to use. DOE plans to remove the fuel from this facility by
1996.
DOE has used Test Area North for commercial reactor fuel research. The large Test Area North
Hot Shop and Hot Cells have supported the Loss of Fluid Test and commercial nuclear fuel testing,
including dry cask storage demonstration. Test Area North stores special case commercial fuel
(including Three Mile Island Unit 2 core debris) and DOE experimental fuel similar to commercial
nuclear fuel.
Test Reactor Area has historically operated a number of test reactors, but the Advanced Test
Reactor and its associated Critical Facility are the only reactors now operating. Most spent nuclear
fuel at this area is associated with the Test Reactor Area reactors, which utilized aluminum-based
fuels. In addition, DOE stores small amounts of special case commercial, foreign, and Power Burst
Facility spent nuclear fuel at Test Reactor Area in the Materials Test Reactor basin. All spent nuclear
fuel in storage at the Test Reactor Area is in water-filled pools (DOE 1993).
2.1.3 Spent Nuclear Fuel Mission
The INEL spent nuclear fuel mission is to manage DOE-owned spent fuel cost-effectively and in
a way that protects the safety of INEL workers, the public, and the environment. As the lead
laboratory for the DOE Spent Nuclear Fuel Program, the INEL provides support to the Office of Spent
Fuel Management and coordinates the development of an integrated program for DOE.
The main focus of near-term activities is the accurate quantification and characterization of
DOE-owned spent nuclear fuel, identification of spent nuclear fuel management facilities and their
conditions, identification of safe interim storage for existing and new spent nuclear fuel, and
identification of technologies and requirements to place DOE spent nuclear fuel in safe interim storage.
Long-term activities include the development of final waste acceptance criteria requirements and
stabilization technologies for alternate fuel disposition, construction of facilities to stabilize fuel to
meet waste disposal requirements, processing of the fuel to a final waste form, and transportation of
the waste form for disposition.
2.2 Regulatory Framework for Spent Nuclear Fuel Management
This section summarizes State of Idaho laws and regulations that apply to spent nuclear fuel
management at the INEL. Volume 1, Section 7.2, provides summary information for Federal laws and
regulations, Executive Orders, and DOE Orders. Volume 2, Chapter 2, provides information on
National Environmental Policy Act reviews related to site-specific decisions that have potential
environmental impacts. Volume 2, Chapter 7, provides information on regulatory permits that the
INEL holds or for which it has applied.
The Idaho Environmental Protection and Health Act (Idaho Code, Title 39, Chapter 101 et seq.)
establishes general provisions for the protection of the environment and public health. The Act created
the Idaho Department of Health and Welfare and its Division of Environmental Quality, thereby
consolidating all state public health and environmental protection activities in one department. The
Act authorizes the Department to promulgate standards, rules, and regulations related to water and air
quality, noise reduction, and solid waste disposal; and grants authority to issue required permits,
collect fees, establish compliance schedules, and review plans for the construction of sewage and
public water treatment and disposal facilities.
The Idaho Water Pollution Control Act (Idaho Code, Title 39, Chapter 36) authorizes the
Department of Health and Welfare to protect the waters of Idaho. This law contains general language
on the prevention of water pollution and the provision of financial assistance to municipalities.
The Idaho Department of Health and Welfare is also responsible for the enforcement and
implementation of the Hazardous Waste Management Act of 1983, as amended (Idaho Code, Title 39,
Chapter 44), which provides for the protection of health and the environment from the effects of
improper or unsafe management of hazardous wastes and for the establishment of a tracking or
manifesting system for these wastes. This program is intended to be consistent with, and not more
stringent than, the Federal regulations established under the Resource Conservation and Recovery Act
(RCRA). At this time, Idaho has primacy over hazardous and mixed waste regulations promulgated
through July 1, 1990, by the U.S. Environmental Protection Agency. The Hazardous Waste
Management Act sets forth requirements for the development of plans that address the identification of
hazardous wastes; unauthorized treatment, storage, release, use, or disposal of these wastes; and permit
requirements for hazardous waste facilities. Under the authority of this Act, the Idaho Department of
Health and Welfare has promulgated rules and regulations on the transportation, monitoring, reporting,
and record keeping of hazardous wastes.
Several INEL facilities have air quality permits from the State, and operate in compliance with
permit conditions. Permit applications are currently pending with the State for proposed new or
modified emission sources. In April 1991 DOE submitted an inventory of all potential INEL
radioactive and criteria pollutant emission sources to the State. The inventory contains the information
necessary for the State to issue the INEL a Permit to Operate.
The Idaho Department of Health and Welfare, Division of Environmental Quality, Air Quality
Bureau, conducts annual inspections of the INEL to determine if the operating portions of the site are
in compliance with the Rules for the Control of Air Pollution in Idaho. The most recent inspections
were in January 1994. In addition, pursuant to 40 CFR Part 61.94(H), DOE submits to the State an
annual report documenting compliance with National Emission Standards for Hazardous Air Pollutants
at the INEL.
2.3 Spent Nuclear Fuel Management Program at the INEL
In 1992 the Secretary of Energy directed the Assistant Secretary for Environmental Restoration
and Waste Management to develop an integrated, long-term spent nuclear fuel management program.
In response to this request, DOE created the Office of Spent Fuel Management (EM-37). This office,
which has strategic programmatic responsibilities, has designated the INEL as the program support
organization for the DOE Spent Nuclear Fuel Program. In this role, the INEL provides technical
support to the Office of Spent Fuel Management and develops site communication and integration for
the national program.
As identified in the Spent Fuel Working Group Report on Storage of the Department's Spent
Nuclear Fuel and Other Reactor Irradiated Nuclear Materials and Their Environmental, Safety and
Health Vulnerabilities, Volume I (DOE 1993), some of the current storage facilities at the INEL are
inadequate for extended interim storage, and additional storage facilities or modifications might be
necessary. In February 1994, DOE issued, Plan of Action to Resolve Spent Nuclear Fuel
Vulnerabilities, Phase I (DOE 1994a), followed by a Phase II Plan in April 1994 (DOE 1994b) and a
Phase III Plan in October 1994 (DOE 1994c), which identified specific corrective actions to address
the spent nuclear fuel vulnerabilities. At the INEL, many of the corrective actions have been
completed or are currently underway. The spent nuclear fuel storage pools at Test Area North, Power
Burst Facility, and the Underwater Fuel Storage Facility do not comply with new facility regulatory
requirements. The INEL plans to move spent nuclear fuel from the CPP-603 Underwater Fuel Storage
Facility by December 31, 2000. To stabilize this fuel for storage, the INEL also plans to install
canning equipment in the Irradiated Fuel Storage Facility hot cell. This equipment is scheduled for
operation by late 1995. To the extent of its existing capability, DOE could consolidate spent nuclear
fuel at the Power Burst Facility, the Idaho Chemical Processing Plant, and the Test Area North at the
Idaho Chemical Processing Plant as a result of implementing the management alternatives described in
Chapter 3. These activities and other planned actions for which National Environmental Policy Act
review will be completed before the Record of Decision of this EIS were analyzed under the No-
Action Alternative (see Chapter 3).
Each of the specific INEL spent nuclear fuel Plan of Action projects could result in emissions,
worker exposures, and other potential environmental impacts. The potential environmental impacts
that could result from each project or corrective action item were not analyzed individually but were
collectively enveloped by the spent nuclear fuel management activities reported and analyzed for each
alternative. Successful completion of the corrective actions would significantly reduce the near-term
environmental, safety, and health risks associated with spent fuel storage at INEL.
The INEL has provided support in the development of dry at-reactor storage of special case
commercial spent nuclear fuel in accordance with the requirements of the Nuclear Waste Policy Act of
1982 and its 1987 amendments. Dry-storage demonstrations and research at the INEL contributed to
the granting of NRC licenses to several utilities for the construction and operation of dry-storage
facilities at reactor sites. Research at these facilities is demonstrating the technical feasibility and the
economics of adding dry storage capacity in metal or concrete spent fuel storage casks at reactor sites.
3. SPENT NUCLEAR FUEL MANAGEMENT ALTERNATIVES
Chapter 3 describes the alternatives for spent nuclear fuel management as they relate to the Idaho
National Engineering Laboratory (INEL) and summarizes and compares potential environmental
consequences for each alternative. Chapter 5 contains full descriptions of the consequences of
implementing the alternatives.
3.1 Description of Alternatives
DOE has identified five spent nuclear fuel management alternatives:
Alternative 1 - No Action
Alternative 2 - Decentralization (2a, 2b, and 2c)
Alternative 3 - 1992/1993 Planning Basis
Alternative 4 - Regionalization (4a and 4b)
Alternative 5 - Centralization (5a and 5b)
Table 3-1 summarizes the actions that would result from the implementation of these alternatives
at the INEL. For each alternative, this table summarizes the proposed transportation, stabilization,
storage, research and development, and naval-type fuel examination activities. For alternatives 2, 4,
and 5, it identifies a number of options.
The analysis of each alternative considers, as appropriate, existing and projected spent nuclear
fuel inventories, existing spent nuclear fuel wet and dry storage facilities, the construction of storage
facilities and associated stabilization facilities to achieve interim management objectives, and the
relocation of the spent nuclear fuel as appropriate to proposed interim storage facilities.
Table 2-2 lists existing spent nuclear fuel storage facilities with associated type(s) of storage and
fuel. Table 3-2 lists the potential facilities and projects required for specific alternatives. DOE has
based the potential environmental consequences for each alternative on the existing and proposed
facilities and projects listed in Tables 2-2 and 3-2, respectively.
Table 3-1. Summary of spent nuclear fuel management alternatives at the Idaho National Engineering Laboratory. (Page 1)
Table 3-1. (Page 2) Table 3-1. (Page 3) Table 3-2. Potential spent nuclear fuel projects required for each alternative(a).
The alternatives involving the interim storage of naval spent nuclear fuel at sites other than the
INEL include a transition period, which would start on June 1, 1995, and continue for approximately
3 years. During this period, approximately 80 shipments of naval spent nuclear fuel would occur to the
Expended Core Facility for examination and subsequent shipment to the Idaho Chemical Processing
Plant for storage. After this transition period, DOE would phase out the Expended Core Facility such
that the worker total at the facility would decline to about 10 by 2001. Appendix D describes this
transition period.
3.1.1 Alternative 1: No Action
Table 3-1 lists the basic actions expected under this alternative. This alternative would be
restricted to the minimum actions necessary for the continued safe and secure management of spent
nuclear fuel. Table 3-3 lists the existing inventory of spent nuclear fuel at the INEL. This alternative
is not a status quo condition in terms of spent nuclear fuel receipts (unlike Alternative 3, under which
operations would continue in accordance with the 1992/1993 planning basis). Rather, DOE would
maintain spent nuclear fuel close to defueling or current storage locations with minimal facility
upgrades or replacements.
DOE would continue the operation of the following existing spent nuclear fuel-related facilities:
the Fuel Storage Area/Fluorinel Dissolution Process Cell; CPP-603 Underwater Fuel Storage Facility
(until 2000); Irradiated Fuel Storage Facility; Underground Storage Facility; Power Burst Facility
storage canal; Advanced Test Reactor canal; Advanced Reactivity Measurement Facility; Coupled Fast
Reactivity Measurement Facility; Materials Test Reactor canal; Test Area North Pool and Test Pad;
Argonne National Laboratory - West Hot Fuel Examination Facility, Radioactive Scrap and Waste
Facility, Transient Reactor Test Facility, Zero Power Physics Reactor, and Neutron Radiography
Reactor pool. Table 2-2 lists the type(s) of storage and spent nuclear fuels associated with each facility.
3.1.1.1 Transportation. Under this alternative, the INEL would neither receive nor ship spent
nuclear fuel except for naval spent fuel during a transition period. DOE would continue to transfer the
Advanced Test Reactor canal spent nuclear fuel to the Idaho Chemical Processing Plant. In addition,
DOE could transfer other spent nuclear fuel at the INEL site (e.g., Test Reactor Area, Test Area North
Pad, Power Burst Facility storage canal, Experimental Breeder Reactor-II, and Naval Nuclear
Table 3-3. Spent nuclear fuel inventory for each alternative by 2035 (metric tons of heavy metal). ,b,c
____________________________________________________________________________________________________________________________________
Fuel Type 1. 2. 3. 4a. 4b(1)e 5a. 5b.
No Decentralization 1992/1993 Regionalization Regionalization Centralization Centralization
Action(d) Planning by Fuel Type by Geography at Other at the INEL
Basis (INEL) DOE Sites
____________________________________________________________________________________________________________________________________
Naval-type 10.23 N/Cf +55.00 +55.00 +55.00 -10.23 +55.00
Aluminum-clad 2.91 11.02 +12.09 -2.91 +5.85 -2.91 +210.18
Hanford None None None None +2,103.17 None +2,103.17
Graphite 11.60 N/C +16.00 +16.01 +16.01 -11.60 +16.01
Special case 122.88 +0.03 +26.69 +33.63 +2.30 -122.88 33.63
commercial
Stainless-steel- 77.43 +1.08 +1.19 +19.08 +12.69 -77.43 +19.08
clad
Zircaloy-clad 49.09 +0.67 +0.670 +28.90 +15.75 -49.09 +28.90
Other 0.01 +0.82 +0.82 +1.69 +0.28 -0.01 +1.69
Net increase (+)/ - +13.62 +112.47 +151.41 +2,211.05 -274.14 +2,467.66
decrease (-)
TOTAL 274.14 287.76 386.61 425.55 2,485.19 0 2,741.80
-----------------------------
a. Source: Wichmann (1995).
b. To convert metric tons to tons, multiply by 1.10. Heavy metals are uranium, plutonium, and thorium.
c. The values may not sum exactly due to rounding.
d. The No-Action Alternative represents the present inventory and projections and serves as the basis for
determining the net increase or decrease for each type of spent nuclear fuel for each of the other alternatives.
e. Regionalization 4b(2), Regionalization by Geography (Elsewhere), assumes all spent nuclear fuel inventories at
the INEL go to the Nevada Test Site or Hanford Site. Inventories for 4b(2) would equal those listed for
Alternative 5a.
f. N/C = No change from the No-Action Alternative.
_____________________________________________________________________________________________________________________________________
Propulsion Program prototype reactors at the Naval Reactors Facility) to the Idaho Chemical
Processing Plant to the extent of its storage capability.
3.1.1.2 Stabilization. Due to the deteriorated condition of some of the fuel in the CPP-603
Underwater Fuel Storage Facility, additional canning and characterization capabilities would be
necessary to stabilize this fuel for safe transport and subsequent storage. DOE has scheduled the
installation and operation of new fuel canning and characterization equipment in the Irradiated Fuel
Storage Facility, which could provide these capabilities, by late 1995. (The installation of such
equipment would be a minor upgrade and would have a smaller extent than similar actions described
under Alternatives 3, 4, and 5.) DOE could perform other required stabilization of spent nuclear fuel
at the INEL in either the Remote Analytical Laboratory or the Fluorinel Dissolution Process Hot Cell.
3.1.1.3 Storage. DOE has identified the CPP-603 Underwater Fuel Storage Facility as one of
five complex-wide spent nuclear fuel storage facilities that exhibit the greatest vulnerabilities according
to selected criteria and, therefore, has selected this facility for priority attention (DOE 1993b). As part
of the August 9, 1993, agreement between the Secretaries of the Department of Energy and the
Department of the Navy and the Governor of Idaho to phase out storage operations in the 45-year old
CPP-603 facility, one goal of this and the other alternatives would be to remove spent nuclear fuel from
underwater storage in the North and Middle Basins of the CPP-603 facility by the end of 1996 and
from the South Basin of this facility by the end of 2000 (DOE 1993a). DOE would relocate this
material to the Fuel Storage Area at the Idaho Chemical Processing Plant.
At the Argonne National Laboratory-West, the spent nuclear fuel stored at the Hot Fuel
Examination Facility and the Radioactive Scrap and Waste Facility, primarily Experimental Breeder
Reactor-II fuel and blanket elements, would remain in dry storage until its potential processing in the
Fuel Cycle Facility. At the Experimental Breeder Reactor-II site, DOE would use dry storage with the
exception of the Neutron Radiography Reactor pool fuel. The Test Area North Pool Fuel Transfer
project would continue, resulting in the relocation of Test Area North spent pool contents into dry cask
storage at the Idaho Chemical Processing Plant by 1998. The dry cask storage required for this project
is not related to the Dry Fuels Storage Facility.
DOE would start no new projects to increase spent nuclear fuel storage capacity because there is
sufficient storage capacity to meet No-Action storage needs. The planning of spent nuclear fuel storage
projects such as the Dry Fuels Storage Facility and Additional Increased Rack Capacity for the Fuel
Storage Area would stop.
3.1.1.4 Research and Development. There would be only limited spent nuclear fuel
research and development. Existing spent nuclear fuel management research and development projects
would continue. Existing facilities such as the Process Improvement Facility, the Remote Analytical
Laboratory, and the Pilot Plant Facility would support continuing research and development work.
3.1.1.5 Naval-Type Fuel Examination. After a transition period, DOE would cease
shipments of naval spent nuclear fuel to the INEL and would phase out the Expended Core Facility.
DOE would make onsite shipments of the "library fuel" (a representative sampling of different fuel
types maintained for reference purposes) and the spent nuclear fuel that originated at the prototype sites
at the Naval Reactors Facility to the Idaho Chemical Processing Plant.
3.1.2 Alternative 2: Decentralization
Under this alternative, DOE could transport fuel for safety or research and development
activities. In addition, DOE could undertake actions for safety it deemed desirable, though not
essential, and could perform spent nuclear fuel treatment and research and development. As listed in
Table 3-3, the anticipated spent nuclear fuel inventory for this alternative would be slightly greater than
the inventory for Alternative 1, with the increase consisting primarily of aluminum-clad and stainless-
steel-clad spent nuclear fuel from university and foreign research and experimental reactors.
3.1.2.1 Transportation. This alternative assumes that the INEL would accept primarily
limited shipments of spent nuclear fuel from offsite sources into the Fuel Storage Area (e.g., DOE or
university reactors) after the Record of Decision for this EIS (1995). Onsite transfers could occur from
the Fuel Storage Area to the Storage Facility or the Irradiated Fuel Storage Facility. DOE would
consolidate the spent nuclear fuel in the Advanced Test Reactor and in the Materials Test Reactor and
Power Burst Facility canals at the Idaho Chemical Processing Plant for canning, characterization, and
storage.
As in the No-Action Alternative, there would be a transition period during which the Naval
Nuclear Propulsion Program would ship naval spent nuclear fuels to the Expended Core Facility for
examination and subsequent shipment to the Idaho Chemical Processing Plant for storage.
Section 3.1.2.5 describes the transportation of naval spent fuels that would occur after the transition
period.
3.1.2.2 Stabilization. DOE would use the canning and characterization equipment identified in
Section 3.1.1.2 to stabilize spent nuclear fuel removed from the CPP-603 Underwater Fuel Storage
Facility for interim underwater storage.
3.1.2.3 Storage. As in Alternative 1, DOE would transfer the spent nuclear fuel in the
CPP-603 Underwater Fuel Storage Facility to the Fuel Storage Area by 2000. DOE would continue to
use the Underground Storage Facility and the Irradiated Fuel Storage Facility for existing spent nuclear
fuel inventory and transfers of other spent nuclear fuel based on safety analyses. DOE would upgrade
or increase fuel storage capacity at the INEL as required.
The Test Area North Pool Fuel Transfer project would result in the relocation of the contents of
Test Area North spent nuclear fuel into dry storage at a pad at the Idaho Chemical Processing Plant.
3.1.2.4 Research and Development. The development of technology for the disposition of
spent nuclear fuel would continue. Research and development activities would include laboratory and
pilot plant testing, continued repository performance assessments and waste acceptance criteria
development, and the characterization of spent nuclear fuel. Shipments of samples or selected spent
nuclear fuel assemblies to offsite DOE facilities would be necessary.
3.1.2.5 Naval-Type Fuel Examination. DOE would consider three options for naval reactor
spent nuclear fuel receipt and shipment. Under options 2a and 2b, DOE would stop shipments of naval
spent nuclear fuel to the INEL and would shut down the Expended Core Facility. Option 2c would
enable the continued receipt of naval-type fuel for examination at the Expended Core Facility and its
return to the originating shipyards for storage in transport casks. Chapter 3 of Appendix D further
describes these options. As with Alternative 1, each option would require approximately a 3-year
transition period. During this period, DOE would transport spent nuclear fuel in shipping containers to
the Expended Core Facility, unload the containers, and use them to support additional refuelings and
defueling.
3.1.3 Alternative 3: 1992/1993 Planning Basis
This alternative is consistent with DOE plans at the INEL before the injunction that stopped spent
nuclear fuel shipment to the INEL; it assumes a 40-year planning horizon for the continued
transportation, receipt, stabilization, and storage of spent nuclear fuel. As with Alternative 1, DOE
would continue the maintenance and operation of existing spent nuclear fuel-related facilities; however,
some consolidation of INEL facilities could occur. DOE would send newly generated spent nuclear
fuel to either the INEL or the Savannah River Site. DOE would assess the construction of new
facilities to accommodate current and projected spent nuclear fuel management requirements.
The amount of spent nuclear fuel at the INEL under this alternative would be greater than that for
either Alternative 1 or 2 (see Table 3-3) because this alternative assumes that the INEL would
manage, before stabilization and disposal, its present inventory (see Alternative 1) plus additional
receipts of DOE spent nuclear fuel, including the following:
- Naval-type spent nuclear fuel
- Approximately half of the aluminum-clad spent nuclear fuel from university and foreign
research and experimental reactors
- All Training Reactor Isotopics General Atomics (TRIGA) spent nuclear fuels from the
Hanford Site and approximately half of that from foreign, DOE, and university reactors
- Fort St. Vrain spent nuclear fuel from Public Service of Colorado
- Special case commercial pressurized water reactor and boiling water reactor spent nuclear
fuel from the DOE facility in West Valley, New York
- Miscellaneous spent nuclear fuel types from such DOE sites as Los Alamos, New Mexico,
and Oak Ridge, Tennessee, and from university reactors and other locations
3.1.3.1 Transportation. DOE would consolidate the spent nuclear fuel in the Test Reactor
Area (Advanced Test Reactor canal, Materials Test Reactor canal, and Coupled Fast Reactivity
Measurements Facility and Advanced Reactivity Measurement Facility canal) and the Power Burst
Facility at the Idaho Chemical Processing Plant for canning and dry storage.
The INEL would receive and temporarily store new spent nuclear fuels in the Fuel Storage Area.
Transfers could occur from the Fuel Storage Area to the Underground Storage Facility or the Irradiated
Fuel Storage Facility or, when available, the dry storage vaults at the proposed Dry Fuels Storage
Facility.
At present, DOE is transferring spent nuclear fuel from the Advanced Test Reactor Canal to the
Idaho Chemical Processing Plant. DOE would maintain this canal for the storage and management of
its recyclable fuel assemblies until the reactor no longer had a mission. The Experimental Breeder
Reactor-II spent nuclear fuel in storage would remain at Argonne National Laboratory-West. As with
Alternative 2, the Test Area North Pool Fuel Transfer project would result in the relocation of the
contents of the Test Area North spent nuclear fuel pool to dry storage at a pad at the Idaho Chemical
Processing Plant.
3.1.3.2 Stabilization. DOE would complete a new Canning and Characterization Facility with
appropriate inspection, stabilization, and packaging equipment to stabilize new receipts of spent nuclear
fuel and to prepare fuel currently in underwater storage for dry storage. This facility would be an
integral part of the Dry Fuels Storage Facility that DOE would complete under this alternative. Until
the Dry Fuels Storage Facility is in service, DOE would use the canning and characterization
equipment described under Alternative 1 to stabilize spent nuclear fuel removed from the CPP-603
Underwater Fuel Storage Facility for interim underwater storage.
3.1.3.3 Storage. As with Alternative 2, DOE would upgrade or increase dry fuel storage
capacity at the INEL as required. DOE would complete the Fuel Storage Area increased Rack
Capacity project in 1997. Coupled with stringent fuel management and, if necessary, temporary
storage of some aluminum fuel in stainless steel racks, this project would allow the Fuel Storage Area
to accept all of the project spent nuclear fuel receipts until the Additional Increased Rack Capacity
project would be completed in 2001. The Additional Increased Rack Capacity project would allow the
Fuel Storage Area to accept the projected spent nuclear fuel receipts until the Dry Fuels Storage
Facility project would become available in 2005. The INEL would receive the Fort St. Vrain spent
nuclear fuel in the Irradiated Fuel Storage Facility on a space-available basis or in the new vault storage
in the Dry Fuels Storage Facility. Modifications to the Irradiated Fuel Storage Facility cask handling
equipment would be necessary to accept the new Fort St. Vrain shipping casks.
DOE would continue to use the Underground Storage Facility and the Irradiated Fuel Storage
Facility for current inventory and for transfers of other fuel inventories based on safety analyses.
Based on these safety analyses, upgrades would be limited to those required for facility safety
improvements and for making transfers safely.
3.1.3.4 Research and Development. Spent nuclear fuel research and development would
continue as planned, with the construction of a Technology Development Facility. The
Electrometallurgical Process Demonstration Project at Argonne National Laboratory - West Fuel Cycle
Facility would continue. In addition, Argonne National Laboratory would implement the EBR-II
Blanket Processing project under this alternative. The Dry Fuels Storage Facility would develop and
demonstrate technology for the dry storage of selected DOE highly enriched uranium fuels.
3.1.3.5 Naval-Type Fuel Examination. The practice of transporting spent nuclear fuel from
naval reactors to the Expended Core Facility at the INEL would resume. After an examination, DOE
would transfer such fuel to the Idaho Chemical Processing Plant for interim storage pending final
disposition. Under this alternative, the Naval Nuclear Propulsion Program would complete the
Expended Core Facility Dry Cell Construction project.
3.1.4 Alternative 4: Regionalization
This alternative assumes that DOE would base the spent nuclear fuels shipped between DOE sites
and the receipt of fuels from other locations primarily on either geography or fuel type. Alternative 4
offers two options for the redistribution of existing and new spent nuclear fuel:
- Option 4a assumes that DOE would base the spent nuclear fuels shipped between DOE sites
and the receipt of fuels from other locations at the INEL, Hanford Site, or the Savannah
River Site primarily on fuel type.
- Option 4b assumes that DOE would base the spent nuclear fuels shipped between DOE sites
and the receipt of fuels on geography. There would be a single western site at either the
Hanford Site, INEL or Nevada Test Site. Option 4b(1) in which the INEL is the western
regional site is essentially the same as Alternative 5b. Option 4b(2) in which INEL ships all
SNF to another western regional site is the same as Alternative 5a.
3.1.4.1 Transportation. Under option 4a, the INEL would receive all Zircaloy- and
stainless-steel-clad spent nuclear fuel. This redistribution would optimize DOE spent nuclear fuel
management.
The spent nuclear fuel inventory involved under option 4a would be greater than those for
Alternative 1, 2, or 3 because this alternative assumes that the INEL would manage its present
inventory plus the following additional spent nuclear fuels (see Table 3-3) prior to stabilization and
disposal:
- Naval-type spent nuclear fuel
- All spent nuclear fuel except aluminum-clad fuel and Hanford spent nuclear fuel
- All Training Reactor Isotopics General Atomics spent nuclear fuels from the Hanford Site
- Fort St. Vrain spent nuclear fuel from Public Service of Colorado
- Special case commercial pressurized water reactor and boiling water reactor spent nuclear
fuel from the DOE facility in West Valley, New York
Under option 4b(1), DOE would regionalize all western DOE SNF at the INEL. DOE would
transport all spent nuclear fuel at other western sites to the INEL. Because the fuel inventory for this
alternative would be within 15 percent of that for Alternative 5b, analyses for this option conservatively
assume that environmental impacts would be the same as those for as Alternative 5b - Centralization at
INEL.
Under option 4b(2), DOE would regionalize all western DOE SNF at either the Nevada Test Site
or Hanford Site. DOE would transport spent nuclear fuel at the INEL to the selected western site. As
such, this option would be the same as Alternative 5a - Centralization at Other DOE Sites.
3.1.4.2 Stabilization. DOE would stabilize the spent nuclear fuels it would retain at the INEL
as planned for Alternative 3, with the construction of such new facilities as a canning and
characterization facility and the Dry Fuels Storage Facility. Options 4a and 4b(1) would require such a
facility for the receipt and storage of spent nuclear fuel, while option 4b(2) would require stabilization
capabilities for shipping spent nuclear fuel. For spent nuclear fuel that the INEL would ship to other
regional sites, the receiving site would perform any stabilization beyond that required for
transportation.
3.1.4.3 Storage. Under option 4a, DOE would increase dry storage capacity and undertake
facility upgrades similar to those described for Alternative 3, with replacements and additions as
appropriate. Under option 4b(1), DOE would increase dry storage capacity and undertake facility
upgrades similar to those described for Alternative 5b, with replacements and additions as appropriate.
Option 4b(2) would not require increased storage capacity and, therefore, there would be no facility
upgrades.
3.1.4.4 Research and Development. As with Alternative 3, this alternative would include
the continuation of activities related to the treatment of spent nuclear fuel, including research and
development (e.g., Electrometallurgical Process Demonstration Project), and the construction of the
Dry Fuels Storage Facility. DOE would initiate pilot programs as needed to support future decisions
on spent nuclear fuel management and disposition. DOE would use historic data on spent nuclear fuel
to provide the bounding case for a determination of the impacts associated with potential pilot program
activities.
3.1.4.5 Naval-Type Fuel Examination. Under options 4a and 4b(1), the transportation of
spent nuclear fuel from naval reactors to the Expended Core Facility at the INEL would resume. As
with Alternative 1, under option 4b(2) DOE would phase out shipments of naval-type spent nuclear fuel
to the INEL and would phase out the Expended Core Facility.
3.1.5 Alternative 5: Centralization
Under this alternative, DOE would send all current and future spent nuclear fuel inventories from
both DOE and the Naval Nuclear Propulsion Program to one DOE site for interim storage until final
disposition.
The two options under Alternative 5 encompass the extreme ranges of spent nuclear fuel
inventories that DOE could store at the INEL (i.e., all or none of the inventory). Under option 5a,
DOE would ship the INEL spent nuclear fuel inventory off the site to the Hanford Site, the Savannah
River Site, the Nevada Test Site, or the Oak Ridge Reservation. Under option 5b, DOE would ship all
existing spent nuclear fuel to the INEL.
This alternative would bound the maximum number of spent nuclear fuel-related actions that DOE
could reasonably undertake at any site. DOE would have to build new facilities at the selected site to
accommodate the increased inventories. Shipments of spent nuclear fuel to the sites not selected as the
centralized destination would continue as an interim action pending the construction of necessary
storage and examination facilities at the selected site. DOE would then transfer all spent nuclear fuel to
the selected site, and the other sites would close their spent nuclear fuel facilities. Before DOE would
ship spent nuclear fuel from the originating site, it would characterize and can all spent nuclear fuel as
necessary.
The locations from which spent nuclear fuel would originate, in addition to the Hanford Site and
Savannah River Site, would include Argonne National Laboratory - East, Babcock and Wilcox,
Brookhaven National Laboratory, General Atomics, Los Alamos National Laboratory, Oak Ridge
National Laboratory, Sandia National Laboratories, West Valley, and Fort St. Vrain. This alternative
would also include fuel that might be returned to the United States following irradiation or testing.
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 disposition. DOE
would use historic data on spent nuclear fuel to provide a foundation case for determining the impacts
associated with potential pilot program activities.
3.1.5.1 Alternative 5a - Centralization at Other DOE Sites.
3.1.5.1.1 Transportation - This option assumes that the INEL would consolidate and
prepare all existing and projected onsite spent nuclear fuel for shipment to another DOE facility: the
Hanford Site, the Savannah River Site, the Nevada Test Site, or Oak Ridge.
3.1.5.1.2 Stabilization - The DOE would construct a canning and characterization facility
at the Idaho Chemical Processing Plant to accept the different types of INEL spent nuclear fuel in
various shipping casks and storage containers, and to stabilize these fuel types before their shipment to
the selected DOE facility.
3.1.5.1.3 Storage - As in Alternative 1, DOE would complete the CPP-603 Underwater
Fuel Storage Facility pool inventory transfer to existing dry storage facilities by 2000.
DOE would not
build the Dry Fuels Storage Facility. DOE would then close all spent nuclear fuel-related facilities at
the INEL with the exception of those in direct support of operating reactors, such as the Advanced Test
Reactor canal or the Argonne National Laboratory-West Hot Fuel Examination Facility and Fuel Cycle
Facility. This closure would require the establishment of a major surveillance and maintenance
operation until DOE determined the disposition of these facilities. The timeframe for closure would
depend on the following factors:
- The time necessary to stabilize the spent nuclear fuel in the CPP-603 Underwater Fuel
Storage Facility
- The time necessary for the selected DOE site to prepare facilities qualified to accept the spent
nuclear fuel
- The time necessary for the procurement and licensing of shipping containers that would be
compatible with the selected receiving DOE site
The spent nuclear fuel inventory that DOE would export off the INEL site for Alternative 5a is
the same quantity listed for Alternative 1 (see Table 3-3).
3.1.5.1.4 Research and Development - Under this option there would be a phaseout of
all research and development activities, although the Electrometallurgical Process Demonstration
Project would continue at the Argonne National Laboratory - West Fuel Cycle Facility (but would
stabilize only spent nuclear fuel currently on the site).
3.1.5.1.5 Naval-Type Fuel Examination - As with Alternative 1, DOE would phase out
shipments of naval-type spent nuclear fuel to the INEL and would phase out the Expended Core
Facility.
3.1.5.2 Alternative 5b - Centralization at the INEL.
3.1.5.2.1 Transportation - This option assumes that the INEL would receive all DOE and
naval-type spent nuclear fuel (see Table 3-3).
3.1.5.2.2 Stabilization - The Hanford Site, the Savannah River Site, and other DOE
facilities would stabilize as necessary, spent nuclear fuel for safe transportation to the Idaho Chemical
Processing Plant.
The Hanford Site, the Savannah River Site, and other DOE facilities would procure
an undetermined number of additional casks and install cask handling equipment as necessary. DOE
would complete an expanded Dry Fuels Storage Facility at the INEL, which would include a new
Canning and Characterization Facility similar to that described for Alternative 3. This facility would,
if needed, repackage the spent nuclear fuel into compatible canisters for dry storage. Other new
facility projects would be the same as those described for Alternative 3. In addition, DOE would begin
stabilizing for safe storage all complex-wide spent nuclear fuel, as necessary, in existing facilities at the
Idaho Chemical Processing Plant. Upgrades and new facilities would be necessary to support long-
term fuel stabilization for ultimate disposition; this would address criticality (unplanned and
uncontrolled nuclear fission) concerns about the disposal of spent nuclear fuel in a potential Federal
repository.
3.1.5.2.3 Storage - Projects and activities for storage of spent nuclear fuel would be similar
to those described for Alternative 3, except that accelerated schedules for the Increased Rack Capacity
and Additional Increased Rack Capacity projects would be necessary to accommodate the increased
fuel receipts.
In addition, the schedule for the Dry Fuel Storage Facility project would have to be
accelerated and its scope expanded. For example, the Increased Rack Capacity project may have to be
completed in late 1996, the Additional Increased Rack Capacity project may have to be completed in
late 1998, and the Expanded Dry Fuels Storage Facility project may have to be completed in 2002. If
the Expanded Dry Fuels Storage Facility would become available even earlier, it could eliminate the
need for the Additional Increased Rack Capacity project.
3.1.5.2.4 Research and Development - DOE would conduct maximum spent nuclear
fuel research and development under this option.
As with Alternative 4, the Electrometallurgical
Process Demonstration Project would continue at the Argonne National Laboratory - West.
3.1.5.2.5 Naval-Type Fuel Examination - Similar to Alternative 3, the practice of
transporting spent nuclear fuel from naval reactors to the Expended Core Facility at the INEL would
resume.
3.2 Comparison of Alternatives
Chapter 5 analyzes the environmental consequences of the alternatives. Tables 3-4 through 3-6
summarize and compare the potential impacts associated with each alternative from the information in
Chapter 5 for construction, normal operations, and accidents, respectively.
A review of the impacts of the alternatives, as presented in Chapter 5, indicates that impacts
would be minimal or negligible in most areas. Further, most areas with measurable impacts would
have no appreciable differences among alternatives.
In general, the levels of potential impacts associated with Alternatives 1 through 4 (option 4a)
would be similar because the amounts of spent nuclear fuel that DOE would manage at the INEL under
these alternatives would be on the same order of magnitude (e.g., 300 to 450 MTHM) and activities
would extend throughout the full 40-year management period. The lowest level of overall potential
impact at the INEL would occur under Alternative 4b(2) - Regionalization by Geography (Elsewhere)
and Alternative 5a - Centralization at Other DOE Sites because DOE would ship INEL spent nuclear
fuel off the site well before the management period ended in 2035. Alternative 5b and Alternative
4b(1), under which DOE would ship all or nearly all spent nuclear fuel to the INEL, would result in the
greatest potential onsite impacts.
4. AFFECTED ENVIRONMENT
Table 3-4. Comparison of impacts from construction. (Page 1)
Table 3-4. (Page 2)
Table 3-4. (Page 3)
Table 3-5. Comparison of impacts from normal operations. (Page 1)
Table 3-5. (Page 2)
Table 3-5. (Page 3)
Table 3-6. Comparison of impacts from accidents.
4.1 Overview
Chapter 4 describes the existing environment at the Idaho National Engineering Laboratory
(INEL) site and the surrounding region. It emphasizes areas that the proposed spent nuclear fuel
management alternatives could affect. The information in this chapter provides the existing
environmental conditions against which the Department of Energy (DOE) can measure the potential
environmental effects of the alternatives. It supports the assessment of the potential environmental
consequences that Chapter 5 discusses. DOE used the discussion of the Affected Environment in
Volume 2 of this EIS as input for this chapter.
4.2 Land Use
The INEL site encompasses 570,914 acres (2,310.4 square kilometers) in Butte, Bingham,
Jefferson, Bonneville, and Clark Counties, Idaho. This section describes existing land uses at the INEL
and in the surrounding region, and land use plans and policies applicable to the surrounding area.
4.2.1 Existing and Planned Land Uses at the INEL
Categories of land use at the INEL include facility operations, grazing, general open space, and
infrastructure such as roads. Facility operations include industrial and support operations associated
with energy research and waste management activities (DOE also conducts such activities at its Idaho
Falls facilities). In addition, DOE uses INEL land for recreation and environmental research associated
with the designation of the INEL as a National Environmental Research Park.
Much of the INEL is open space that DOE has not designated for specific uses. Some of this
open space serves as a buffer zone between INEL facilities and other land uses. Facilities and
operations use about 2 percent of the total INEL site area (11,400 acres or 46 square kilometers).
Public access to most facility areas is restricted. Approximately 6 percent of the INEL, or
32,985 acres (133.5 square kilometers), is devoted to public roads and utility rights-of-way that cross
the site. Recreational uses include public tours of general facility areas and the Experimental Breeder
Reactor-I (a National Historic Landmark), and controlled hunting, which is generally restricted to
0.5 mile (0.8 kilometer) inside the INEL boundary.
Cattle and sheep grazing occupies between 300,000 and 350,000 acres (1,200 and 1,400 square
kilometers). The U.S. Sheep Experiment Station uses a 900-acre (3.6-square-kilometer) portion of this
land, at the junction of Idaho State Highways 28 and 33, for a winter feed lot for approximately 6,500
sheep. Grazing is not allowed within 2 miles (3.2 kilometers) of any nuclear facility and, to avoid the
possibility of milk contamination by long-lived radionuclides, dairy cattle are not permitted on the site.
The Department of the Interior's Bureau of Land Management grants and administers rights-of-way and
grazing permits. Figure 4.2-1 shows selected land uses at the INEL and in the surrounding region.
Figure 4.2-1 Selected land uses at the INEL and in the surrounding region. The INEL site is within the Medicine Lodge Resource Area (approximately 140,415 acres or
568.3 square kilometers in the eastern and southern portions of the INEL site) and the Big Butte
Resource Area (430,499 acres or 1,742 square kilometers in the central and western portions); the
Bureau of Land Management administers both of these areas. Under Resource Management Plans, the
Bureau manages portions of these Resource Areas for grazing and wildlife habitat. No mineral
exploration or development is allowed on INEL land.
DOE land use plans and policies applicable to the INEL include the INEL Institutional Plan -
Fiscal Year 1994 - 1999 (DOE-ID 1993c) and the INEL Technical Site Information Report (DOE-ID
1993a). The Institutional Plan provides a general overview of INEL facilities, outlines strategic
program directions and major construction projects, and identifies specific technical programs and
capital equipment needs. The Technical Site Information Report presents a 20-year master plan for
development activities at the site. Under the scope of these planning documents, energy research and
waste management activities would continue in existing facility areas and, in some instances, expand
into currently undeveloped site areas. These documents also describe environmental restoration, waste
management, and spent nuclear fuel activities. Projected land use scenarios for the next 25 to 50 years
include the outgrowth of current functional areas and the possible development of waterfowl production
ponds in existing grazing areas.
No onsite land use restrictions due to Native American treaty rights would exist for any of the
alternatives described in this EIS. The INEL does not lie within any of the land boundaries established
by the Fort Bridger Treaty, and the entire INEL site is land occupied by the U.S. Department of
Energy. Therefore, the provisions in the Fort Bridger Treaty that allows the Shoshone-Bannock
Indians to hunt on unoccupied lands of the United States do not apply to the INEL site.
4.2.2 Existing and Planned Land Use in Surrounding Areas
The Federal government, the State of Idaho, and private parties own the lands surrounding the INEL
site. Land uses on Federally owned land consist of grazing, wildlife management, range land, mineral
and energy production, and recreational uses. State-owned lands are used for grazing, wildlife
management, and recreational purposes. Privately owned lands are used primarily for grazing, crop
production, and range land.
Small communities and towns near the INEL boundaries include Mud Lake to the east; Arco,
Butte City, and Howe to the west; and Atomic City to the south. The larger communities of Idaho
Falls, Rexburg, Blackfoot, and Pocatello and Chubbock are to the east and southeast of the INEL site.
The Fort Hall Indian Reservation is to the southeast of the INEL. Recreation and tourist attractions in
the region around the INEL include the Craters of the Moon National Monument, Hell's Half Acre
Wilderness Study Area, Black Canyon Wilderness Study Area, Camas National Wildlife Refuge,
Market Lake State Wildlife Management Area, North Lake State Wildlife Management Area,
Yellowstone National Park, Grand Teton National Park, Jackson Hole Recreation Complex, Targhee
and Challis National Forests, and the Snake River.
Lands surrounding the INEL site are subject to Federal and state planning laws and regulations.
Federal rules and regulations that require public involvement in their implementation govern planning
for and use of Federal lands and their resources. Land use planning in the State of Idaho is derived
from the Local Planning Act of 1975 (State of Idaho Code 1975). Because the State currently has no
land use planning agency, the Idaho legislature requires each county to adopt its own land use planning
and zoning guidelines. County plans that are applicable to lands bordering the INEL site include the
Clark County Planning and Zoning Ordinance and Interim Land Use Plan (Clark County 1994);
Bonneville County Comprehensive Plan (Bonneville County 1976); Bingham County Zoning Ordinance
and Planning Handbook (Bingham County 1986); Jefferson County Comprehensive Plan (Jefferson
County 1988); and Butte County Comprehensive Plan (Butte County 1992). Land use planning for
INEL facilities within the Idaho Falls city limits is subject to Idaho Falls planning and zoning
restrictions (City of Idaho Falls 1989, 1992).
All county plans and policies accept development adjacent to previously developed areas to
minimize the need to extend infrastructure improvements and to avoid urban sprawl. Because the
INEL is remote from most developed areas, INEL lands and adjacent areas are not likely to experience
residential and commercial development; no new development is planned near the INEL site.
However, DOE expects recreational and agricultural uses to increase in the surrounding area in
response to greater demand for recreational areas and the conversion of range land to crop land.
4.3 Socioeconomics
This section presents a brief overview of current socioeconomic conditions within a region of
influence where approximately 97 percent of the INEL workforce lived in 1991 (DOE-ID 1991). The
INEL region of influence is a seven-county area comprised of Bingham, Bonneville, Butte, Clark,
Jefferson, Bannock, and Madison Counties. The region of influence also includes the Fort Hall Indian
Reservation and Trust Lands (home of the Shoshone-Bannock Tribes) in Bannock, Bingham, Caribou,
and Power Counties.
4.3.1 Employment
Historically, the regional economy has relied predominantly on natural resource use and
extraction. Today, farming, ranching, and mining remain important components of the regional
economy. Idaho Falls is the retail and service center for the region of influence, and Pocatello has
evolved into an important processing and distribution center and site of higher education institutions.
4.3.1.1 Region. The labor force in the region of influence increased from 92,159 in 1980 to
104,654 in 1991, an average annual growth rate of approximately 1.2 percent. In 1991 the region of
influence accounted for approximately 18 percent of the total state labor force of 504,000
(ISDE 1992). As listed in Table 4.3-1, the projected labor force in the region of influence will reach
108,667 by 1995.
Unemployment rates varied considerably among the counties of the region of influence in 1991,
ranging from 2.6 percent in Clark County to 6.3 percent in Bannock and Bingham Counties. Since
1980 the average annual unemployment rate for the region has ranged from 5.3 percent in 1989 to
8.3 percent in 1983. In 1991 the average annual unemployment rate for the region of influence was
5.5 percent compared to the statewide average of 6.2 percent (ISDE 1992).
Employment in the region of influence increased from 86,261 in 1980 to 98,898 in 1991, an
average annual growth rate of approximately 1.3 percent. As listed in Table 4.3-1, employment is
projected to increase to 101,450 by 1995.
Table 4.3-1. Projected labor force, employment, and population for the INEL region of influence,
1995-2004.
1995 1996 1997 1998 1999 2000 2001 2002 2003 2004
Labor Force 108,667 109,607 110,547 111,487 112,427 113,367 114,308 115,248 116,188 117,128
Employment 101,450 102,328 103,205 104,083 104,960 105,838 106,716 107,593 108,471 109,348
Population 247,990 251,518 255,096 258,726 262,406 266,140 268,667 271,219 273,795 276,395
Source: ISDE (1992); SAIC (1994); ISDE (1991); ISDE (1986).
4.3.1.2 Idaho National Engineering Laboratory. INEL plays a substantial role in the
regional economy. During Fiscal Year 1990, INEL directly employed approximately
11,100 personnel, accounting for almost 12 percent of total regional employment. The estimated
population directly supported by INEL employment was approximately 38,000 persons, or 17 percent
of the total regional population. The major employers at INEL are DOE-ID, DOE-ID contractors,
Argonne National Laboratory-West, and the Naval Reactors Facility (see Figure 4.3-1). In 1992, the
total direct INEL employment was approximately 11,600 jobs (DOE-ID 1994). Projections as of
January 1995 indicate that the total number of jobs at INEL will decrease to approximately 8,620 in
Fiscal Year 1995 and to approximately 7,250 in Fiscal Year 2004 (Tellez 1995). Projected decreases
in INEL employment are primarily related to contractor consolidation, which accounts for 64 percent
of the projected losses between Fiscal Year 1994 and Fiscal Year 2004, and to reduced activities at the
Naval Reactors Facility, which accounts for 33 percent of the projected job losses. Contract changes
at DOE-ID resulted in the consolidation of several contracts under one contract. The consolidation
eliminated redundant administrative activities previously performed by each individual contractor and
offered early retirement or other options to impacted INEL contractor employees.
4.3.2 Population and Housing
4.3.2.1 Population. From 1960 to 1990, population growth in the region of influence
mirrored statewide growth. During this period, the region's population increased at an average annual
rate of approximately 1.3 percent, while the growth rate for the State was 1.4 percent. Between 1980
and 1990, population growth in the region of influence approximately equaled that of the State with an
average growth rate of 0.6 percent per year. The region of influence had a 1990 population of
219,713, which comprised 22 percent of the total State population of 1,006,749. Based on population
and employment trends, the population in the region of influence will reach approximately
248,000 persons by 1995 (Table 4.3-1).
Figure 4.3-1. Historic and projected employment at the Idaho National Engineering Laboratory, 1990- 2004.
In 1990, the most populous counties were Bannock and Bonneville, which together contained
over 60 percent of the seven-county total (Figure 4.3-2). Butte and Clark were the least populous of
the counties in the region of influence. The largest cities in the region of influence are Pocatello and
Idaho Falls, with 1990 populations of approximately 46,000 and 44,000, respectively. In 1990, the
Fort Hall Indian Reservation and Trust Lands contained 5,113 residents, most of whom (52 percent)
resided in Bingham County.
4.3.2.2 Housing. Bonneville and Bannock Counties (which respectively include the cities of
Idaho Falls and Pocatello) provided 67 percent of the 73,230 year-round housing units in the region of
influence in 1990 (see Table 4.3-2). Of this number, approximately 70 percent were single-family
units, 17 percent were multifamily units, and 13 percent were mobile homes. Most of the multifamily
units (75 percent) were in Bonneville and Bannock Counties. About 29 percent of the occupied
housing units in the region were rental units and 71 percent were homeowner units (USBC 1992).
The median value of owner-occupied housing units ranged from $37,300 in Clark County to
$68,700 in Madison County, and median monthly rents ranged from $243 in Butte County to $366 in
Bonneville County. In 1990, there were 1,510 occupied housing units on the Fort Hall Indian
Reservation and Trust Lands (USBC 1992) and a vacancy rate of 14 percent.
4.3.3 Community Services
This assessment considers the following selected community services in the region of influence:
public schools, law enforcement, fire protection, hospital services, and solid waste disposal.
Table 4.3-3 summarizes pertinent characteristics of these services for the region of influence.
Seventeen public school districts and three nonpublic schools provide educational services for
about 58,000 children in the region of influence. Of these students, about 6,500 were dependents of
INEL-related employees. During the 1990-1991 academic year, most public school districts spent an
average of $3,000 to $4,000 per student annually. Higher education in the region is provided by the
University of Idaho, Idaho State University, Brigham Young University, Ricks College, and the
Eastern Idaho Technical College.
Seven county sheriff's offices, 12 city police departments, and the Idaho State Police provide law
enforcement services in the region. There was a total of 479 sworn officers and 100 other law
Figure 4.3-2. Historic and projected total population for the counties of the region of influence, 1940 through 2004.
Table 4.3-2. Number of housing units, vacancy rates, median house value, and median monthly rent
by county and region of influence.
Homeowner housing units Rental units
County Number of Vacancy rates Median value Number Vacancy rates Median
units ($) of units monthly rent
($)
Bannock 16,447 2.4 53,300 7,467 10.3 294
Bingham 9,010 2.0 50,700 2,955 9.2 284
Bonneville 17,707 1.9 63,700 7,375 6.2 366
Butte 780 4.6 41,400 302 16.2 243
Clark 177 1.7 37,300 114 9.6 281
Jefferson 4,000 2.0 54,300 992 4.1 314
Madison 3,522 1.3 68,700 2,392 2.8 299
Region of
influence 51,674 2.1 - 21,556 4.6 -
a. Source: USBC (1992).
enforcement personnel in 1991, more than 59 percent of whom served Bannock and Bonneville
Counties.
Eighteen fire districts in the region of influence operate 30 fire stations staffed by 180 paid and
approximately 300 volunteer firefighters. Bingham, Bonneville, Butte, Clark, and Jefferson Counties,
which surround the INEL, have developed emergency plans to be implemented in the event of a
radiological or hazardous materials emergency. Each emergency plan identifies facilities with
extremely hazardous substances and defines transportation routes for these substances. The emergency
plans also include procedures for notification and response, listings of emergency equipment and
facilities, evacuation routes, and training programs.
Eight hospitals serve the region of influence with more than 900 licensed beds and a capacity of
nearly 128,000 patient-days per year. Occupancy rates range from 22.0 to 61.7 percent in the region
(IDHW 1990). County governments and the Blackfoot, Dubois, Idaho Falls, and Pocatello fire
departments provide regional ambulance services. A private ambulance company serves residents in
Butte County. Four quick-response units, two medical helicopters, and two clinics specializing in
emergency medical services also serve the region of influence (Hardinger 1990; U.S. West Directories
1992).
Table 4.3-3. Summary of public services available in the region of influence.
County
Public Service Bannock Bingham Bonneville Butte Clark Jefferson Madison
Schools
Number of public school districts 2 5 3 1 1 3 2
Total enrollment 15,455 11,311 17,896 765 166 5,339 5,967
Number of INEL-related students (excluding 485 1,532 4,040 301 5 134 47
military)
Health Care Delivery
Number of hospitals 3 2 1 1 0 0 1
Number of licensed beds 309 238 311 4 - - 52
Law Enforcement
Number of sworn law enforcement officers 151 65 143 4 2 18 43
Total personnel per 1000 population 2.5 2.0 2.2 1.3 6.3 1.6 1.9
Fire Protection
Number of fire stations 9 7 6 2 1 4 1
Number of firefighters 166 96 121 15 7 63 24
Number of firefighting vehicles 37 25 24 3 1 11 6
Municipal Solid Waste Disposal
Number of landfills meeting EPAb regulations1c 3d 1e 2 0f 1 0f
Expected lifespan in years 30 3-6 50 30 - 2 -
a. Source: IDE (1991); IDHW (1990); IDLE (1991); Kouris (1992a); and Kouris (1992b).
b. EPA = U.S. Environmental Protection Agency.
c. Fort Hall Mine Landfill is being redesigned to meet EPA standards.
d. Aberdeen Landfill may close due to noncompliance with EPA standards.
e. A new landfill is replacing Bonneville County Landfill.
f. Madison and Clark Counties are evaluating a regional landfill for use after 1993.
Municipal solid waste generated in the region of influence is transported to county landfills. In
1992, twelve landfills served the region of influence. Four landfills (one each in Bannock, Clark,
Jefferson, and Madison Counties) will close without replacement before reaching their planned
capacity due to noncompliance with new Environmental Protection Agency standards (CFR 1991a).
4.3.4 Public Finance
In Fiscal Year 1991, total county revenues for the region of influence amounted to approximately
$90 million (see Table 4.3-4). County governments receive most of their revenues from taxes and
intergovernmental transfers. In 1991 the total assessed value of taxable property in the region of
influence was about $4.5 billion. In addition to property tax revenues, local governments (cities and
counties) also receive revenue from sales tax disbursements and revenue-sharing programs. These two
sources provide approximately 60 to 85 percent of the total revenues received by each county.
Table 4.3-4. Total revenues and expenditures by county, Fiscal Year 1991.
County Total Total
revenues ($) expenditures ($)
Bannock 16,232,274 14,216,708
Bingham 11,434,200 10,708,011
Bonnevilleb 50,186,650 51,850,100
Butte 1,417,684 1,397,012
Clark 1,236,849 1,086,379
Jefferson 4,408,236 4,566,074
Madison 5,249,432 5,662,080
Seven-county region 90,165,325 89,486,364
a. Sources: Ghan (1992); Bingham County (circa 1992); McFadden (circa 1992); Swager & Swager
(1992a); Swager & Swager (1992b); Draney, Searle, and Associates (1992); Schwendiman &
Sutton (1992).
b. Bonneville County's financial statements and total revenue data include special accounts for
schools, cities, cemeteries, fire districts, ambulance districts, and other special accounts not found in
other county budgets. The majority of intergovernmental revenue is used to fund these accounts.
Although DOE as a Federal agency is exempt from paying state or local taxes, INEL employees
and contractors are not. In 1992, INEL employees paid an estimated $60 million in Federal
withholding tax and $24 million in state withholding tax.
In 1991 the major categories of county government expenditures were general government
services, 27 percent; road maintenance, 18 percent; public safety, 16 percent; health and welfare
programs, 16 percent; sanitation and public works, 9 percent; debt service, 3 percent; trust remittances,
2 percent; and other expenditures, 9 percent.
4.4 Cultural Resources
This section discusses cultural resources at the INEL, including prehistoric and historic
archeological sites and historic sites and structures, and traditional resources that are of cultural or
religious importance to local Native Americans. It also discusses paleontological localities on the
INEL site.
4.4.1 Archeological Sites and Historic Structures
As summarized in the INEL Draft Management Plan for Cultural Resources (Miller 1992), the
INEL contains a rich and varied inventory of cultural resources. This includes fossil localities that
provide an important paleontological context for the region and the many prehistoric archeological
sites that are preserved within it. These latter sites, including campsites, lithic workshops, cairns, and
hunting blinds, among others, are also an important part of the INEL inventory because they provide
information about the activities of aboriginal hunting and gathering groups who inhabited the area for
approximately 12,000 years. In addition, archeological sites, pictographs, caves, and many other
features of the INEL landscape are also important to contemporary Native American groups for
historic, religious, and traditional reasons. Historic sites, including the abandoned town of
Powell/Pioneer, a northern spur of the Oregon Trail known as Goodale's Cutoff, many small
homesteads, irrigation canals, sheep and cattle camps, and stage and wagon trails, document the use of
the area during the late 1800s and early 1900s. Finally, the many scientific and technical facilities
inside the INEL boundaries have preserved important information on the historic development of
nuclear science in America.
To date, more than 100 cultural resource surveys have been conducted over approximately
4 percent of the area on the INEL site. These surveys, most of which have occurred near major
facility areas, have identified 1,506 archeological resources, including 688 prehistoric sites, 38 historic
sites, 753 prehistoric isolates, and 27 historic isolates (Miller 1992; Gilbert and Ringe 1993). These
numbers do not include architectural properties associated with the creation and operation of the INEL.
Until formal significance evaluations (archeological testing and historic records searches) have been
completed, all cultural sites in this inventory are considered to be potentially eligible for nomination to
the National Register of Historic Places. However, all the isolates have been categorized as unlikely
to meet eligibility requirements (Yohe 1993).
Due to the relatively high density of prehistoric sites on the INEL and the need to consider these
resources during Federal undertakings, DOE has sponsored a preliminary study, which resulted in the
development of a predictive model, to identify areas where densities of sites are highest and where the
potential impacts to significant archeological resources, as well as costs of compliance, would increase
correspondingly (Ringe 1993). This information provides guidance for INEL project managers in the
selection of appropriate areas for new construction. However, it does not take the place of inventories
that are required by the National Historic Preservation Act before ground-disturbing projects can start
(NHPA 1966 as amended).
The predictive model, constructed using a multivariate statistical technique on environmental
variables associated with areas with and without sites, indicates that prehistoric cultural resources
appear to be concentrated in association with certain definable physical features of the land. In this
context, very high densities of resources are likely to occur along the Big Lost River and Birch Creek,
atop buttes, and within craters and caves. The Lemhi Mountains, the Lake Terreton basin, and a 1.75-
mile- (2,800-meter-) wide zone along the edge of local lava fields probably contain a fairly high
density of sites. Within the extensive flows of basaltic lava and along the low foothills of the Lemhi
Mountains, site density is classified as moderate, and the lowest density of prehistoric resources
probably occurs in the floodplain of the Big Lost River and the alluvial fans emerging from the Birch
Creek Valley, in the sinks, and in the recent Cerro Grande lava flow. However, a classification of low
or medium density does not eliminate the possibility that significant resources exist in those areas.
Although the predictive model has not been tested, it is useful as a planning guide for defining areas
most likely to contain archeological resources based on past surveys.
Although there has been no systematic inventory of historically significant facilities associated
with the creation and operation of the INEL, a preliminary study indicated that all INEL facilities will
require evaluation (Braun et al. 1993). The Experimental Breeder Reactor-I is a National Historic
Landmark listed in the National Register of Historic Places. To date, however, few of the other
properties have been formally evaluated for eligibility to the National Register. Memoranda of
Agreement between DOE, the Idaho State Historic Preservation Office, and the National Advisory
Council on Historic Preservation establish that certain structures at Test Area North (DOE 1993b) and
Auxiliary Reactor Area (DOE 1993a) are eligible for nomination, and outline specific techniques for
preserving the historic value of the areas in conformance with the requirements of the Historic
American Building Survey and the Historic American Engineering Record. Other facilities on the
INEL site are likely to require similar efforts if DOE schedules them for major modification,
demolition, or abandonment.
4.4.2 Native American Cultural Resources
Because Native American people believe the land is sacred, the entire INEL reserve is culturally
important to them. Cultural resources, to the Shoshone-Bannock peoples, include all forms of
traditional lifeways and usage of all natural resources. This includes not only prehistoric archeological
sites, which are important in a religious or cultural heritage context, but also features of the natural
landscape, air, plant, water, or animal resources that might have special significance. These resources
may be affected by changes in the visual environment (construction, ground disturbance, or
introduction of a foreign element into the setting), dust particles, or by contamination. Geographically,
the INEL is included within a large territory once inhabited by and still of importance to the
Shoshone-Bannock Tribes. Plant resources used by the Shoshone-Bannock Tribes that are located on
or near the INEL site are listed in Table 4.4-1. Areas significant to the tribes would include the
buttes, wetlands, sinks, grasslands, juniper woodlands, Birch Creek, and the Big Lost River.
Five Federal laws prompt consultation between Federal agencies and Indian Tribes: the National
Environmental Policy Act (NEPA 1969), the National Historic Preservation Act (NHPA 1966 as
amended), the American Indian Religious Freedom Act (AIRFA 1978), the Archeological Resources
Protection Act (ARPA 1979), and the Native American Graves Protection and Repatriation Act
(NAGPRA 1990). In accordance with these directives and in consideration of its Native American
Policy (DOE 1990a and DOE 1992a), DOE is developing procedures at the INEL for consultation and
coordination with the Shoshone-Bannock Tribes of the Fort Hall Reservation. DOE has committed to
additional interaction and exchange of information with the Shoshone-Bannock Tribes, and has
outlined this relationship in a formal Working Agreement with these tribes (DOE 1992c). In addition,
the Cultural Resources Management Plan for the INEL (Miller 1992) and the curation agreement for
permanent storage of archaeological materials will be completed by June 1996. The Cultural
Resources Management Plan will define procedures for involving the tribes during the planning stages
of project development and the curation agreement will provide for the repatriation of burial goods in
accordance with NAGPRA.
4.4.3 Paleontological Resources
There are 31 known fossil localities at the INEL site. Available information suggests that the
region has relatively abundant and varied paleontological resources. Preliminary analyses suggest that
Table 4.4-1. Plants used by the Shoshone-Bannock tribes that are located on or near the INEL.
Plant Family Type of Use Location Abundance
Desert Parsley medicine, food scattered over site common
Milkweed food, tools roadsides scattered, uncommon
Sagebrush medicine, tools throughout the site common, abundant
Balsamroot food, medicine around buttes common but scattered
Thistle food scattered throughout site common but scattered
Gumweed medicine disturbed areas common
Sunflower medicine, food roadside common
Dandelion food, medicine throughout site common
Beggar's Ticks food disturbed areas throughout site common, abundant
Tansymustard food, medicine disturbed areas common
Cactus food throughout the site common, abundant
Honeysuckle food, tools Big Southern Butte common on butte
Goosefoot food throughout site common, abundant
Russian Thistle food disturbed areas throughout site common, abundant
Dogwood food, medicine, tools Webb Springs, Birch Creek common where found
Juniper medicine, food, tools throughout site common to abundant
Gooseberry food scattered throughout site common
Mentha arvensis medicine Big Lost River uncommon
Wild onion food, medicine, dye throughout site common
Caloehortus spp. food buttes common
Fireweed food throughout site common
Pine food, tools, medicine Big Southern Butte common on butte
Douglas Fir medicine Big Southern Butte common on butte
Plantain medicine, food throughout site uncommon
Wildrye food, tools throughout site common, abundant
Indian Ricegrass food throughout site common, abundant
Bluegrass food, medicine throughout site common, abundant
Serviceberry food, tools, medicine buttes common where found
Chokeberry food, medicine, tools, fuel buttes common where found
Wood's Rose food, smoking, medicine, Big Lost River, Big common, abundant
ritual Southern Butte
Red Raspberry food, medicine Big Southern Butte uncommon
Willow medicine throughout site in moist areas common
Coyote Tobacco smoking, medicine Big Lost River, Webb Springs uncommon
Cattail food, tools sinks, outflow from facilities uncommon
Source: Andersen et al. (1995).
these materials are most likely to occur in association with archeological sites; in areas of basalt flows;
in deposits of the Big Lost River, Little Lost River, and Birch Creek; in deposits of Lake Terreton and
playas; in some wind and sand deposits; and in sedimentary interbeds or lava tubes within local lava
flows (Miller 1992).
4.5 Aesthetic and Scenic Resources
4.5.1 Visual Character of the INEL Site
The Bitterroot, Lemhi, and Lost River mountain ranges border the INEL site on the north and
west. Persons can see volcanic buttes near the southern boundary of the INEL from most locations on
the site and from the Fort Hall Reservation. Most of the INEL site consists of open undeveloped land,
covered predominantly by large sagebrush and grasslands (see Section 4.9). Pasture and irrigated
farmland border much of the INEL site (see Section 4.2).
Although the INEL has a master plan, it has not established specific visual resource standards.
The nine facility areas on the INEL site are generally of low density, look like commercial or
industrial complexes, and are spread across the site. Structures in the facility areas range in height
from 10 feet to approximately 100 feet (3 to 30 meters). About 90 miles (145 kilometers) of paved
public highway run through the INEL site (see Section 4.11). Although many INEL facilities are
visible from these highways, most facilities are located more than 0.5 mile (0.8 kilometer) from public
roads.
4.5.2 Scenic Areas
The Craters of the Moon National Monument is about 15 miles (24 kilometers) southwest of the
INEL site's western boundary. The Monument is located in a designated Wilderness Area, which
must maintain Class I (very high) air quality standards or minimal degradation, as defined by the
Clean Air Act (CAA 1990; CFR 1990; CFR 1991b). Under Section 169a of the Clean Air Act, air
quality includes visibility and scenic view considerations.
Lands adjacent to the INEL under Bureau of Land Management jurisdiction are Visual Resource
Management Class II areas (BLM 1984; BLM 1986), which urge preservation and retention of the
existing character of the landscape. Lands inside the INEL boundaries are Class III and IV areas, the
most lenient classes in terms of modification. The Bureau of Land Management is considering the
Black Canyon Wilderness Study Area, which is adjacent to the INEL, for a Wilderness Area
designation (BLM 1986); if approved, this would result in an upgrade from Visual Resource
Management Class II to a Class I.
Features of the natural landscape have special significance to the Shoshone-Bannock tribes. The
visual environment of the INEL site is within the visual range of Fort Hall Reservation.
4.6 Geology
This section describes the geology of the INEL and the surrounding area. Section 4.6.1
characterizes the general geology, while section 4.6.2 describes the natural resources of the area.
Sections 4.6.3 and 4.6.4 describe seismic and volcanic hazards, respectively.
4.6.1 General Geology
The site is on the Eastern Snake River Plain (Figure 4.6-1). The Plain forms a broad northeast-
trending, crescent-shaped trough with low relief composed primarily of surface basaltic lava flows
formed 1.2 million to 2,100 years ago. The Plain features thin, discontinuous, and interbedded
deposits of wind-blown loess and sand; water-borne alluvial fan, lacustrine, and floodplain alluvial
sediments; and rhyolitic domes formed 1,200,000 to 300,000 years ago (Kuntz et al. 1990)
(Figure 4.6-2). Mountains and valleys of the Basin and Range Province, which trend north to
northwest and consist of folded and faulted rocks that are more than 70 million years old, bound the
Plain on the north and south. The Yellowstone Plateau bounds the Plain on the northeast. The major
episode of Basin and Range faulting began 20 to 30 million years ago and continues today, most
recently associated with the October 28, 1983, Borah Peak earthquake [moment magnitude 6.9,
magnitude 7.3 on the Richter scale with a resulting peak ground acceleration of 0.022 to 0.078 at the
INEL (Jackson 1985)], which occurred along the Lost River fault, approximately 100 kilometers
(62 miles) from site facilities and the 1959 Hebgen Lake Earthquake, moment magnitude 7.5,
approximately 150 kilometers (93 miles) from the INEL (Figure 4.6-1).
The northeast-trending volcanic terrain of the Plain has a markedly different geologic history and
tectonic pattern than the folded and faulted terrain of the northwest-trending Basin and Range. The
Basin and Range faults have not been observed on or across the Plain. Four northwest-trending
volcanic rift zones, attributed to basaltic eruptions that occurred 4 million to 2,100 years ago, lie
across the Plain at the INEL (Bowman 1995; Hackett and Smith 1992; Kuntz et al. 1990).
The seismic characteristics of the Eastern Snake River Plain and the adjacent Basin and Range
Province are also different. Earthquakes and active faulting are associated with the Basin and Range
tectonic activity. The Plain has historically experienced few and small earthquakes (King et al. 1987;
Pelton et al. 1990; WCC 1992; Jackson et al. 1993).
In 1979 the INEL drilled a geothermal exploration well to 3,159 meters (10,365 feet).
Researchers measured a temperature of 142yC (288yF) but identified no commercial quantities of
geothermal fluids (IDWR 1980). Mineral resources include several quarries or pits inside the INEL
boundary that supply sand, gravel, pumice, silt, clay, and aggregate for road construction and
maintenance, new facility construction and maintenance, waste burial activities, and ornamental
landscaping cinders. During excavations, DOE might study the gravel pits to characterize the local
surficial geology of the site. Outside the site boundary, mineral resources include sand, gravel,
pumice, phosphate, and base and precious metals (Strowd et al. 1981; Mitchell et al. 1981). The
geologic history of the Plain makes the potential for petroleum production at the INEL very low.
4.6.3 Seismic Hazards
The distribution of earthquakes at and near the INEL from 1884 to 1989 clearly shows that the
Plain has a remarkably low rate of seismicity, whereas the surrounding Basin and Range has a fairly
high rate (Figure 4.6-3, WCC 1992). The mechanism for faulting and generation of earthquakes in the
Basin and Range is attributed to northeast-southwest directed crustal extension.
Several investigators have suggested hypotheses for the low rate of seismic activity within the
Plain compared to the activity in both the Centennial Tectonic Belt and the Intermountain Seismic
Belt:
- Smith and Sbar (1974) and Brott et al. (1981) suggest that high crustal temperatures beneath
the Plain and adjacent region inside the seismic parabola (Figure 4.6-1) result in ductile
deformation (aseismic creep), in contrast to the brittle deformation (rock fracture) that occurs
in the Basin and Range.
- Anders et al. (1989) suggest that the Plain and the adjacent region inside the seismic
parabola (Figure 4.6-1) have increased integrated lithospheric strength. They propose that
the presence of mid-crustal basic intrusive rock strengthens the crust so that it is too strong
to fracture (see also Smith and Arabasz 1991).
Figure 4.6-3. Earthquakes with magnitudes greater than 2.5 from 1884 to 1989. - Parsons and Thompson (1991) propose that magma dike injection suppresses normal faulting
and associated seismicity by altering the local tectonic stress field. As dikes are injected in
volcanic rift zones, they push apart the surrounding rocks and decrease differential stress,
thereby preventing earthquakes from occurring.
- Anders and Sleep (1992) propose that the introduction of mantle-derived magma into the
midcrust beneath the Plain has decreased faulting and earthquakes by lowering the rate of
deformation.
The markedly different tectonic and seismic histories of the Plain and Basin and Range provinces
reflect the dissimilar deformational processes acting in each region. Both regions are subjected to the
same extensional stress field (Weaver et al. 1979; Zoback and Zoback 1989; Pierce and Morgan 1992;
Jackson et al. 1993); however, crustal deformation occurs through dike injection in the Plain and
through large-scale normal faulting in the Basin and Range (Rodgers et al. 1990; Parsons and
Thompson 1991; Hackett and Smith 1992).
Major seismic hazards include the effects from ground shaking and surface deformation (faulting,
tilting). Other potential seismic hazards (e.g., avalanches, landslides, mudslides, soil settlement,
and soil liquefaction) are not likely to occur at the INEL because the local geologic conditions are not
conducive to them. Based on the seismic history and the geologic conditions, earthquakes greater than
moment magnitude 5.5 (and associated strong ground shaking and surface fault rupture) are not likely
to occur in the Plain. However, moderate to strong ground shaking from earthquakes in the Basin and
Range can affect the INEL. Researchers use patterns of seismicity and locations of mapped faults to
assess potential sources of future earthquakes and to estimate levels of ground motion at the site.
The sources and maximum magnitudes of earthquakes that could produce the maximum levels of ground
motions at all INEL facilities include the following (WCC 1990; WCC 1992):
- A moment magnitude 7.0 earthquake at the southern end of the Lemhi fault along the Howe
and Fallert Springs segments
- A moment magnitude 7.0 earthquake at the southern end of the Lost River fault along the
Arco segment
- A moment magnitude 5.5 earthquake associated with dike injection in either the Arco or
Lava Ridge-Hell's Half Acre Volcanic Rift Zone and the Axial Volcanic Zone
- A "random" moment magnitude 5.5 earthquake occurring in the Eastern Snake River Plain
Figure 4.6-4 shows a facility-specific example of the relationship of the peak ground acceleration
on the INEL to the annual frequency of occurrence of seismic events on various seismic sources in the
region, including the four events described above (WCFS 1993). The curves refer specifically to the
site of the Idaho Chemical Processing Plant in the south-central INEL and might not apply directly to
other INEL areas. Ground motion contributions from seismic sources not shown on Figure 4.6-4
(i.e., Intermountain seismic belt and Yellowstone Region) are significantly smaller because of their
distant locations or lower estimated maximum magnitudes. The INEL Natural Phenomena Committee
determines INEL seismic design-basis events based on studies such as those performed by Woodward
Clyde Consultants (1990) and Woodward Clyde Federal Services (1993).
A maximum horizontal ground surface acceleration of 0.24g at the Idaho National Engineering
Laboratory is estimated to result from an earthquake that could occur once every 2,000 years (DOE
1994). 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, standards, and site-specific
procedures. Section 5.15 describes the potential impacts of postulated seismic events.
4.6.4 Volcanic Hazards
Volcanic hazards at the INEL can come from sources inside or outside Plain boundaries. These
hazards include the effects of lava flows, ground deformation (fissures, uplift, subsidence), volcanic
earthquakes (associated with magmatic processes as distinct from earthquakes associated with
tectonics), and ash flows or airborne ash deposits (Bowman 1995). Most of the basalt volcanic
activity occurred from 4 million to 2,100 years ago in the INEL area. The most recent and closest
volcanic eruption occurred 2,100 years ago at the Craters of the Moon, 25 kilometers (15 miles)
southwest of the INEL (Kuntz et al. 1992). The rhyolite domes along the Axial Volcanic Zone formed
between 1.2 million and 300,000 years ago and have a recurrence interval of about 200,000 years.
Therefore, the probability of future dome formation affecting INEL facilities is very low.
Figure 4.6-4. Contribution of the seismic sources to the mean peak acceleration at the Idaho Chemical Processing Plant.
Catastrophic Yellowstone eruptions have occurred three times in the past 2 million years, but the
INEL is more than 160 kilometers (70 miles) from the Yellowstone Caldera rim and high-altitude
winds would not disperse Yellowstone ash in the direction of INEL. Due to the infrequency, great
distance, and unfavorable dispersal, pyroclastic flows or ash fallout from future Yellowstone eruptions
should not impact the INEL.
Basaltic lava flows and eruptions from fissures or vents might occur. Based on a probability
analysis of the volcanic history in the Big Southern Butte area (Volcanism Working Group 1990), the
conditional probability that basaltic volcanism would affect a south-central INEL location is less than
2.5 y 10-5 per year (once per 40,000 years or longer), where the risk associated with Axial Volcanic
Zone volcanism is greatest. The estimated probability of volcanic impact on INEL facilities farther
north, where both silicic and basaltic volcanism have been older and less frequent, is less than 10-6 per
year (once every million years or longer). The statistics of 116 measured INEL-area lava flow lengths
and areas were used to define the two lava flow hazard zones (Figure 4.6-5). The hazard for a
particular site within or near a volcanic zone is much lower, typically by an order of magnitude or
more, and must be assessed on a site-specific basis (Bowman 1995).
Figure 4.6-5. Map of the INEL showing locations of volcanic rift zones and lava flow hazard zones. 4.7 Air Quality
This section describes the air resources of the INEL site and the surrounding area. The
discussion includes the climatology and meteorology of the region, descriptions of nonradiological and
radiological air contaminant emissions, and a characterization of existing and projected levels of air
pollutants. The analysis includes both existing facilities and those that were expected (at the time the
analysis was performed) to be operational before June 1, 1995. Additional detail and background
information on the material presented in this section is presented in Appendix F, Section F-3, of
Volume 2.
4.7.1 Climatology and Meteorology
The Eastern Snake River Plain climate exhibits low relative humidity, wide daily temperature
swings, and large variations in annual precipitation. Average seasonal temperatures measured on the
INEL site range from -7.3yC (18.8yF) in winter to 18.2yC (64.8yF) in summer, with an annual average
temperature of about 5.6yC (42yF). Temperature extremes range from a summertime maximum of
39.4yC (103yF) to a wintertime minimum of -45yC (-49yF). The annual average relative humidity is
50 percent, with monthly average maximum values ranging from 59 percent in July to 89 percent in
February and December, and with monthly average minimum values ranging from 16 percent in June
and July to 47 percent in January (Clawson et al. 1989).
Annual precipitation is light, averaging 221.2 millimeters (8.71 inches), with monthly extremes
of zero to 127 millimeters (5 inches). The maximum 24-hour precipitation rate is 46 millimeters
(1.8 inches). The greatest short-term precipitation rates are attributable primarily to thunderstorms,
which occur approximately two or three days per month during the summer. The average annual
snowfall is 701 millimeters (27.6 inches), with a maximum of 1,516 millimeters (59.7 inches) and a
minimum of 173 millimeters (6.8 inches) (Clawson et al. 1989).
The INEL site is in the belt of prevailing westerlies; however, the mountain ranges bordering the
Eastern Snake River Plain normally channel these winds into a southwest wind. Most offsite locations
experience the predominant southwest-northeast wind flow of the Eastern Snake River Plain, although
subtle terrain features near some locations cause considerable variations from this flow regime. The
annual average wind speed measured at the 6.1-meter (20-foot) level at the Central Facilities Area
Weather Station is 3.4 meters per second (7.5 miles per hour). Monthly average values range from
2.3 meters per second (5.1 miles per hour) in December to 4.2 meters per second (9.3 miles per hour)
in April and May (Clawson et al. 1989). The highest hourly average near-ground wind speed
measured onsite is 22.8 meters per second (51 miles per hour) from the west-southwest, with a
maximum instantaneous gust of 34.9 meters per second (78 miles per hour) (Clawson et al. 1989).
Figure 4.7-1 presents the frequency of wind speed and wind direction at three meteorological
monitoring sites on the INEL site from 1988 to 1992. The wind directions presented in the figure are
the direction from which the wind blows. The three wind-roses demonstrate the effects of terrain on
predominant wind directions and wind speed. The winds at the Test Area North monitoring station are
predominantly from the north-northwest, whereas the winds from the other stations are predominantly
from the southwest.
Air pollutant dispersion is a result of the processes of transport and diffusion of airborne
contaminants in the atmosphere. Transport is the movement of a pollutant in the wind field, while
diffusion refers to the process whereby turbulent eddies dilute a pollutant plume. The temperature
gradient of the atmosphere (i.e., the change in temperature with altitude) can restrict or enhance the
vertical diffusion of pollutants. Lapse rate conditions, which tend to enhance vertical diffusion, occur
slightly less than 50 percent of the time. Conversely, thermal stratification or inversion conditions,
which inhibit vertical diffusion, occur slightly more than 50 percent of the time. The height to which
the pollutants can freely diffuse is the mixing depth, while the layer of air from the ground to the
mixing depth is the mixed layer. Estimates of the monthly average depth of the mixed layer range
from 400 meters (1,312 feet) in December to 3,000 meters (9,843 feet) in July. With calm winds and
mostly clear skies, nocturnal inversions begin forming after sunset and dissipate about 1 to 2 hours
after sunrise. These inversions are often ground-based, meaning the atmospheric temperature increases
with height from the ground (Clawson et al. 1989).
Other than thunderstorms, severe weather is uncommon. Five funnel clouds (tornadoes not
touching the ground) and no tornadoes were reported on the site between 1950 and 1988. Visibility in
the region is good because of the low moisture content of the air and minimal sources of visibility-
reducing pollutants. From Craters of the Moon National Monument, the seasonal visual range is from
130 to 155 kilometers (81 to 97 miles) (Notar 1993).
4.7.2 Air Quality
4.7.2.1 Nonradiological Air Quality. The INEL is in the Eastern Idaho Intrastate Air
Quality Control Region (AQCR 61). Neither the INEL nor any of the surrounding counties is
Figure 4.7-1. Depiction of annual average wind direction and speed at INEL meteorological monitoring stations.
designated as a nonattainment area (CFR 1992b) for the National Ambient Air Quality Standards
(CFR 1991b). Ambient air quality data monitored in the vicinity of the INEL indicate that the site is
in compliance with applicable air quality standards (DOE 1991a).
The Clean Air Act (CAA 1990) contains requirements to prevent the deterioration of air quality
in areas designated to be in attainment with the ambient air quality standards. These requirements are
administered through a program that limits the increase in specific air pollutants above the levels that
existed in what has been termed a baseline (or starting) year, which is 1977. The requirements specify
maximum allowable ambient pollutant concentration increases or increments. They specify increment
limits for pollutant level increases for the nation as a whole (Class II areas) and prescribe more
stringent increment limits (as well as ceilings) for designated national resources, such as national
forests, parks, and monuments (Class I areas). Three areas in the INEL vicinity are Prevention of
Significant Deterioration Class I ambient air quality areas: Craters of the Moon Wilderness Area,
approximately 53 kilometers (33 miles) to the west-southwest; Yellowstone National Park,
approximately 143 kilometers (89 miles) to the northeast; and Grand Teton National Park,
approximately 145 kilometers (90 miles) to the east-northeast.
DOE evaluates proposed new and modified sources of emissions at INEL to determine the net
emissions increase of all pollutants. The INEL is considered a major source, because facility-wide
emissions of specific regulated air contaminants exceed 227 metric tons (250 tons) per year.
Therefore, a Prevention of Significant Deterioration analysis must be performed for all significant
emission increases of specified regulated pollutants. Levels of significance for net emission increases
range from very small quantities (less than 1 pound) for beryllium up to 91 metric tons (100 tons) per
year for carbon monoxide. Their significance is dependent on the toxicity of the substance. For
radionuclides, significance means any increase in emissions that would result in an offsite dose of 0.1
millirem per year or greater.
Ambient air quality standards for Idaho are the same as the National Ambient Air Quality
Standards but include total suspended particulates and fluorides. The Idaho Department of Health and
Welfare (IDHW) also has ambient concentration limits for hazardous and toxic air pollutants.
Table 4.7-1 lists emission rates of criteria and hazardous and toxic air pollutants.
The types and amounts of nonradiological emissions from INEL facilities and activities are
similar to those from other industrial complexes that are the same sizes as the INEL. Combustion
sources such as boilers and emergency generators emit both criteria and toxic pollutants. Other
Table 4.7-1. Baseline annual average and maximum hourly emission rates of nonradiological air
pollutants at the INEL.
Pollutant Annual average (kg/yr)b,c Maximum hourly (kg/hr)b
Criteria pollutants
Carbon monoxide (CO) 301,000 177
Lead (Pb) 11 0.085
Nitrogen dioxide (NO2) 744,000 545
Particulate matter (PM10)d 302,000 230
Sulfur dioxide (SO2) 202,000 136
Hazardous/toxic air pollutantse
Acetaldehyde 31 0.39
Ammonia 1,600 3.4
Arsenic 4.2 9.0 y 10-4
Benzene 370 16
1,3-Butadiene 220 0.8
Carbon tetrachloride 28 0.08
Chloroform 1.9 5.5 y 10-3
Chromium - trivalent 3.1 2.5 y 10-3
Chromium - hexavalent 0.4 6.2 y 10-4
Cyclopentane 350 0.58
Dichloromethane 620 0.29
Formaldehyde 960 8.9
Hydrazine 8.3 9.5 y 10-4
Hydrochloric acid 1,500 0.34
Mercury 200 0.023
Napthalene 16 2.2
Nickel 270 0.057
Nitric acid 1,500 1.7
Phosphorous 56 0.024
Potassium hydroxide 990 0.24
Propionaldehyde 62 0.24
Styrene 4.7 0.74
Tetrachlorethylene 980 0.11
Toluene 580 56
Trichloroethylene 4.7 0.013
Trimethylbenzene 87 12
a. Source: Volume 2, Table 4.7-2.
b. To convert kilograms to pounds, multiply by 2.2.
c. Annual average values include actual emissions plus projected increases from facilities that will
become opertional after the baseline year.
d. It is conservatively assumed that all particulate matter is PM10 (less than 10 microns in diameter).
e. Hazardous/toxic air pollutants that are listed in State of Idaho regulations and are emitted in levels
that exceed screening criteria.
sources include chemical processing operations, transportation, waste management activities, and
research laboratories.
Table 4.7-2 compares the INEL contribution to air quality to applicable standards and guidelines.
This assessment modelled the INEL air emissions inventory for 1990 using the methodology approved
by the U.S. Environmental Protection Agency to predict the maximum ground-level concentration that
would occur at or beyond the site boundary for each regulated pollutant (EPA 1993b). The Industrial
Source Complex-2 model primarily assessed criteria pollutants, and the SCREEN model assessed toxic
air pollutants. The SCREEN model incorporates meteorological data that tend to overestimate impacts,
and is useful for identifying cases that require additional, more refined assessments. The baseline
concentrations listed in Table 4.7-2 are the sums of the following factors: the concentrations resulting
from potential impacts from current operations and the concentrations resulting from the construction
or operation of planned upgrades or modifications before the implementation of the proposed actions
described in Section 5.7. Background concentrations have not been included because (a) reliable data
on background levels in the INEL environs are not available for most pollutants and (b) background
levels are low and are more than offset by the use of the maximum (as opposed to actual) baseline.
The baseline concentrations represent the maximum calculated concentration occurring at public access
locations (site boundary, public roads, and Craters of the Moon Wilderness Area). A comparison of
the baseline concentrations to applicable Federal and state criteria pollutant and hazardous/toxic air
pollutant guidelines and regulations shows that air quality at INEL is in compliance with those
guidelines and regulations. The 24-hour total suspended particulate background concentration is listed
as 40 micrograms per cubic meter, which is the same as the annual geometric mean value. The annual
sources include chemical processing operations, transportation, waste management activities, and
research laboratories.
4.7.2.2 Radiological Air Quality. The major source of radiation exposure in the Eastern
Snake River Plain is from natural background radiation sources such as cosmic rays; radioactivity
naturally present in soil, rocks, and the human body; and airborne radionuclides of natural origin (such
as radon). Sources of radioactivity related to INEL operations include research and training reactors,
spent nuclear fuel testing and stabilization, irradiated material and fuel examination, nuclear waste
treatment and storage, and depleted uranium armor production.
Radioactive emissions from INEL facilities include the noble gases (argon, krypton, and xenon)
and iodine; particulate fission products such as rubidium, strontium, and cesium; radionuclides formed
Table 4.7-2. Comparison of baseline ambient air concentrations with most stringent applicable
regulations and guidelines at the INEL.
Pollutant Averaging Most stringent Maximum Percent
time regulation or baseline of
guideline concentration standard
(-g/m3)a,b,c (-g/m3)
Criteria pollutants
Carbon monoxide (CO) 8-hour 10,000 280 2.8
1-hour 40,000 610 1.5
Lead (Pb) Calendar 1.5 0.001 <0.1
Quarter
Nitrogen dioxide (NO2) Annual 100 4 4
Particulate matter (PM10) Annual 50 5 10
24-hour 150 80 53
Sulfur dioxide (SO2) Annual 80 6 7.5
24-hour 365 140 37
3-hour 1,300 580 45
Hazardous/toxic air pollutants
Acetaldehyde Annual 4.5 y 10-1 1.1 y 10-2 2
Ammonia Annual 1.8 y 102 6.0 y 100 3
Arsenic Annual 2.3 y 10-4 9.0 y 10-5 39
Benzene Annual 1.2 y 10-1 2.9 y 10-2 24
Butadiene Annual 3.6 y 10-3 1.0 y 10-3 28
Carbon Tetrachloride Annual 6.7 y 10-2 6.0 y 10-3 9
Chloroform Annual 4.3 y 10-2 4.0 y 10-4 <1
Chromium - hexavalent Annual 8.3 y 10-5 6.0 y 10-5 72
Chromium - trivalent Annual 5.0 y 100 3.6 y 10-2 <1
Cylclopentane Annual 1.7 y 104 2.7 y 10-0 <1
Formaldehyde Annual 7.7 y 10-2 1.2 y 10-2 16
Hydrazine Annual 3.4 y 10-4 1.0 y 10-6 <1
Hydrochloric acid Annual 7.5 y 100 9.8 y 10-1 13
Mercury Annual 1.0 y 100 4.2 y 10-2 4
Methylene Chloride Annual 2.4 y 10-1 6.0 y 10-3 3
Napthalene Annual 5.0 y 102 1.8 y 101 4
Nickel Annual 4.2 y 10-3 2.7 y 10-3 65
Nitric Acid Annual 5.0 y 101 6.4 y 10-1 1
Table 4.7-2. (continued).
Pollutant Averaging Most stringent Maximum Percent
time regulation or baseline of
guideline concentration standard
(-g/m3)a,b,c (-g/m3)
Perchloroethylene Annual 2.1 y 100 1.1 y 10-1 5
Phosphorous Annual 1.0 y 100 3.0 y 10-1 30
Potassium hydroxide Annual 2.0 y 101 2.0 y 10-1 1
Proprionaldehyde Annual 4.3 y 100 3.0 y 10-1 7
Styrene Annual 1.0 y 103 1.3 y 100 <1
Toluene Annual 3.8 y 103 3.7 y 102 10
Trichloroethylene Annual 7.7 y 10-2 9.7 y 10-4 1
Trimethylbenzene Annual 1.2 y 103 1.0 y 102 8
a. CFR (1991b).
b. IDHW (1994); the ambient standards for the criteria pollutants are the same as the NAAQS.
c. Standards cited for hazardous/toxic air pollutants are for all new sources constructed or modified
since May 1, 1994, under State of Idaho Regulations for the Control of Air Pollution in the State of
Idaho (IDHW 1994).
Source: Volume 2, Section 4.7.
by neutron activation such as tritium (hydrogen-3), carbon-14, and cobalt-60; and very small quantities
(less than 6 y 10-4 curies per year) of heavy elements such as uranium, thorium, plutonium, and their
decay products. Historically, the radionuclide with the highest emission rate is the noble gas
krypton-85, which is released primarily by the chemical reprocessing of spent nuclear fuel at the Idaho
Chemical Processing Plant. Fuel reprocessing also releases small amounts (less than 0.1 curie per
year) of iodine-129, which is of concern because of its long half-life (16 million years) and biological
properties (iodine isotopes tend to accumulate in the human thyroid). Reactor operations release noble
gas isotopes with short half-lives, including argon-41 and isotopes of xenon (primarily xenon-133,
-135, and -138). Other activities at the INEL, including waste management operations, result in very
low levels of airborne radionuclide emissions (less than 1 y 10-4 curie per year). Table 4.7-3
summarizes airborne radionuclide emissions from INEL facility areas, plus estimated emissions from
projects expected, at the time of the analysis was performed, to become operational before June 1,
1995.
Radioactivity released to the atmosphere can result in human exposure through a number of
pathways, including inhalation, external exposure, and ingestion. DOE conducts physical
Table 4.7-3. Summary of airborne radionuclide emissions from INEL facility areas (curies per year).
Tritium/ Iodines Noble Mixed U/Th/TRUc
Facility carbon-14 gases fission and
activation
productsb
Argonne National 1.0 y 102 -d 1.3 y 104 8.1 y 10-4 1.8 y 10-6
Laboratory-West
Central Facilities Area 2.6 y 100 5.0 y 10-7 - 1.9 y 10-5 9.6 y 10-7
Idaho Chemical Processing 4.3 y 101 6.4 y 10-2 1.0 y 104 3.6 y 10-2 9.4 y 10-9
Plant
Naval Reactors Facility 1.9 y 10-1 6.3 y 10-6 5.7 y 10-1 5.6 y 10-5 -
Power Burst 4.9 y 101 - - 1.3 y 100 9.8 y 10-3
Facility/Waste
Experimental Reduction
Facility
Radioactive Waste - - - 2.6 y 10-5 4.2 y 10-6
Management Complex
Test Area North 1.2 y 10-1 - - 5.6 y 10-6 1.5 y 10-5
Test Reactor Area 1.6 y 102 1.6 y 10-2 3.3 y 103 3.0 y 100 1.8 y 10-6
INEL total 2.1 y 103 1.1 y 10-1 1.2 y 105 5.6 y 100 1.0 y 10-2
a. With the exception of the Idaho Chemical Processing Plant, emissions estimates are based on 1991
operations. Idaho Chemical Processing Plant emissions are based on 1993 emissions but are scaled
upward to reflect operation of the New Waste Calcining Facility at maximum permitted levels.
Anticipated projects in the baseline include the Waste Experimental Reduction Facility (compacting
and sizing operations but not incineration), Argonne National Laboratory-West Fuel Cycle Facility,
and Portable Water Treatment Unit, as described in Appendix F of Volume 2.
b. Mixed fission and activation products that are primarily particulate in nature (for example,
cobalt-60, strontium-90, and cesium-137).
c. U/Th/TRU = Radioisotopes of uranium, thorium, or transuranic elements such as plutonium,
americium, and neptunium.
d. A dash (-) indicates that the emissions for this group are negligibly small or zero.
Source: Volume 2, Table 4.7-1.
measurements (ambient air monitoring) and uses calculation techniques (atmospheric dispersion
modeling) to assess existing levels of radiation (both cosmic and manmade) in and near the site, and to
assess doses to workers and the surrounding population.
The offsite population can receive a radiation dose as a result of radiological conditions directly
attributable to existing INEL operations. DOE assesses such a dose for a maximally exposed
individual and for the population as a whole. The maximally exposed individual is a hypothetical
person whose habits and proximity to the site are such that the person would receive the highest dose
projected to result from sitewide radioactive emissions. The calculated annual dose to this individual
as a result of current and anticipated sitewide emissions is 0.05 millirem (Section 4.7 to Volume 2).
This value is a small fraction of both the National Emission Standards for Hazardous Air Pollutants
dose limit of 10 millirem per year (CFR 1992a) and the dose received from natural background
sources of 351 millirem per year (Section 4.7 to Volume 2). Figure 4.7-2 compares these dose rates.
The collective annual dose to the surrounding population, determined using 1990 U.S. Census
Bureau data for the total population residing within an 80-kilometer (50-mile) radius from each facility
on the site, is about 0.3 person-rem (Section 4.7 to Volume 2). This value is small in comparison to
the annual dose received by the same population from background sources, which is more than
40,000 person-rem (Section 4.7 to Volume 2).
Workers at each major INEL facility can receive radiation exposures. DOE has based its
assessment of the dose to these workers on contributions from sources at each facility and those
expected to become operational before June 1, 1995. The results of this assessment indicate that the
maximum dose received by a worker at any onsite area is about 4.3 millirem per year (Section 4.7 to
Volume 2), well below the National Emissions Standard for Hazardous Air Pollutants dose limit of
10 millirem per year. The standard applies to the highest exposed member of the public, and is not
applicable to workers. However, it is the most restrictive limit for airborne releases and provides a
useful comparison. This dose value of 4.3 millirem per year includes the maximum projected
operation of the Portable Water Treatment Unit at the Power Burst Facility Area. However, that
operation would be temporary (1 to 2 years) and is not representative of a permanent increase in the
baseline. If this facility were not included, the baseline dose to the worker would be about
0.2 millirem per year.
Figure 4.7-2. Comparison of dose to maximally exposed individual (MEI) to the National Emission Standard for Hazardous Air Pollutants (NESHAP) dose limit and the dose from background sources.
4.8 Water Resources
This section describes existing regional and site hydrologic conditions and discusses the quality
of surface and subsurface water and water use and rights. The subsurface water section also describes
the vadose zone (or unsaturated zone and perched water bodies) located between the land surface and
the water table.
4.8.1 Surface Water
Other than surface-water bodies formed from accumulated runoff during snowmelt or heavy
precipitation and manmade infiltration and evaporation ponds, there is little surface water at the site.
The following sections discuss regional drainage conditions, local runoff, floodplains, and
surface-water quality. Figure 4,8-1 supports discussions in this section.
4.8.1.1 Regional Drainage. The INEL is in the Pioneer Basin, a closed drainage basin that
includes three main surface-water bodies--the Big and Little Lost Rivers and Birch Creek. These
water bodies drain mountain watersheds directly west and north of the site. However, most of the
surface-water flow is diverted for irrigation before it reaches site boundaries (Barraclough et al. 1981),
resulting in little or no flow for several years inside the site boundaries (Pittman et al. 1988).
The Big Lost River drains approximately 3,755 square kilometers (1,450 square miles) of land
before reaching the site. Approximately 48 kilometers (30 miles) upstream of Arco, Idaho, Mackay
Dam controls and regulates the flow of the river, which continues southeast past the towns of Moore
and Arco and onto the Eastern Snake River Plain. The river channel then crosses the southwestern
boundary of the site, where the INEL Diversion Dam controls surface-water flow. During heavy
runoff events, the dam diverts surface water to a series of natural depressions, designated as spreading
areas. The Big Lost River continues northeasterly across the site to an area of natural infiltration
basins (playas or sinks) near Test Area North. In dry years, surface water does not usually reach the
western boundary of the site, and because the INEL is located in a closed drainage basin, surface
water never flows off the site.
Birch Creek drains an area of approximately 1,943 square kilometers (750 square miles). In the
summer, upstream of the site, surface water from Birch Creek is diverted to provide irrigation and
Figure 4.8-1. Selected facilities and predicted inundation map for probable maximum flood-induced overtopping failure of Mackay Dam at INEL.
to produce hydropower. In the winter, water flow crosses the northwest corner of the site, entering a
manmade channel 6.4 kilometers (4 miles) north of Test Area North, where it then infiltrates into
channel gravels.
The Little Lost River drains an area of approximately 1,826 square kilometers (705 square
miles). Strearnflow is diverted for irrigation north of Howe, Idaho. Surface water from the Little Lost
River has not reached the site in recent years; however, during high stream flow years, water will
reach the site and infiltrate into the subsurface (E(3&G 1984).
4.8.1.2 Local Runoff. Surface water generated from local precipitation will flow into
topographic depressions (lower elevations than the surrounding terrain) on the site. This surface water
either evaporates or infiltrates into the ground, increasing subsurface saturation and enhancing
subsurface migration (Wilhelmson et al. 1993).
Localized flooding can occur at the site when the ground is frozen and melting snow combines
with heavy spring rains. Test Area North was flooded in 1969 (Koslow and Van Haaften 1986), In
1969 extensive flooding caused by snowmelt occurred in the lower Birch Creek Valley (Koslow 1984)
Studies have shown that both the 25- and 100-year, 24-hour rainfall/snowmelt storm event could cause
flooding within the Radioactive Waste Management Complex (Dames & Moore 1992). The drainage
system, including dikes and erosion prevention features designed to mitigate potential surface water
flooding, are being upgraded.
4.8.1.3 Floodplains. Intermittent surface-water flow and the INEL Diversion Dam (built in
1958 and enlarged in 1984) have effectively prevented flooding from the Big Lost River onto the site.
However, onsite flooding from the river could occur if high water in the Mackay Dam or the Big Lost
River were coupled with a darn failure. Koslow and Van Haaften (1986) examined the consequences
of structural failure of the Mackay Dam due to a seismic event, coupled with a probable maximum
flood (the largest flood assumed possible in an area), This scenario predicts flood waters overtopping
the INEL Diversion Dam and spreading at the Idaho Chemical Processing Plant, Naval Reactors
Facility, and the Test Area North Loss-of-Fluid Test Facility (Figure 4.8-1). In the event of a
combined Mackay Dam failure and a 100-year flood (flood that occurs on an average of every
100 years), flooding along the Big Lost River would also occur, with low velocities and water depths
on the INEL (Koslow and Van Haaften 1986). The area inundated under the Mackay Dam failure
scenarios probably would use more than the 100- or 500-year floodplains for the Big Lost River at the
INEL. A 100-year floodplain study for the INEL is in progress.
4.8.1.4 Surface-Water Ouality, Water quality in the Big and Little Lost Rivers and Birch
Creek is similar and has not varied a great deal over the period of record. Measured physical,
chemical, and radioactive parameters have not exceeded applicable drinking water quality standards.
Chemical composition is determined primarily by the mineral composition of the rocks in the
mountain ranges northwest of the site and by the chemical composition of irrigation water in contact
with the surface water (Robertson et al. 1974; Bennett 1990).
Site activities do not directly affect the quality of surface water outside the site because
discharges from site facilities are to manmade seepage and evaporation basins or stormwater injection
wells. Effluents are not discharged to natural surface waters. In addition, surface water does not flow
directly off the site (Hoff et al. 1990). However, water from the Big Lost River, as well as seepage
from evaporation basins and stormwater injection wells, does infiltrate the Snake River Plain Aquifer
(Robertson et al. 1974; Wood and Low 1988; Bennett 1990). These areas are inspected, monitored,
and sampled as stipulated in the INEL Stormwater Pollution Prevention Program (DOE-ID 1 993b).
4.8.2 Subsurface Water
Subsurface water at the site occurs in the Snake River Plain Aquifer and the vadose zone. This
section describes regional and local hydrogeologic conditions, vadose zone hydrology, perched water,
and subsurface-water quality. Generally, the term "groundwater" refers to usable quantities of water
that enter freely into wells under confined and unconfined conditions within an aquifer (Driscoll 1989).
4.8.2. 1 Regional Hydrogeology. The INEL overlies the Snake River Plain Aquifer, the
largest aquifer in Idaho (Figure 4.8-2). This aquifer underlies the Eastern Snake River Plain and
covers an area of approximately 24,900 square kilometers (9,611 square miles). Groundwater in the
aquifer generally flows south and southwestward across the Snake River Plain. The estimated water
storage in the aquifer is 2.5 x 1012 cubic meters (2 billion acre-feet, which is about the same as the
volume of water contained in Lake Erie) (Robertson et al. 1974). A typical irrigation well can yield as
much as 13.9 x 106 cubic meters (3.7 x 10(9) gallons) per year of water if pumped every day
(Garabedian 1989). The Snake River Plain Aquifer is among the most productive aquifers in the
nation.
The drainage basin recharging the Snake River Plain Aquifer covers an area of approximately
90,643 square kilometers (35,000 square miles). The aquifer is recharged by infiltration of irrigation
Figure 4.8-2. Location of the INEL, Snake River Plain, and generalized groundwater flow direction of the Snake River Plain Aquifer.
water, seepage from stream channels and canals, underflow from tributary stream valleys extending
into the watershed, and direct infiltration from precipitation (Garabedian 1989). Most recharge occurs
in surface water-irrigated areas and along the northeastern margins of the plain. Groundwater
discharges primarily from the aquifer through springs that flow into the Snake River and from
pumping for irrigation. Major springs and seepages that flow from the aquifer are located near the
American Falls Reservoir (southwest of Pocatello) and the Thousand Springs area between Milner
Dam and King Hill (near Twin Falls).
4.8.2.2 Local Hydrogeology. The INEL site covers 2,305 square kilometers (890 square
miles) of the north-central portion of the Snake River Plain Aquifer. Depth to groundwater from the
land surface at the site ranges from approximately 61 meters (200 feet) in the north to over 274 meters
(900 feet) in the south (Pittman et al. 1988) (see Figure 4.8-3). Groundwater flow is generally toward
the south-southwest, and the upper surface is primarily unconfined (not overlain by impermeable soil
or bedrock). However, the aquifer behaves as if it were partially confined because of localized
geologic conditions. The occurrence and movement of groundwater in the aquifer depends on the
geologic setting and the recharge and discharge of water within that setting. Most of the aquifer
consists primarily of numerous relatively thin, basaltic lava flows with interbedded sediments
extending to depths of 1,067 meters (3,500 feet) below the land surface (Irving 1993). Most of the
groundwater migrates horizontally through fractured, basaltic interflow zones (broken and rubble
zones) that occur at various depths. Water also migrates vertically along joints and the interfingering
edges of interflow zones (Garabedian 1986). Sedimentary interbeds restrict the vertical movement of
groundwater. The variability in how the aquifer stores and transmits water increases the difficulty in
aquifer investigations and modeling.
The rate at which water moves through the ground depends on the hydraulic gradient (change in
elevation and pressure with distance in a given direction) of the aquifer, the effective porosity
(percentage of void spaces), and hydraulic conductivity (capacity of a porous media to transport water)
of the soil and bedrock. Because aquifer porosity and hydraulic conductivity decrease with depth,
most of the water in the aquifer moves through the upper 61 to 152 meters (200 to 500 feet) of the
basalts. Estimated flow rates within the aquifer range from 1.5 to 6.1 meters (5 to 20 feet) per day
(Barraclough et al. 1981).
The aquifer's ability to transmit water (transmissivity), and its ability to store water (storativity)
are important physical properties of the aquifer. In general, the hydraulic characteristics of the aquifer
enable the easy transmission of water, particularly in the upper portions.
Figure 4.8-3. Hydrostratigraphy scross the INEL and water table surface. Recharge to the aquifer originates off the site from precipitation in the mountains to the west and
north. Most of the inflow to the aquifer results from the underflow of groundwater along
alluvial-filled valleys adjacent to the Eastern Snake River Plain and adjacent surface-water drainages
(i.e., Big and Little Lost Rivers and Birch Creek). In addition, recharge at the site is related to the
amount of precipitation, particularly snowfall, for a given year (Barraclough et al. 1981).
4.8.2.3 Vadose Zone Hydrology The vadose (unsaturated) zone extends from the land
surface down to the water table. Within the vadose zone, water and air occupy openings in the
geologic materials. Subsurface water in the vadose zone is referred to as vadose water. At the site
this complex zone consists of surface sediments (primarily clay and silt, with some sand and gravel)
and many relatively thin basaltic lava flows, with some sedimentary interbeds. Thick surficial deposits
occur in the northern part of the site, which thin to the south where basalt is exposed at the surface.
The vadose zone protects the groundwater by filtering many contaminants through adsorption,
buffering dissolved chemical wastes, and slowing the transport of contaminated liquids to the aquifer.
The vadose zone also protects the aquifer by storing large volumes of liquid or dissolved contaminants
released to the environment through spills or migration from disposal pits or ponds, allowing natural
decay processes to occur.
Travel times for water through the vadose zone are important for an understanding of
contaminant movement. The flow rates in the vadose zone depend directly on the extent of fracturing,
the percentage of sediments versus basalt, and the moisture content of vadose zone material. Flow
increases under wetter conditions and slows under dryer conditions.
4.8.2.4 Perched Water. Locally, saturated conditions that exist above the water table are
called perched water. Perched water occurs when water migrates vertically and laterally from the
surface until it reaches an impermeable layer (Irving 1993). As perched water spreads laterally,
sometimes for hundreds of meters, it moves over the edges of the impermeable layer and continues
downward. Several perched water bodies can form between the land surface and the water table.
In general, perched water bodies slow the downward migration of fluids that infiltrate into the
vadose zone from the surface because the downward flow is not continuous. The occurrence of
perched water at the site is related to the presence of disposal ponds or other surface-water bodies,
which studies have detected at the Idaho Chemical Processing Plant, Test Reactor Area, and Test Area
North. For example, a 1986 field study at the Idaho Chemical Processing Plant showed that perched
water occurs in three areas at possibly three depth zones, ranging from approximately 9 meters
(30 feet) to 98 meters (322 feet) below the ground surface and extending laterally as much as
1 ,097 meters (3,600 feet). In general, the chemical concentrations, shape, and size of these bodies
have fluctuated over time in response to the volume of water discharged to the infiltration ponds
(Irving 1993).
4.8.2.5 Subsurface Water Quality. Natural water chemistry and contaminants originating at
the site affect subsurface water quality. The INEL Groundwater Protection Management Program
conducts monitoring programs. This program collects samples from surface water, perched water, and
aquifer wells to identify contaminants and contaminant migration to and within the aquifer.
4.8.2,5.1 Natural Water Chemistry - Several factors determine the natural groundwater
chemistry of the Snake River Plain Aquifer beneath the site. These factors include the weathering
reactions that occur as water interacts with minerals in the aquifer and the chemical composition of
(1) groundwater originating outside the site; (2) precipitation falling directly on the land surface; and
(3) streams, rivers, and runoff infiltrating the aquifer (Wood and Low 1986, 1988). The chemistry of
the groundwater is different, depending on the source areas. For example, groundwater from the
northwest contains calcium, magnesium, and bicarbonate leached from sedimentary rocks, and
groundwater from the east contains sodium, fluorine, and silicate resulting from contact with volcanic
rocks (Robertson et al. 1974).
Although the natural chemical composition of groundwater beneath the site does not exceed the
Environmental Protection Agency drinking water standards for any component, the natural chemistry
affects the mobility of contaminants introduced into the subsurface from INEL activities. Many
dissolved contaminants adsorb (or attach) to the surface of rocks and minerals in the subsurface,
thereby retarding the movement of contaminants in the aquifer and inhibiting further migration of
contamination. However, many naturally occurring chemicals compete with contaminants for
adsorption sites on the rocks and minerals or react with contaminants to reduce their attraction to rock
and mineral surfaces.
4.8.2.5.2 Groundwater Quality - Previous waste discharges to unlined ponds and deep
wells have introduced radionuclides, nonradioactive metals, inorganic salts, and organic compounds to
the subsurface.
Table 4.8-1 summarizes the highest detected concentrations of contaminants observed
in the aquifer between 1987 and 1992, concentrations near the site boundary, Environmental Protection
Agency maximum contaminant levels, and DOE Derived Concentration Guides. The following
Table 4.8-1. Highest dtected contaminant concentrations in groundwater at the Idaho National Engineering Laboratory (1987 to 1992).
paragraphs discuss each category of contaminants and comparisons of observed concentrations to
maximum contaminant levels.
Radionuclides - In general, radionuclide concentrations in the Snake River Plain Aquifer beneath
the site have decreased since the mid-1980s because of changes in disposal practices, radioactive
decay, adsorption of radionuclides to rocks and minerals, and dilution by natural surface water and
groundwater entering the aquifer (Pittman et al. 1988; Orr and Cecil 1991; Bargelt et al. 1992).
Radionuclides released and observed in the soil and groundwater include tritium, strontium-90,
iodine-129, cobalt-60, cesium-137, plutonium-238, plutonium-239/240, and americium-241 (Golder
Associates 1994). Most of these radionuclides have been observed at the Idaho Chemical Processing
Plant and Test Reactor Area facility areas. However, radionuclides have also been observed in the
Test Area North disposal well.
Concentrations of radionuclides in the aquifer have decreased over time. This decrease is attributed
to reduced discharges, adsorption, radioactive decay, and improved waste management practices. As
of 1992, concentrations of iodine-l29, cobalt-60, tritium, strontiurn-90, and cesium-137 had exceeded
the EPA maximum contaminant levels for radionuclides in drinking water in localized areas inside the
INEL boundary. Currently, there are no individual maximum contaminant levels for plutonium-238,
plutonium-239, plutonium-240, and americium-24 1. However, these radionuclides have not been
detected above the established limits for gross radioactivity or the proposed adjusted gross alpha
activity maximum contaminant level for drinking water (Golder Associates 1994; Mann et al. 1988;
Orr and Cecil 1991).
Extremely low concentrations of iodine- 129 and tritium have migrated outside site boundaries. In
1992, iodine- 129 concentrations were well below the maximum contaminant levels in two wells
approximately 6 and 13 kilometers (4 and 8 miles) south of the site boundary (Mann 1994). Tritium
concentrations were much below maximum contaminant levels just south of the site boundary in 1985.
By 1988 the tritium plume encompassed by the 500 picocurie per liter contour was back inside the site
boundary, and its size has continued to decrease (Pittman et al. 1988; Otr and Cecil 1991; Orr et al.
1991). Cobalt-60, strontium-90, cesium-i 3?, plutonium-238, plutonium-240!241, and americium-241
have not been detected outside the site boundaries.
Nonradioactive Metals - The INEL has released sodium, chromium, lead, and mercury on the
site and into the subsurface through unlined ponds and deep wells. Of these metals, the INEL released
sodium in the greatest quantity from waste treatment processes; however, sodium is not toxic and does
not have an established maximum contaminant level. In 1988 chromium concentrations exceeding the
maximum contaminant level were measured near the Test Reactor Area. Lead and mercury have
occurred at concentrations below the maximum contaminant level near the Idaho Chemical Processing
Plant (Orr and Cecil 1991).
Inorganic Salts - Human activities at the site have released chloride, sulfate, and nitrate into
the subsurface. Although chloride and sulfate releases have occurred, only nitrate has exceeded
maximum contaminant levels (near the Idaho Chemical Processing Plant in 1981). Disposal of nitrates
to the injection well and infiltration ponds at the Idaho Chemical Processing Plant account for the
elevated nitrate levels in the central portion of the site. By 1988 the levels of nitrate decreased to
below the maximum contaminant level. Irrigation in the Mud Lake area might be causing these
contaminants to enter the northeastern portion of the site in concentrations comparable to those in
nearby irrigated areas (Orr et al. 1991; Robertson et al. 1974; Edwards et al. 1990).
Organic Compounds - Concentrations of volatile organic compounds have been detected in
the aquifer beneath the site. However, many of these compounds were detected at amounts below the
detection limit (0.002 milligram per liter), or two parts per billion, which is the lowest concentration at
which a specific analytical method can detect a contaminant. However, concentrations of the
following compounds exceeding the maximum contaminant levels have occurred in and near the Test
Area North disposal well: carbon tetrachloride, chloroform, l,2-cis-dichloroethylene,
1,1 -dichloroethylene, 1 ,2-trans-dichloroethylene, trichloroethylene, tetrachloroethylene, and vinyl
chloride (Leenheer and Bagby 1982; Mann and Knobel 1987; Mann 1990; Liszewski and Mann 1992).
4.8.2.5.3 Perched Water Quality - Wastewater discharges from INEL operations have
infiltrated into the vadose zone and created most of the perched water beneath the site.
Studies have
detected elevated concentrations of the following contaminants in samples: tritium, cesium-l37,
cobalt-60, chromium, and sulfate concentrations in deep perched water near the Test Reactor Area, and
strontium-90 in perched water near the Idaho Chemical Processing Plant and at Test Area North
(Irving 1993; Schafer-Perini 1993). DOE has not yet measured potential concentrations of
contaminants in all INEL perched water bodies. In general, the chemical concentrations, shape, and
size of these bodies have fluctuated over time in response to the volume of water discharged to the
infiltration ponds.
4.8.3 Water Use and Rights
The INEL does not withdraw or use surface water for site operations, nor does it discharge
effluents to natural surface water. However, the three surface-water bodies at or near the site (Big and
Little Lost Rivers and Birch Creek) have the following designated uses: agricultural water supply,
cold-water biota, salmonid spawning, and primary and secondary contact recreation. In addition,
waters in the Big Lost River and Birch Creek have been designated for domestic water supply and as
special resource waters.
Groundwater use on the Snake River Plain includes irrigation, food processing and aquaculture,
and domestic, rural, public, and livestock supply. Water use for the upper Snake River drainage basin
and the Snake River Plain Aquifer was 16.4 billion cubic meters (4.3 trillion gallons) per year in 1985,
which was more than 50 percent of the water used in Idaho and approximately 7 percent of
agricultural withdrawals in the nation. Most of the water withdrawn from the Eastern Snake River
Plain [1.8 billion cubic meters (0.47 trillion gallons) per year] is for agriculture. The aquifer is the
source of all water used at the INEL. Site activities withdraw water at an average rate of 7.4 million
cubic meters (1.9 billion gallons) per year (DOE-ID 1993e). However, the baseline annual withdrawal
rate dropped to 6.5 million cubic meters (1.7 billion gallons) in 1995. The average annual withdrawal
is equal to approximately 0.4 percent of the water consumed from the Eastern Snake River Plain
Aquifer, or 53 percent of the maximum annual yield of a typical irrigation well. Of the quantity of
water pumped from the aquifer, a substantial portion is discharged to the surface or subsurface and
eventually returned to it (DOE-ID l993d,e).
A sole-source aquifer, as designated by the Safe Drinking Water Act (SDWA 1974) is one that
supplies 50 percent of the drinking water consumed in the area overlying the aquifer. Sole-source
aquifer areas have no alternative source or combination of sources that could physically, legally, and
economically supply all those who obtain their drinking water from the aquifer. Because groundwater
supplies 100 percent of the drinking water consumed within the Eastern Snake River Plain (Gaia
Northwest 1988) and an alternative drinking water source or combination of sources is not available,
the Environmental Protection Agency designated the Snake River Plain Aquifer a sole-source aquifer
in 1991 (FR 1991b).
DOE holds a Federal Reserved Water Right for the INEL, which permits a water pumping
capacity of 2.3 cubic meters (80 cubic feet) per second and a maximum water consumption of
43 million cubic meters (11.4 billion gallons) per year for drinking, process water, and noncontact
cooling. Because it is a Federal Water Right, the site's priority on water rights dates back to the
establishment of the INEL.
4.9 Ecological Resources
This section describes the biotic resources - flora, fauna, threatened and endangered species,
and wetlands - on the INEL site, which are typical of the Great Basin and Columbia Plateau.
Because the proposed actions are most likely to affect areas near existing major facilities, this section
emphasizes the biotic resources in those areas. However, because the proposed actions could affect
other resources outside such areas (e.g., more mobile species like pronghorn, Antilocapra americana),
it also describes biotic resources for the entire INEL site.
4.9.1 Flora
Vegetation on the INEL site is primarily of the shrub-steppe type and is a small fraction of the
45,000 square kilometers (111.2 million acres) of this vegetation type in the Intermountain West. The
15 vegetation associations on the INEL site range from primarily shadscale-steppe vegetation at lower
altitudes through sagebrush- and grass-dominated communities to juniper woodlands along the foothills
of the nearby mountains and buttes (Rope et al. 1993; Kramber et al. 1992; Anderson 1991). These
associations can be grouped into six basic types: juniper woodland, grassland, shrub-steppe (which
consists of "sagebrush-steppe" and "salt desert shrubs"), lava, bareground-disturbed, and wetland
vegetation. Shrub-steppe vegetation, which is dominated by big sagebrush (Artemisia tridentata),
saltbush (Atriplex spp.), and rabbitbrush (Chrysothamnus spp.) covers more than 90 percent of the
INEL. Grasses include cheatgrass (Bromus tectorum), Indian ricegrass (Oryzopsis hymenoides),
wheatgrasses, (Agropyron spp.), and squirreltail (Sitanion hysterix). Herbaceous plants include phlox
(Phlox spp.), wild onion (Allium spp.), milkvetch (Astragalus spp.), Russian thistle (Salsola kali), and
various mustards. Work being conducted by Idaho State University will provide additional
information on INEL plant communities and the status of sensitive plant species.
Facility and human-disturbed (grazing not included) areas cover only about 2 percent of the
INEL. Introduced annuals, including Russian thistle and cheatgrass, frequently dominate disturbed
areas. These species usually are less desirable to wildlife as food and cover, and compete with more
desirable perennial native species. These disturbed areas serve as a seed source, increasing the
potential for the establishment of Russian thistle and cheatgrass in surrounding less-disturbed areas.
Vegetation inside facility boundaries is generally disturbed or landscaped. Species richness on the
INEL is comparable to that of like-sized areas with similar terrain in other parts of the Intermountain
West. Plant diversity is typically lower in disturbed and modified areas.
4.9.2 Fauna
The INEL site supports animal communities characteristic of shrub-steppe vegetation and
habitats. More than 270 vertebrate species occur, including 46 mammal, 204 bird, 10 reptile, 2
amphibian, and 9 fish species (Arthur et al. 1984; Reynolds et al. 1986). Common small-mammal
genera include mice (Reithrodontomys spp. and Peromyscus spp.), chipmunks (Tamias spp.),
jackrabbits (Lepus spp.), and cottontails (Sylvilagus spp.).
Songbirds and passerines commonly observed at the INEL include the American robin (Turdus
migratorius), horned lark (Eremophila alpestris), black-billed magpie (Pica pica), sage thrasher
(Oreoscoptes montanus), Brewer's sparrow (Spizella breweri), sage sparrow (S. belli), and western
meadowlark (Sturnella neglecta), while resident upland gamebirds include the sage grouse
(Centrocercus urophasianus), chukar (Alectoris chukar), and grey partridge (Perdix perdix). Common
migratory bird species, which use the INEL for part of the year, include a variety of waterfowl
[e.g., mallard (Anas platyrhynchos), northern pintail (Anas acuta), and Canada goose (Branta
canadensis)] and raptors [e.g., Swainson's hawk (Buteo swainsoni), rough-legged hawk (B. lagopus),
and American kestrel (Falco sparverius)].
The most abundant big-game species that occurs on the INEL is the pronghorn, but mule deer
(Odocoileus hermonius), moose (Alces alces), and elk (Cervus elaphus) are present in small numbers
as transients. Other large mammals observed on the INEL include the coyote (Canis latrans), which is
common across the site, and the badger (Taxidea taxus) and bobcat (Felis rufus), both of which are
present across the site but are much less abundant. Fish, including kokanee salmon (Oncorhynchos
nerka), rainbow trout (Oncorhynchos mykiss), and mountain whitefish (Prosopium williamsoni), occur
on the INEL only when the Big Lost River flows onto the site (as a result of heavy rain- or snowfall
in the mountains to the northwest); they are not full-time residents.
A number of researchers have studied effects of radiation exposure from contaminated areas at
INEL on small mammals and birds, and have concluded that subtle sublethal effects (e.g., reduced
growth rates and life expectancies) can occur in individual animals as a result of radiation exposure.
However, they can attribute no population or community-level impacts to such exposures (Halford and
Markham 1978; Evenson 1981; Arthur et al. 1986; Millard et. al 1990).
The monitoring of radionuclide levels outside the boundaries of the various INEL facilities and
off the INEL site has detected radionuclide concentrations above background levels in individual plants
and animals (Markham 1974; Craig et al. 1979; Markham et al. 1982; Morris 1993), but these limited
data suggest that populations of exposed animals (e.g., mice and rabbits) as well as animals that feed
on these exposed animals (e.g., eagles and hawks) are not at risk.
4.9.3 Threatened, Endangered, and Sensitive Species
State and Federal regulatory agency lists (Lobdell 1992, 1995), the Idaho Department of Fish and
Game Conservation Data Center list, and information from site surveys provided the information to
identify Federal- and state-protected, candidate, and sensitive species that potentially occur on the
INEL. This information identified two Federal endangered (bald eagle, and peregrine falcon) and nine
Federal Category 2 candidate (white-faced ibis, northern goshawk, ferruginous hawk, burrowing owl,
long-eared myotis, small-footed myotis, pygmy rabbit, Townsend's western big-eared bat, and Idaho
pointheaded grasshopper) species as animals that potentially occur on the INEL site (Table 4.9-1).
Five animal species listed by the state as Species of Special Concern occur on the site. No frequent
observations of the Federal- or state-listed animal species have occurred near any of the facilities
where proposed actions would occur. This analysis did not identify any Federal- or state-listed plant
species as potentially occurring on the INEL site. Eight plant species identified by other Federal
agencies and the Idaho Native Plant Society as sensitive, rare, or unique occur on the site (Chowlewa
and Henderson 1984).
4.9.4 Wetlands
The U.S. Fish and Wildlife Service National Wetlands Inventory has identified more than 130
areas inside the boundaries of the INEL that might possess some wetlands characteristics. Surveys
conducted in the fall of 1992 indicate that these possible wetlands cover about 1.4 percent (33 square
kilometers or 8,206 acres) of the INEL site (Hampton et al. 1993). Approximately 70 percent of these
possible wetlands areas occur near the Big Lost River and its spreading areas and playas, near the
Birch Creek Playa, and in an area north of and in the general vicinity of Argonne National
Laboratory-West. Limited riparian (riverbank) communities with mature trees along the Big Lost
River (Reynolds 1993) reflect the intermittent flow in the river (1986 and 1993 were the last two years
with flow reported on the site). The remainder of the possible wetlands are scattered throughout the
INEL site. In 1994, INEL began evaluating these potential wetlands to determine if they meet the
Corps of Engineers definition of jurisdictional wetlands (COE 1987). Approximately 20 wetlands are
near facilities and are mostly manmade (e.g., industrial waste and sewage treatment ponds, borrow
pits, and gravel pits).
Table 4.9-1. Threatened and endangered species, special species of concern, and sensitive species that may be found on the INEL.
Name Statusa Comments
BIRDS Northern goshawk (Accipiter gentilis) C2, SSC, FS, BLM The ferruginous hawk nests on and migrates through the INEL. This
Burrowing owl (Athene cunicularia) C2, BLM species is found throughout the INEL but is observed more frequently
Ferruginous hawk (Buteo regalis) C2, SSC, BLM in juniper woodlands. The peregrine falcon has been observed rarely
Swainson's hawk (Buteo swainsoni) BLM in winter, but has not been observed during other seasons. The last
Great egret (Casmerodius albus) SSC sighting was in 1993 (Morris 1993). It is not known to nest on the
Merlin (Falco columbarius) SSC, BLM INEL and is not commonly observed near facilities (Reynolds 1993a).
Peregrine falcon (Falco peregrinus) E The bald eagle is a winter resident and is locally common in the far
Gyrfalcon (Falco rusticolus) BLM north end and on the western edge of the INEL near Howe (Reynolds
Common loon (Gavia immer) SSC, FS 1993a). It is not known to nest on the INEL and is not commonly
Bald eagle (Haliaeetus leucocephalus) E observed near facilities (Reynolds 1993). The white-faced ibis, which
Long-billed curlew (Numenius americanus) SPS, BLM uses aquatic and riparian habitats, is an uncommon migrant at the
American white pelican (Pelecanus erythrorhynchos) SSC INEL. The long-billed curlew is known to nest on the north end of
White-faced ibis (Plegadis chihi) C2 the INEL near agricultural lands. The northern goshawk is a casual
migrant through the INEL.
MAMMALS Merriam's shrew (Sorex merriami) SPS The pygmy rabbit is common on the INEL, but its distribution is
Pygmy rabbit (Brachylagus (Sylvilagus) idahoensis) C2, BLM, SSC patchy (Reynolds et al. 1986). Roosts and hibernation caves for
California myotis (Myotis californicus) SSC Townsend's western big-eared bat occur on the INEL. All are over
Fringed myotis (Myotis thysanodes) SSC 7 kilometers (3 miles) from facilities. Brood caves might exist on the
Western pipistrelle (Pipistrellus hesperus) SSC, BLM site but have not been located.
Townsend's western big-eared bat (Plecotus townsendii) C2, SSC, FS, BLM
Long-eared myotis (Myotis evotis) C2
Small-footed myotis (Myotis subulatus) CS
PLANTS Lemhi milkvetch (Astragalus aquilonius) BLM, FS, INPS The 8 plant species identified as sensitive, rare, or unique that are
Painted milkvetch (Astragalus ceramicus var. apus) 3c, INPS-M known to occur on the INEL occur primarily at a distance from INEL
Winged-seed evening primrose (Camissonia pterosperma) BLM, INPS-S facilities and are uncommon on the INEL because they require unique
Nipple cactus (Coryphantha missouriensis) INPS-M microhabitat conditions.
Spreading gilia (Ipomopsis (Gilia) polycladon) BLM, INPS-2
King's bladderpod (Lesquerella kingii var. cobrensis) INPS-M
Tree-like oxytheca (Oxytheca dendroidea) INPS-S
Sepal-tooth dodder (Cuscuta denticulata) INPS-1
INSECTS Idaho pointheaded grasshopper (Acrolophitus pulchellus) C2, BLM Occurs just north of the INEL.
a. Key: C2 = Federal Category 2 species. BLM = Bureau of Land Management monitored. INPS-S = Idaho Native Plant Society sensitive.
3c = No longer considered for Federal listing. FS = U.S. Forest Service monitored. INPS-M = Idaho Native Plant.
E = Federal and state endangered species. INEL = Idaho National Engineering Laboratory. INPS-1 = Idaho Native Plant Society State Priority 1.
SSC= State species of special concern. SPS = State protected species. INPS-2 = Idaho Native Plant Society State Priority 2.
4.10 Noise
The major noise sources at the INEL occur primarily in developed operational areas. These
sources include facilities; equipment and machines (e.g., cooling towers, transformers, engines, pumps,
boilers, steam vents, paging systems, construction equipment, and materials-handling equipment);
aircraft; and bus, car, truck, and railroad traffic. At the INEL boundary, which is more than
3 kilometers (2 miles) from any facility, noise from most sources is barely distinguishable from
background noise levels. Some disturbance of wildlife activities could occur at the INEL as a result of
noise from operational and construction activities. The State of Idaho and the counties in which the
INEL is located have not established any regulations that specify acceptable community noise levels,
with the exception of prohibitions on nuisance noise.
Existing INEL-related noises of public significance are from the transportation of people and
materials to and from the site and in-town facilities via buses, trucks, private vehicles, helicopters, and
freight trains. During the normal workweek, most of the 4,000 to 5,000 employees who work on the
site (as opposed to those working in Idaho Falls) travel daily by buses from surrounding communities
(see Section 4.3). In addition, 300 to 500 private vehicles travel to the INEL site from surrounding
communities each day (see Section 4.11). Noise measurements along U.S. Highway 20 about
15 meters (50 feet) from the roadway indicate that the sound level from traffic ranges from 64 to 86
decibels, A-weighted (dBA) (Abbott et al. 1990), and that the primary source is buses (71 to 81 dBA).
While few people reside within 15 meters (50 feet) of the roadway, the results indicate that INEL
traffic noise might be objectionable to members of the public residing near principal highways or busy
bus routes. The acoustic environment along the INEL site boundary in rural areas and at nearby areas
away from traffic noise is typical of a rural location, with the day-night sound level (DNL) in the
range of 35 to 50 dBA (EPA 1974).
Public exposure to aircraft noise is due in part to INEL-related activities. Air cargo and business
travel of INEL personnel via commercial air transport is a significant fraction of all such travel in and
out of regional airports. Onsite INEL security patrol and surveillance flights do not adversely affect
individuals off the site because of the INEL's remoteness. For INEL helicopter flights that originate
or terminate in Idaho Falls, members of the public are exposed to the unique noises produced by these
aircraft. Because the number of flights per day is limited and most flights occur during nonsleeping
hours, public exposure to aircraft nuisance noise is not great.
Normally only one train per day serves the INEL, via the Scoville spur. Noise sources related to
rail transport include those from diesel engines, wheel-track contact, and whistle warnings at rail
crossings. Even with only one or two exposures to these sources per day, individuals residing near the
railroad tracks might find the noises mildly objectionable.
4.11 Traffic and Transportation
Roads are the primary access to and from the INEL site. Commercial shipments are transported
via truck and plane, some bulk materials are transported via rail, and waste is transported by road and
rail. This section discusses the existing traffic volumes, transportation routes, transportation accidents,
and waste and materials transportation, including baseline radiological exposures from waste and
materials transportation. This section summarizes the information in Lehto (1993).
4.11.1 Roadways
4.11.1.1 Infrastructure Regional and Site Systems. Figure 4.11 - 1 shows the existing
regional highway system. Two interstate highways serve the regional area. Interstate 15 (1-15), a
north-south route that connects several cities along the Snake River, is approximately 40 kilometers
(25 miles) east of the INEL site. 1-86 intersects 1-15 approximately 64 kilometers (40 miles) south of
the INEL site, and provides a primary linkage from I-li to points west. 1-15 and US 91 are the
primary access routes to the Shoshone-Bannock reservation. US 20 and US 26 are the main access
routes to the southern portion of the INEL site. Idaho State Routes 22, 28, and 33 pass through the
northern portion of the INEL; State Route 33 provides access to the northern INEL site facilities.
Table 4.11-1 lists the baseline (1991) traffic for several of these access routes. The level of service of
these segments is currently designated "free flow," which is defined as "operation of vehicles is
virtually unaffected by the presence of other vehicles."
The INEL has developed an onsite road system of approximately 140 kilometers (87 miles) of
paved surface, including about 29 kilometers (18 miles) of service roads that are closed to the public.
Most of the roads are adequate for the current level of normal transportation activity and could handle
some increased traffic volume. DOE plans to reconstruct several deteriorating INEL roads built in the
1950s that have been and will continue to be used to transport heavier-than-normal loads.
4.11.1.2 Infrastructure Idaho Falls. Approximately 4,000 DOE and contractor personnel
administer and support INEL work at offices in Idaho Falls. DOE shuttle vans provide hourly
transport between in-town facilities. One of the busiest intersections is Science Center Drive and
Fremont Avenue, which serves Willow Creek Building, Engineering Research Office Building, INEL
Figure 4.11-1. Transportation routes in the vicinity of the INEL. (not available in electronic copy).
Table 4.11-1. Baseline traffic for selected highway segments. Electronic Technology Center, and DOE Office Buildings. This intersection is congested during peak
weekday hours, but it is designed for the current traffic.
4.11.1.3 Transit Modes. Four major modes of transit use the regional highways, community
streets, and INEL site roads to transport people and commodities: DOE buses and shunle vans, DOE
motor pool vehicles, commercial trucks, and personal vehicles. Table 4.11-2 summarizes the baseline
miles for INEL-related traffic.
Table 4.11-2. Baseline annual vehicle miles traveled for Idaho National Engineering Laboratory- related traffic. a
4.11.2 Railroads
Figure 4.11-1 shows the Union Pacific Railroad lines in southeastern Idaho. Idaho Falls receives
railroad freight service from Butte, Montana, to the north, and from Pocatello and Salt Lake City to
the south, The Union Pacific Railroad's Blackfoot-to-Arco branch, which crosses the southern portion
of the INEL, provides rail service to the site for the shipment of spent nuclear fuel and other waste,
bulk commodities, and radioactive materials. This branch connects with a DOE-owned spur line at
Scoville Siding, then links with developed INEL areas. Table 4.11-3 lists rail shipments for Fiscal
Years 1988 through 1992.
Table 4.11-3. Loaded rail shipments to and from the Idaho National Engineering Laboratory (1988-1992)a.
4.11.3 Airports and Air Traffic
Commercial airlines provide Idaho Falls with jet aircraft passenger and cargo service, as well as
commuter service to both the Idaho Falls and Pocatello airports. In addition, local charter service is
available in Idaho Falls, and private aircraft use the major airport and many other fields in the area.
Total landings at the Idaho Falls airport for 1991 and 1992 were 5,367 and 5,598, respectively. The
Idaho Falls and Pocatello airports collectively record nearly 7,500 landings annually.
Non-DOE air traffic over the INEL site is limited to altitudes greater than 305 meters
(1,000 feet) over buildings and populated areas, and non-DOE aircraft are not permitted to use the site.
The primary air traffic at the INEL site is DOE helicopters, which are used for security and emergency
purposes. These helicopters have specific operations stations and duties.
4.11.4 Accidents
From 1987 through 1992, the average motor vehicle accident rate was 0.94 accident per million
kilometers (1.5 accidents per million miles) for INEL vehicles, which compares with an accident rate
of 1.5 accidents per million kilometers (2.4 accidents per million miles) for all DOE complex vehicles
and 8 accidents per million kilometers (12.8 accidents per million miles) nationwide for all motor
vehicles (Lehto 1993). There are no recorded rail or air accidents associated with the INEL and, to
date, no fatal air traffic accidents have involved flights through either the Idaho Falls or Pocatello
airports.
4.11.5 Transportation of Waste, Materials, and Spent Nuclear Fuel
Hazardous, radioactive, industrial commercial, and recyclable wastes are transported oYl the INEL
site. Federal and State regulations and requirements govern the transportation of hazardous and
radioactive materials (Lehto 1993). Hazardous materials include commercial chemical products and
hazardous wastes that are nonradioactive; they are regulated and controlled based on their chemical
toxicity. Onsite spent nuclear fuel comes from Argonne National Laboratory - West, the Naval
Reactors Facility, and the Advanced Test Reactor; it is transported by truck to various onsite storage
and research and development facilities.
This assessment used six years of data (1987 through 1992) to establish a baseline of radiological
doses from incident-free, onsite total nonnaval spent nuclear fuel transportation at the INEL.
Table 4.11-4 lists the results in terms of cumulative doses (1995-2035) and health effects. These doses
do not include onsite naval shipments, which are assessed in Attachment A to Appendix D of
Volume 1 of this ElS. The baseline includes no offsite shipments, which are addressed in
Appendixes D and I.
Table 4.11-4. Cumulative dose and cancer fatalities from incident-free onsite shipment of nonnaval spent nuclear fuel at the Idaho National Engineering Laboratory for 1995 through 2035. (a,b)
4.12 Occupational and Public Health and Safety
4.12.1 Radiological Health and Safety
DOE Order 5480.11, "Radiation Protection for Occupational Workers" (DOE 1992b), limits the
radiation dose that INEL workers can receive to 5 rem per year; administrative controls further limit a
worker dose to 2 rem per year, except under unusual circumstances. In addition, DOE has established
a comprehensive program, known as ALARA (As Low As Reasonably Achievable), to ensure the
reduction of occupational doses to the extent practicable.
The largest fraction of the occupational dose received by INEL workers is from external
radiation. Internal radiation doses constitute a small fraction of the occupational dose. Personnel who
could receive annual external radiation exposures with measured doses greater than 0.1 rem receive a
thermoluminescent dosimeter that they must wear at all times during work on the site. DOE used
recorded doses for 1987 to 1991 as a baseline for routine site operations for this EIS. During this
period, the INEL monitored about 6,000 workers annually for radiation exposure. About 32 percent of
those individuals received measurable radiation doses. Monitoring reports indicate that, from 1987 to
1991, 20 individuals (most of whom were maintenance and construction workers employed by
M-K Ferguson at the Idaho Chemical Processing Plant) received annual doses larger than 2 rem
(4 individuals in 1987, 1 in 1989, and 15 in 1990).
From 1987 to 1991, the average occupational dose to individuals who had received measurable
doses was 0.156 rem per year, resulting in an average collective dose (the number of monitored
workers receiving measurable doses was about 32 percent or 1,920) of about 300 person-rem. The
resulting number of expected excess latent cancer fatalities would be less than 1 for each year of
operation.
This analysis based the doses to the maximally exposed individual and offsite population on
baseline radioactive concentrations associated with normal operations. The baseline dose to the
maximally exposed individual is 5.6 y 10-2 millirem, which corresponds to a latent fatal cancer
probability of 2.8 y 10-8. The baseline population dose is 7.0 y 10-2 person-rem which, corresponds to
a latent fatal cancer incidence of less than 1 (4 y 10-5) annually and less than 1 (1 y 10-3) over
40 years.
4.12.2 Nonradiological Exposure and Health Effects
DOE used the air quality data in Table 4.7-2 to evaluate health impacts associated with potential
exposure to two compound classes: criteria pollutant and toxic. This analysis has based health effects
on air emissions only, and not water pathways, because none of the alternatives would involve the
discharge of pollutants to surface waters or the subsurface. Table 4.7-2 lists 5 criteria pollutant and
26 toxic compounds. The classification of two of the toxic compounds (benzene and formaldehyde) as
carcinogens was consistent with EPA designations published in the Integrated Risk Information System
(IRIS) data base (DOE 1991b). However, this data base does not include sufficient data to perform a
quantitative inhalation cancer risk assessment.
To obtain a hazard index, this analysis evaluated toxic and criteria pollutant compound health
effects by adding hazard quotients for each compound. The EPA Risk Assessment Guidance for
Superfund (EPA 1989) describes this approach. The hazard quotient is the ratio of compound
concentration or dose to a Reference Concentration (RfC) or Dose (RfD). For compounds without
listed Reference Concentration or Dose values, the analysis used appropriate State of Idaho standards.
The use of the noncancer hazard index assumes a level of exposure (standard) below which adverse
health effects would be unlikely. The hazard index is not a statistical probability; therefore, it cannot
be interpreted as such.
This analysis based toxic and criteria pollutant compound hazard index values for the maximally
exposed individual on the maximum concentrations for the compounds at the INEL site boundary,
public access roads inside the INEL site boundary, and the Craters of the Moon Wilderness Area.
Because the hazard index for criteria pollutants is less than 1, no adverse health effects would be likely
from routine operations for either workers or the maximally exposed individual. Because the hazard
index for toxic pollutants exceeds 1, the potential for carcinogenic health risks could exist. However,
varying spacial and temporal distributions of the concentrations of individual air pollutants make it
unlikely that any individual would be exposed to all the pollutants all the time. Since individual
hazard indices for the toxic compounds are less than 1, adverse health effects are not expected.
4.12.3 Occupational Health and Safety
Total injury and illness incidence rates at the INEL varied from an annual average of 1.8 to
4.9 per 200,000 work hours from 1987 to 1991. During this time, total lost workday cases ranged
from a low of 1 per 200,000 work hours in 1988 and 1989 to a high of 2.6 per 200,000 work hours in
1991. The rates appear higher for 1991 because of a 1990 change in reporting requirements for
injuries and illnesses. INEL rates for 1987 to 1989 are below overall DOE rates (2.9 total injury and
illness incidence and 1.4 total lost workday cases per 200,000 work hours) and Bureau of Labor
Statistics rates (8.5 total injury and illness incidence and 4.0 total lost workday cases per 200,000 work
hours). For 1990 and 1991, INEL rates are slightly above overall DOE rates, but below Bureau of
Labor Statistics rate.
There were 1,337 total recordable injury and illness cases at the INEL from 1987 to 1991, for an
average of 8,385 employees working 79,654,000 hours. Of these cases, 114 (8.5 percent) were
occupational illnesses, of which 48 percent were repeated trauma disorders and 30 percent were
classified as skin diseases or disorders. One fatality occurred at the INEL between 1987 and 1991
when an employee was struck and killed by a forklift.
4.13 Idaho National Engineering Laboratory Services
This section discusses water, electricity, fuel capacities and consumption, wastewater disposal,
and security and emergency protection at INEL facilities.
4.13.1 Water Consumption
A system of about 30 wells, with pumps and storage tanks, provides the water supply for the
INEL site. Because of the distance between site facility areas, the water supply system for each
facility is independent. The site uses no natural surface water. The City of Idaho Falls water supply
system, which includes about 16 wells, provides water to DOE and contractor facilities in the city.
A Water Rights Agreement between DOE and the State of Idaho regulates groundwater use at
the INEL site. Under this agreement, INEL has claim to 2,300 liters per second (36,000 gallons per
minute) of groundwater, not to exceed 43 billion liters (11 billion gallons) per year (Teel 1993). DOE
has not measured the total pumping rate from the aquifer, which would depend on the number of
pumps operating. There is a slight possibility that the site could exceed the regulated pumping rate for
very short periods, such as during recovery from an extended power outage when many pumps would
run to refill depleted storage tanks.
The average INEL site water consumption from 1987 through 1991 was 7.4 billion liters
(1.9 billion gallons) per year, based on the cumulative volumes of water withdrawn from the wells
(Teel 1993). The projected baseline usage for 1995 will be about 6.5 billion liters (1.7 billion
gallons). The estimated average water consumption of Idaho Falls facilities is 300 million liters
(80 million gallons) per year.
4.13.2 Electricity Consumption
The Antelope substation supplies commercial electric power to the INEL site through two feeders
to the Federally owned Scoville substation. The Scoville substation supplies electric power directly to
the INEL electric power distribution system (Teel 1993). The contract with Idaho Power Company to
supply electric power to the INEL site provides "up to 45,000 kilowatts monthly" at 13.8 kilovolts
(IPC/DOE 1986). Hydroelectric generators along the Snake River in southern Idaho and the Bridger
and Valmy coal-fired thermal electric generation plants in southwestern Wyoming and northern
Nevada, respectively, generate the electric power supplied by Idaho Power. The Experimental Breeder
Reactor-II can also provide approximately 12 to 15 megavolt-amperes of capacity for the electric
power loop (Teel 1993).
The rated capacity of the INEL site power transmission loop line is 124 megavolt-amperes. The
peak demand on the system from 1990 through 1993 was about 40 megavolt-amperes, and the average
usage was slightly less than 217,000 megawatt-hours per year (Teel 1993). This usage rate should
decrease by about 4 percent by 1995.
The INEL facilities in Idaho Falls receive electric power from the City of Idaho Falls, which
operates four hydroelectric power generation plants on the Snake River along with substation and
distribution facilities. The Bonneville Power Administration, which operates hydroelectric plants on
the Columbia River system, supplies supplemental power to the City of Idaho Falls. In 1993, Idaho
Falls facilities used 31,500 megawatt-hours of electricity (Teel 1993).
4.13.3 Fuel Consumption
Fuels consumed at the INEL site include several liquid petroleum fuels, coal, and propane. All
fuels are transported to the site for storage and use. Natural gas is the only reported fuel consumed at
the INEL Idaho Falls facilities; the Intermountain Gas Company provides this fuel through a system of
underground lines (Teel 1993).
The average annual fuel consumption at the INEL site from 1990 through 1993 was as follows:
fuel oil, 10,578,000 liters (2,795,000 gallons); diesel fuel, 5,690,000 liters (1,500,000 gallons); and
propane gas, 568,000 liters (150,000 gallons). The INEL also uses about 8,200 metric tons
(9,000 tons) of coal. Fuel storage is provided at each facility and inventories are restocked as
necessary. No fossil fuel shortage has ever occurred at the INEL site (Teel 1993).
4.13.4 Wastewater Disposal
Sanitary wastewater systems at the smaller onsite facility areas consist primarily of septic tanks
and drain fields. The larger areas, such as Central Facilities Area, Idaho Chemical Processing Plant,
and Test Reactor Area, have wastewater treatment facilities. The City of Idaho Falls wastewater
treatment system serves the Idaho Falls facilities (Teel 1993).
The average annual wastewater discharge volume at the INEL site from 1989 through 1991 was
537 million liters (142 million gallons). The wastewater from DOE and contractor-operated facilities
in Idaho Falls is not metered but is estimated to be 300 million liters (80 million gallons) per year.
The primary causes of the difference between water pumped and estimated wastewater discharge are
evaporation from ponds and cooling towers, irrigation of landscaped areas, and discharge of unmetered
wastewater (Teel 1993). Some industrial wastewater, such as steam condensate, is also discharged to
evaporation ponds and injection wells.
4.13.5 Security and Emergency Protection
This section describes the fire protection and prevention, security, and emergency preparedness
resources for the INEL site and the surrounding areas. This discussion includes the INEL Fire
Department, DOE and INEL Emergency Preparedness, and DOE and INEL Security. DOE established
an Emergency Management System that incorporates all applicable requirements for emergency
planning, preparedness, and response at the INEL. Each INEL facility must prepare an Emergency
Plan that contains detailed contingency plans and emergency procedures.
4.13.5.1 DOE Fire Department. The contractor-operated Fire Department staffs and operates
three fire stations on the INEL that support the entire site. Each station has the equipment and
expertise to respond to explosions, fires, spills, and medical emergencies. These stations are on the
north end at Test Area North, at Argonne National Laboratory-West, and at the Central Facilities Area.
Each station has a minimum of one engine company capable of supporting any fire emergency in its
assigned area. The Fire Department has a staff of 44 firefighters and 11 support personnel and
operates with a minimum critical staff of 7 firefighters at any time. In addition to providing
firefighting services, the Fire Department provides the INEL ambulance, emergency medical technician
(EMT), and hazardous material response services. The Fire Department has mutual aid agreements
with other firefighting organizations, such as the Bureau of Land Management and the Cities of Idaho
Falls, Blackfoot, and Arco. Through these agreements, the Idaho Falls Fire Department serves DOE
facilities in the City of Idaho Falls.
4.13.5.2 DOE and INEL Emergency Preparedness. Each DOE INEL contractor
administers and staffs its own emergency preparedness program under the direction and supervision of
DOE. All contractor programs for emergency control and response are compatible. The Warning
Communication Center is in the DOE Headquarters building and staffed by the INEL prime contractor
with DOE oversight; it is the communication and overall control center for support to onscene
commanders in charge of an emergency response. The DOE emergency preparedness system includes
mutual aid agreements with all regional county and major city fire departments, police, and medical
facilities. Through the agreements, the Idaho Falls emergency preparedness organizations serve DOE
facilities in the City of Idaho Falls.
4.13.5.3 DOE and INEL Security. DOE has oversight responsibility for safeguards and
security at the INEL. The security program has three categories: security operations, personnel
security, and safeguards. The security operations division provides asset protection (classified matter,
special nuclear material, facilities, and personnel) and technical security (computer and information).
Under this category, DOE administers the INEL protective force, which is supplied by contract. The
personnel security staff processes personnel security clearances. The safeguards department is
responsible for the management and accountability of special nuclear materials. The INEL protective
force, consisting of 200 armed guards and 350 support personnel, provides the onsite personnel who
administer the programs. Each INEL contractor has a safeguards and security staff, divided in a
similar manner, to manage the security associated with its facilities. Contractor safeguards and
security staffs range from about 5 to 60 persons, depending on the size and complexity of the
associated facilities. Each staff works with the INEL protective forces.
4.14 Materials and Waste Management
This section summarizes the management of materials and wastes (high-level, transuranic, mixed
low-level, low-level, hazardous, industrial and commercial solid wastes and hazardous materials) at the
INEL and Idaho Falls facilities, and presents an overview of the current status of the various waste
types generated, stored, and disposed at the INEL.
The total amount of waste generated and disposed has been reduced through waste minimization
and treatment. The INEL attains waste minimization by reducing or eliminating waste generation, by
recycling, and by reducing the volume, toxicity, or mobility of waste before storage or disposal. In
addition, the site has achieved volume reduction of radioactive wastes through more intensive
surveying, waste segregation, and use of administrative and engineering controls.
The quantitative data presented in this section are from Volume 2 of this EIS, unless otherwise
noted.
4.14.1 High-Level Waste
At present, about 11,900 cubic meters (4,970 cubic yards calcine solid and 2,140,000 gallons
liquid) of high-level waste are in storage at the INEL Idaho Chemical Processing Plant (see Figure 2-1
for locations of major waste management facilities). This facility blends liquid waste, consisting of
aluminum and zirconium wastes from past spent nuclear fuel reprocessing, and sodium-bearing wastes,
and processes them through calcination to produce a granular calcine solid. Because of the
termination of reprocessing, the site no longer generates liquid high-level waste, with the exception of
high-level waste residues. Liquid high-level wastes generated by prior reprocessing activities are
solidified at the site. At present, the site generates liquid waste that is not directly the result of
reprocessing. The site manages this liquid as high-level waste. The site will calcine the liquid
high-level waste that does not contain sodium, and as much sodium-bearing high-level waste as
practicable by January 1, 1998, in accordance with the Amended Order Modifying Order of June 28,
1993, United States District Court for the District of Idaho, December 22, 1993. The projected 1995
baseline for high-level waste generation is 750 cubic meters (980 cubic yards) annually (EG&G 1993).
4.14.2 Transuranic Waste
About 65,000 cubic meters (85,000 cubic yards) of transuranic and alpha-contaminated low-level
wastes are retrievably stored and 62,000 cubic meters (81,000 cubic yards) of transuranic waste
(Morton and Hendrickson 1995) have been buried at the Radioactive Waste Management Complex at
the INEL. At present, no facilities can dispose of transuranic waste; however, DOE ultimately intends
to retrieve, repackage, certify, and ship stored transuranic wastes at the INEL to a potential Federal
repository for final disposition. DOE has not determined the disposition of alpha-contaminated low-
level waste and buried waste. Since the October 1988 ban by the State of Idaho prohibiting shipments
of transuranic waste to the INEL, DOE has shipped only minor amounts of transuranic waste
generated on the site to the INEL Radioactive Waste Management Complex for interim storage. At
present, there are no treatment facilities for transuranic wastes at the INEL. The projected 1995
baseline for transuranic waste generation is 6 cubic meters (8 cubic yards) annually (EG&G 1993).
4.14.3 Mixed Low-Level Waste
At present, DOE accepts only mixed low-level waste generated at the INEL for treatment and
disposal at the INEL. DOE stores mixed low-level waste generated at the INEL at interim storage
facilities until treatment systems become available or operational. A total of 1,800 cubic meters
(2,400 cubic yards) of mixed low-level waste interim storage capacity is available at the INEL.
Current mixed low-level waste interim storage is approximately 1,100 cubic meters (1,400 cubic
yards). Treatment technologies exist for much of the mixed low-level waste generated at the INEL,
and waste minimization eliminates potential sources of mixed low-level waste before generation. The
projected 1995 baseline for mixed low-level waste is 525 cubic meters (687 cubic yards) annually
(EG&G 1993).
4.14.4 Low-Level Waste
Through 1991, DOE disposed of 145,000 cubic meters (190,000 cubic yards) of low-level waste
at the Radioactive Waste Management Complex. In 1991, the total available low-level waste disposal
capacity at the complex was 37,000 cubic meters (48,000 cubic yards). DOE has curtailed low-level
waste treatment since 1991 while waiting for updated safety documentation and an environmental
impact assessment for the Waste Experimental Reduction Facility. The INEL stores low-level waste
awaiting treatment on asphalt or concrete pads at the Waste Experimental Reduction Facility and in
radioactive waste storage containers at the generating facilities. The projected 1995 baseline for low-
level waste generation is 4,270 cubic meters (5,585 cubic yards) annually (EG&G 1993).
4.14.5 Hazardous Waste
DOE collects hazardous waste generated at the INEL and stores it temporarily at the Hazardous
Waste Storage Facility before shipping it off the site. The Hazardous Waste Storage Facility has
adequate storage capacity [approximately 64 cubic meters (84 cubic yards)] to manage the quantities of
hazardous waste generated at the INEL. The site recycles, reuses, or reprocesses such waste if
possible, and might replace some hazardous substances with nonhazardous substances.
4.14.6 Industrial/Commercial Solid Waste
DOE disposes of the industrial and commercial solid waste generated at the site in the INEL
Landfill Complex at the Central Facilities Area. The Landfill Complex has approximately
910,000 square meters (225 acres) of land available for solid waste disposal, including the remaining
area at Landfill III, which is currently in use. The estimated capacity of the INEL Landfill Complex
will be sufficient to dispose of INEL waste for 30 to 50 years; however, capacity of the current
excavations will be filled by 1998. DOE has proposed expanding the excavation. Volume 2 of this
EIS describes the landfill expansion project. The industrial and commercial solid waste landfill
currently in use is in a 48,000-square-meter (12-acre) gravel pit area north of Disposal Area II. DOE
does not expect to store solid waste intended for disposal. Waste segregation occurs at each INEL
facility so recyclable materials do not enter the solid waste stream. The average annual volume of
waste disposed at the Central Facilities Area landfill from 1988 through 1992 was approximately
52,000 cubic meters (68,000 cubic yards) (also the projected 1995 baseline) (EG&G 1993).
4.14.7 Hazardous Materials
The INEL 1993 chemical inventory lists 774 hazardous chemicals. The number and the total
weight of hazardous chemicals used on the site and at individual facilities change daily in response to
use. The annual Superfund Amendments and Reauthorization Act reports for the INEL facilities
include year-to-year inventories.
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 U.S. Department of Energy (DOE) used the
environmental consequence analyses of nonnaval spent nuclear fuel management from Volume 2 as
input for this chapter; however, DOE made necessary adjustments to accommodate the differences
between Volume 1 and Volume 2 alternatives. In addition, DOE adjusted the 10-year planning
horizon for Volume 2 alternatives to 40 years for Volume 1.
As described in Chapter 1, this chapter analyzes only nonnaval DOE actions; however,
Section 5.16, "Cumulative Impacts and Impacts from Connected or Similar Actions," includes impacts
from the Naval Nuclear Propulsion Program and nonnaval DOE impacts that are cumulative. The
Appendix B restriction of analysis to nonnaval actions results in Alternative 2 (options 2a, 2b, and 2c)
becoming a single alternative.
Chapter 5 addresses potential impacts from construction and normal operations for each element
of the affected environment described in Chapter 4. In addition, it provides potential consequences
from accidents and several types of summary information. In cases where the consequence analysis
does not result in a distinction among the alternatives, this chapter describes the consequences without
division by alternative to avoid needless repetition. Tables 3-4 through 3-6 in Section 3.2 summarize
and compare the potential impacts associated with each alternative.
5.2 Land Use
Alternatives 1, 2, 4b(2), and 5a [No Action, Decentralization, Regionalization by Geography
(Elsewhere), and Centralization at other DOE sites] would have the least impact on land use, affecting
0.8 acre (0.003 square kilometer); Alternatives 4b(1) [Regionalization by Geography (INEL)] and
5b (Centralization at the INEL) would result in the greatest changes, impacting nearly 31 acres
(0.12 square kilometer).
Overall environmental impacts on land use by any of the alternatives would be small because
DOE would build new facilities in developed areas that it has already dedicated to industrial use and
that previous activities have disturbed. Under all the alternatives, proposed activities would be
consistent with the existing land use plans discussed in Section 4.2 and would be similar to uses in
existing developed areas on the site. None of the proposed activities would involve land outside the
INEL boundaries, and no effects on surrounding land uses or local land use plans should occur.
No onsite land use restrictions due to Native American treaty rights would exist for any of the
alternatives described in this EIS. Potential impacts on Native American and other cultural resources
are discussed in Section 5.4 (Cultural Resources) and in Appendix L (Environmental Justice).
5.3 Socioeconomics
This section describes the potential effects of the spent nuclear fuel alternatives on the
socioeconomic resources of the region of influence described in Section 4.3. Tables 5.3-1 and 5.3-2
list proposed changes in the INEL-related workforce and population. Figure 5.3-1 shows these
proposed changes.
5.3.1 Methodology
This section addresses socioeconomic impacts in terms of both direct and secondary employment
and population effects. Direct effects are changes in INEL employment that DOE expects to occur
under each alternative and include construction and operations phase impacts. Secondary effects
include indirect and induced impacts. Indirect effects are impacts to regional businesses and
employment resulting from changes in DOE regional purchases or nonpayroll expenditures. Induced
effects are impacts to regional businesses and employment that result from changes in payroll spending
by affected INEL employees. The total economic impact to the region is the sum of direct and
secondary effects.
The bases for the estimated direct impacts in this section are project summary data that DOE
developed in cooperation with INEL contractors. Employment impacts represent actual changes in
INEL staffing; they do not include changes in staffing due to a reassignment of the existing INEL
workforce. The projected decline in baseline INEL activity is not part of any alternative and therefore,
a comprehensive analysis of potential impacts was not included. Projected declines in baseline site
employment are presented in Figure 5.3-1 in order to provide the reader with a framework for
evaluating potential employment and population impacts. This assessment used RIMS II to estimate
total employment impacts with multipliers that the U.S. Bureau of Economic Analysis developed
specifically for the INEL region of influence. A comprehensive discussion of the methodology is
provided in Appendix F-1 of Volume 2. Cumulative impacts on socioeconomic resources in the
region are discussed in Section 5.16.
Table 5.3-1. Estimated changes in employment and population for Alternatives 3, 4a, 4b(1) and 5b,
1995 - 2004.
Factor 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004
Direct employment 0 0 0 0 250 250 375 375 375 375
Secondary 0 0 0 0 352 352 528 528 528 528
employment
Total employment 0 0 0 0 602 602 903 903 903 903
change
Change in ROIb 0.0 0.0 0.0 0.0 0.5 0.5 0.8 0.8 0.8 0.7
labor force (%)
Change in ROI 0.0 0.0 0.0 0.0 0.6 0.6 0.8 0.8 0.8 0.8
employment (%)
Population change 0 0 0 0 2,027 2,027 3,040 3,040 3,040 3,040
Change in ROI 0.0 0.0 0.0 0.0 0.8 0.8 1.1 1.1 1.1 1.1
population (%)
a. Sources: Johnson (1995); USBEA (1993); USBC (1992).
b. ROI = region of influence.
Table 5.3-2. Estimated changes in employment and population for Alternatives 4b(2) and 5a,
1995 - 2004.
Factor 1995 1996 1997 1998 1999 2000 2001 2002 2003 2004
Direct employment 50 50 0 0 0 0 0 0 0 0
Secondary 70 70 0 0 0 0 0 0 0 0
employment
Total employment 120 120 0 0 0 0 0 0 0 0
change
Change in ROIa 0.1 0.1 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
labor force (%)
Change in ROI 0.1 0.1 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
employment (%)
Population change 405 405 0 0 0 0 0 0 0 0
Change in ROI 0.2 0.2 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0
population (%)
a. Sources: Johnson (1995); USBEA (1993); USBC (1992).
b. ROI = region of influence.
5.3.2 Alternatives 1 and 2 - No Action and Decentralization
Activities associated with Alternatives 1 and 2 would not result in any additional construction or
operations jobs at the INEL; therefore, implementation of either of these alternatives would have no
impact on socioeconomic resources in the region of influence.
5.3.3 Alternatives 3, 4a, 4b(1), and 5b - 1992/1993 Planning Basis, Regionalization by Fuel Type,
Regionalization by Geography (INEL), and Centralization at the INEL
5.3.3.1 Construction. As listed in Table 5.3-1, construction employment under these
alternatives would peak during the period from 2001 to 2004 with approximately 375 additional direct
jobs per year. When added to the estimated 528 indirect jobs, the total employment impact in the
region would be an addition of approximately 903 jobs. Employment would decline to zero by 2008.
Based on historic data, approximately 97 percent of the new employees who would fill these jobs
would live in the seven-county region of influence. As listed in Table 5.3-1, if all new jobs (903)
were filled by in-migrants to the region, there would be a 0.8-percent increase in the regional labor
force and in regional employment during the peak years. These changes would be minimal and would
have no adverse impacts on socioeconomic resources in the region. In fact, although the
implementation of any of these alternatives would result in an increase over projected employment
levels, as shown in Figure 5.3-1, there would be an overall decline in employment from projected
1995 levels.
Assuming each new employee represented one household and 3.47 persons per household, there
would be a corresponding increase in regional population levels of 1.1 percent (approximately
3,000 people). 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.
5.3.3.2 Operations. Activities associated with Alternatives 3, 4a, 4b(1), and 5b would not
require any additional operations jobs at the INEL. Therefore, the implementation of either of these
alternatives would have no impact on socioeconomic resources in the region of influence.
5.3.4 Alternatives 4b(2) and 5a - Regionalization by Geography (Elsewhere) and Centralization at Other
DOE Sites
5.3.4.1 Construction. As listed in Table 5.3-2, construction employment under these
alternatives would peak during the period from 1995 to 1996 with approximately 50 additional direct
jobs per year. When added to the estimated 70 indirect jobs, the total employment impact in the
region would be approximately 120 jobs. Employment after 1996 would drop to zero.
Figure 5.3-1. INEL employment by SNF alternative relative to site employment projections. (not available in electronic copy)
Based on historic data, approximately 97 percent of the new employees who would fill these jobs
would live in the seven-county region of influence. As listed in Table 5.3-2, if all new jobs (120)
were filled by in-migrants to the region, there would be a 0.1-percent increase in the regional labor
force and in regional employment levels during the peak years. These changes would be minimal and
would have no adverse impacts on socioeconomic resources in the region. In fact, although the
implementation of any of these alternatives would be an increase over projected employment levels
from 1995 to 1996, as shown in Figure 5.3-1, there would be an overall decline in employment from
projected 1995 levels.
Assuming each new employee represented one household and 3.47 persons per household, there
would be a corresponding increase in regional population levels of 0.2 percent (approximately
400 people). 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.
5.3.4.2 Operations. Activities associated with Alternatives 4b(2) and 5a would not result in
any additional operations jobs at the INEL. Therefore, the implementation of either of these
alternatives would have no impact on socioeconomic resources in the region of influence.
5.4 Cultural Resources
This section summarizes the potential impacts of spent nuclear fuel management activities on
cultural resources at the INEL site.
This assessment evaluated both direct and indirect impacts due to the proposed alternatives. At
the INEL, direct impacts to archaeological resources usually would be those associated with ground
disturbance from construction activities. Direct impacts to existing historic structures could result from
demolition, modification, deterioration, isolation from or alteration of the character of the property's
setting; or introduction of visual, audible, or atmospheric elements out of character or that alter the
property's setting. In addition, indirect impacts to archaeological resources could occur due to an
overall increase in activity at the INEL, which could bring a larger workforce closer to significant
sites. Direct impacts to traditional resources could occur through land disturbance, vandalism, or
changes to the environmental settings of traditional use and sacred areas. Impacts could result from
pollution, noise, and contamination that could affect the traditional hunting and gathering areas or the
visual or audible settings of sacred areas.
The potential for adverse impacts on cultural resources would be the least under Alternatives 1,
2, 4b(2), and 5a, which would disturb approximately 0.8 acres (0.003 square kilometer). Impacts
would be minor because surveys of the area to be disturbed found no eligible cultural resources
(Reed et al. 1986; DOE 1993a).
The potential for adverse impacts on cultural resources would be similar under Alternatives 3, 4a,
4b(1), and 5b with the greatest potential under Alternatives 4b(1) and 5b [Regionalization by
Geography (INEL) and Centralization at the INEL], which would involve the disturbance of nearly 31
acres (0.12 square kilometer). Again, impacts would be minimal because surveys of the previously
disturbed area found no eligible cultural resources (Reed et al. 1986). Under these alternatives,
proposed modifications at the Idaho Chemical Processing Plant facilities could adversely affect
historically significant structures and could require consultation with the Idaho State Historic
Preservation Office (Braun et al. 1993).
The Shoshone-Bannock Tribes are also concerned with the potential impact to important Native
American resources from changes in the visual setting, noise, air quality, or water quality. Because
activities associated with spent nuclear fuel management would take place within existing facility areas
currently engaged in similar activities, DOE does not expect any impacts to important Native
American resources from alteration of the visual setting or noise associated with implementation of
any of the alternatives. There could be temporary, minor impacts on air quality from fugitive dust
associated with construction activities. Emissions of radionuclides to the air under normal operations
would be minor and would be well below applicable standards and guidelines. Under normal
operating conditions, radioactive discharges to the soil or directly to the aquifer would not occur.
DOE would minimize the potential for direct and indirect adverse impacts on traditional use
resources from pollution, noise, and contamination through compliance with applicable local, state, and
Federal laws and regulations. Impact avoidance and other mitigation measures for cultural resources
are described in Section 5.20.2.
5.5 Aesthetic and Scenic Resources
None of the alternatives for spent nuclear fuel management at the INEL would have adverse
consequences on scenic resources or aesthetics because DOE would confine the proposed projects to
developed areas. Although the construction of the proposed facilities would produce fugitive dust that
could temporarily affect visibility, the INEL would follow standard construction practices to minimize
both erosion and dust generation. Facility operations under each alternative would not produce
emissions to the atmosphere that would impact visibility.
5.6 Geology
This section discusses the potential effects of the spent nuclear fuel management alternatives on
geologic resources at the INEL site.
Proposed INEL spent nuclear fuel management activities would only have minor localized
impacts on the geology of the site for all the alternatives. Direct impacts to geologic resources at the
site would be associated with the disturbance or extraction of surface deposits to construct new
facilities. These impacts could include excavations into the soil and rock of the site, soil mounding
and banking, and the extraction of aggregate materials from gravel and borrow pits on the site.
Table 5.6-1 lists estimated extractions of aggregate from site gravel pits for all INEL spent nuclear
fuel, environmental restoration, and waste management projects. These values serve to bound the
spent nuclear fuel project usage.
A secondary impact to geological resources from construction activities would be the potential
for increased soil erosion. DOE would minimize any potential soil erosion by the use of Best
Management Practices designed to control stormwater runoff and slope stability.
Table 5.6-1. Estimated INEL gravel/borrow use (cubic meters). ,b
Alternative Estimated Gravel/Borrow Use
1. No Action 158,000
2. Decentralization 158,000
3. 1992/1993 Planning Basis 392,000
4a. Regionalization by Fuel Type 392,000
4b(1) Regionalization by Geography (INEL) 1,772,000
4b(2) Regionalization by Geography (Elsewhere) 296,000
5a. Centralization at other DOE Sites 296,000
5b. Centralization at the INEL 1,772,000
a. Source: EG&G (1994).
b. To convert cubic meters to cubic yards, multiply by 1.31.
5.7 Air Quality and Related Consequences
This section describes the potential nonradiological and radiological impacts to air quality
associated with each alternative. The term "baseline concentrations" is defined as the sum of the
concentrations resulting from potential emissions from current operations and those resulting from
planned upgrades or modifications that DOE would construct or operate prior to any of the proposed
actions described in this EIS. Additional information is provided in Section 5.7 and Appendix F-3 of
Volume 2.
5.7.1 Alternative 1 - No Action
5.7.1.1 Nonradiological Air Quality. Construction activities associated with this alternative
would be limited to upgrading an existing facility. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. DOE assessed
the impacts from construction using the EPA Fugitive Dust Model (FDM) (Winges 1992). The
modeling results showed that the expected construction-related air quality impacts should be temporary
and highly localized.
Minimal spent nuclear fuel activities would occur under this alternative. Therefore, DOE expects
that the ambient concentrations levels from normal operations would be similar to those from baseline.
Table 4.7-1 lists nonradioactive emissions from normal operations. Tables 5.7-1 and 5.7-2 list the
maximum potential concentrations for the proposed alternatives; they are all below applicable
standards and guidelines. Ambient concentrations from Alternative 1 activities will be below
applicable standards and guidelines.
5.7.1.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
No additional facilities that would be in operation for this alternative would produce radionuclide
emissions. Therefore, for normal operations, doses to the maximally exposed individual, the
population, and workers would be equivalent to baseline doses, as listed in Table 5.7-3. Table 5.7-4
lists associated emission rates.
Table 5.7-1. Maximum impacts to nonradiological air quality from spent nuclear fuel - criteria
pollutants. ,b
Pollutant Averaging Applicable Maximum Baseline plus Percent of
time standard baseline maximum standard
(-g/m3) concentration alternativec
(-g/m3) (-g/m3)
Carbon monoxide 1-hr 40,000 610 610 1.5
8-hr 10,000 280 280 2.8
Nitrogen dioxide Annual 100 4 4 4
Lead Quarterly 1.5 0.001 0.001 <0.1
Particulate matter (PM10) 24-hr 150 80 80 53
Annual 50 5 5 10
Sulfur dioxide 3-hr 1,300 580 580 45
24-hr 365 140 140 38
Annual 80 6 6 7.5
a. Source: Section 5.7 of Volume 2 of this EIS and Belanger et al. (1995).
b. Listed concentrations are the maximum of those calculated at the INEL site boundary, public access roads
inside the INEL site boundary, and the Craters of the Moon Wilderness Area.
c. The listed concentrations are the maximums for any of the proposed alternatives.
Table 5.7-2. Maximum impacts to nonradiological air quality from spent nuclear fuel - toxic air
pollutants. ,b
Pollutant Averaging Applicable Maximum Impact from Percent of
time standard baseline maximum standardd
(-g/m3) concentration alternativec
(-g/m3) (-g/m3)
Ammonia Annual 1.8y102 6.0y100 1.8y100 1
Benzene Annual 1.2y10-1 2.9y10-2 2.3y10-2 19
Formaldehyde Annual 7.7y10-2 1.2y10-2 4.4y10-2 57
Methyl isobutyl ketone Annual 2.1y103 (e) 2.6y101 1
Hydrofluoric acid Annual 2.5y101 (e) 1.8y10-2 <0.1
Tributylphosphate Annual 2.5y101 (e) 6.1y10- 0.2
-2
a. Source: Section 5.7 of Volume 2 of this EIS and Raudsep (1995).
b. Listed concentrations are the maximum of those calculated at the INEL site boundary, public access roads
inside the INEL site boundary, and the Craters of the Moon Wilderness Area.
c. The listed concentrations are the maximums for any of the proposed alternatives, plus new or modified
sources expected to become operational after May 1, 1994.
d. In accordance with State of Idaho regulations for toxic air pollutants, the percent of standard is calculated
based on concentrations resulting from the alternatives and from new or modified sources that have become
operational since May 1, 1994.
e. Baseline concentrations for these pollutants were not analyzed because their emissions were below screening
levels.
Table 5.7-3. Annual dose increments by alternative in comparison to the baseline.
Maximally
INEL worker exposed individual Population
Alternative (millirem) (millirem) (person-rem)b
Baseline 4.3y100c 5.6y10-2 3.4y10-1
1. No Action 3.3y10-4 3.5y10-3 1.0y10-1
2. Decentralization 3.3y10-4 3.5y10-3 1.0y10-1
3. 1992/1993 3.3y10-3 8.0y10-3 1.9y10-1
Planning Basisc
4a. Regionalization by Fuel Type 3.3y10-3 8.0y10-3 1.9y10-1
4b(1). Regionalization by Geography 4.2y10-3 4.8y10-2 3.9y10-1
(INEL)d
4b(2). Regionalization by Geography 7.0y10-5 3.9y10-3 8.3y10-2
(Elsewhere)
5a. Centralization at Other DOE 7.0y10-5 3.9y10-3 8.3y10-2
Sites
5b. Centralization at the INEL 4.2y10-3 4.8y10-2 3.9y10-1
a. Source: Section 5.7 of Volume 2 of this EIS.
b. Population dose is calculated based on the projected population in 2000 or 2010 whichever is higher.
c. Baseline worker dose includes the maximum projected operation of the portable water treatment unit at the
Power Burst Facility area. However, the operation would be temporary (1 to 2 years) and is not
representative of a permanent increase in the baseline. If this facility were not included, the baseline dose to
the worker would be about 0.2 millirem per year.
d. Alternative 4b(1) doses are slightly less than Alternative 5b doses.
5.7.2 Alternative 2 - Decentralization
5.7.2.1 Nonradiological Air Quality. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. The modeling
assessment showed that the expected construction-related air quality impacts should be temporary and
highly localized.
Emissions resulting from normal operations under this alternative would include baseline
emissions and those resulting from the startup of the proposed facilities. Emission rates associated
with startup would be less than 1 percent of those from normal operations. Tables 5.7-1 and 5.7-2 list
the maximum concentrations predicted for the proposed alternatives. Ambient concentrations from
Alternative 2 activities would be below applicable standards and guidelines.
Table 5.7-4. Radionuclide emissions by alternative for spent nuclear fuel projects.
Radionuclides and Emission Rates (Ci/yr)
Project and Location Associated H-3/ Co-60 Kr-85 Xe-131m/ Sr-90/ Sb-125 I-129/ Cs-134 Plutonium Am-241 Others
Alternative C-14 Xe-133 Y-90 I-131 Cs-137
TAN Pool Fuel Transfer Project 1, 2, 3, 4a
a. Drying operations 4b(1), 5b 9.6y102 - - - 2.9y10-2 - 3.4y10-2 - 6.6y10-4 2.2y10-4 -
b. Storage operations 3.9y10-1 - - - - - - - - - -
(Test Area North)
Additional Increased Rack Capacity 3, 4a, 4b(1), 5b 2.0y10-1 1.2y10-8 - - 3.8y10-7 1.0y10-4 - 1.3y10-5 - - 3.1y10-6
(Idaho Chemical Processing Plant)
Dry Fuels Storage Facility 3, 4a, 4b(1), 1.8y10-2 1.9y10-6 - - 1.8y10-5 2.2y10-3 4.2y10-3 6.8y10-7 2.6y10-7 - 1.9y10-5
(Idaho Chemical Processing Plant) 4b(2), 5a, 5b
Fort St. Vrain Spent Fuel Storage 3, 4a, 4b(1), 5b - 5.6y10-8 - - 1.8y10-6 - - 2.4y10-7 5.6y10-7 - 2.4y10-7
(Idaho Chemical Processing Plant)
Increased Rack Capacity 3, 4a, 4b(1), 5b 2.0y10-1 1.2y10-8 - - 3.8y10-7 1.0y10-4 - 1.3y10-5 - - 3.1y10-6
(Idaho Chemical Processing Plant)
EBR-II Blanket Treatment (Argonne 3, 4a, 4b(1), 5b 1.6y102 - 4.9y103 5.1y101 - - - - - - -
National Laboratory - West)
Electrometallurgical Process 3, 4a, 4b(1), 8.4y102 - 1.4y104 1.3y102 - - - - - - -
Demonstration Project (Argonne 4b(2), 5a, 5b
National Laboratory - West)
Spent Fuel Processing Facility 4b(1), 5b 3.1y103 1.9y10-6 5.0y105 - 5.8y10-2 1.6y101 4.4y10- 1.8y10-1 7.7y10-3 - 2.1y10-1
-1
a. Source: Appendix F-3 of Volume 2 of this EIS.
5.7.2.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
Emissions resulting from normal operations under this alternative would include the baseline
emissions and those resulting from the startup of the proposed facilities. Table 5.7-4 lists emission
rates for the spent nuclear fuel alternatives, including Decentralization. Table 5.7-3 lists the resulting
doses to the maximally exposed individual, the population, and workers. These values are small in
comparison to the National Emission Standards for Hazardous Air Pollutants dose limit of 10 millirem
per year, the dose limit received from background sources of 351 millirem per year, and the
population dose from background sources of 40,000 person-rem.
5.7.3 Alternative 3 - 1992/1993 Planning Basis
5.7.3.1 Nonradiological Air Quality. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. The modeling
assessment showed that expected construction-related air quality impacts should be temporary and
highly localized.
Emissions resulting from normal operations under this alternative would include baseline
emissions and those resulting from the proposed facilities. Emission rates associated with startup
would be less than 1 percent of those from normal operations. Tables 5.7-1 and 5.7-2 list the
maximum potential concentrations for the proposed alternatives. Ambient concentrations from
Alternative 3 activities would be below applicable standards and guidelines.
5.7.3.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
Emissions resulting from normal operations under this alternative would include baseline
emissions and those resulting from the startup of the proposed facilities. Table 5.7-4 lists emission
rates for the spent nuclear fuel alternatives. Table 5.7-3 lists the resulting doses to the maximally
exposed individual, the population, and workers. These values are small in comparison to the National
Emission Standards for Hazardous Air Pollutants dose limit of 10 millirem per year, the dose limit
received from background sources of 351 millirem per year, and the population dose from background
sources of 40,000 person-rem.
5.7.4 Alternative 4a - Regionalization by Fuel Type
5.7.4.1 Nonradiological Air Quality. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. The modeling
assessment showed that the expected construction-related air quality impacts should be temporary and
highly localized.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the startup of the proposed facilities. Emission rates associated
with startup would be less than 1 percent of those from normal operations. Tables 5.7-1 and 5.7-2 list
the maximum potential concentrations for the proposed alternatives. Ambient concentrations from
Alternative 4 activities would be below applicable standards and guidelines.
5.7.4.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the proposed facilities. Table 5.7-4 lists emission rates for spent
nuclear fuel alternatives including Regionalization. Table 5.7-3 lists the resulting doses to the
maximally exposed individual, the population, and workers. These values are small in comparison to
the National Emission Standards for Hazardous Air Pollutants dose limit of 10 millirem per year, the
dose limit received from background sources of 351 millirem per year, and the population dose from
background sources of 40,000 person-rem.
5.7.5 Alternative 4b(1) - Regionalization by Geography (INEL)
5.7.5.1 Nonradiological Air Quality. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. The modeling
assessment showed that the expected construction-related air quality impacts should be temporary and
highly localized.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the startup of the proposed facilities. Emission rates associated
with startup would be less than 1 percent of those from normal operations. Tables 5.7-1 and 5.7-2 list
the maximum potential concentrations from the proposed alternatives. Ambient concentrations from
Alternative 4b(1) activities would be below applicable standards and guidelines.
5.7.5.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the proposed facilities. Table 5.7-4 lists associated emission rates
for spent nuclear fuel alternatives including Regionalization by Geography (INEL). Table 5.7-3 lists
resulting doses to the maximally exposed individual, the population, and workers. These values are
small in comparison to the National Emission Standards for Hazardous Air Pollutants dose limit of 10
millirem per year, the dose limit received from background sources of 351 millirem per year, and the
population dose from background sources of 40,000 person-rem.
5.7.6 Alternative 4b(2) - Regionalization by Geography (Elsewhere)
5.7.6.1 Nonradiological Air Quality. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. The modeling
assessment showed that the expected construction-related air quality impacts should be temporary and
highly localized.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the startup of the proposed facilities. Emission rates associated
with startup would be less than 1 percent of those from normal operations. Tables 5.7-1 and 5.7-2 list
the maximum potential concentrations from the proposed alternatives. Ambient concentrations from
Alternative 4b(2) activities would be below applicable standards and guidelines.
5.7.6.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the proposed facilities. Table 5.7-4 lists associated emission rates
for spent nuclear fuel alternatives including Regionalization by Geography (Elsewhere). Table 5.7-3
lists resulting doses to the maximally exposed individual, the population, and workers. These values
are small in comparison to the National Emission Standards for Hazardous Air Pollutants dose limit of
10 millirem per year, the dose limit received from background sources of 351 millirem per year, and
the population dose from background sources of 40,000 person-rem.
5.7.7 Alternative 5a - Centralization at Other DOE Sites
5.7.7.1 Nonradiological Air Quality. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. The modeling
assessment showed that the expected construction-related air quality impacts should be temporary and
highly localized.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the startup of the proposed facilities. Emission rates associated
with startup would be less than 1 percent of those from normal operations. Tables 5.7-1 and 5.7-2 list
the maximum potential concentrations from the proposed alternatives. Ambient concentrations from
Alternative 5a activities would be below applicable standards and guidelines.
5.7.7.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the proposed facilities. Table 5.7-4 lists associated emission rates
for spent nuclear fuel alternatives including Centralization at other DOE sites. Table 5.7-3 lists
resulting doses to the maximally exposed individual, the population, and workers. These values are
small in comparison to the National Emission Standards for Hazardous Air Pollutants dose limit of 10
millirem per year, the dose limit received from background sources of 351 millirem per year, and the
population dose from background sources of 40,000 person-rem.
5.7.8 Alternative 5b - Centralization at the INEL
5.7.8.1 Nonradiological Air Quality. Potential impacts to air quality from construction
activities would include fugitive dust and exhaust emissions from support equipment. The modeling
assessment showed that the expected construction-related air quality impacts should be temporary and
highly localized.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from the proposed facilities. Emission rates associated with the startup
of the proposed facilities would be less than 1 percent of those from normal operations. Tables 5.7-1
and 5.7-2 list the maximum potential concentrations from the proposed alternatives. Ambient
concentrations from Alternative 5b activities would be below applicable standards and guidelines.
5.7.8.2 Radiological Air Quality. No radiological impacts to the environment would result
from construction activities.
Emissions resulting from normal operation under this alternative would include baseline
emissions and those resulting from startup of the proposed facilities. Table 5.7-4 lists associated
emission rates for spent nuclear fuel alternatives including Centralization at the INEL. Table 5.7-3
lists resulting doses to the maximally exposed individual, the population, and workers. These values
are small in comparison to the National Emission Standards for Hazardous Air Pollutants dose limit of
10 millirem per year, the dose limit received from background sources of 351 millirem per year, and
the population dose from background sources of 40,000 person-rem.
5.8 Water Resources and Related Consequences
This section discusses potential environmental consequences to water resources under the five
spent nuclear fuel management alternatives. DOE evaluated each alternative with respect to its
impacts on water quality (both surface and subsurface water), water use, and human health.
Any liquid effluents from facilities proposed for the spent nuclear fuel alternatives would be in
tanks or lined evaporation basins. Under normal operating conditions, radioactive discharges to the
soil or directly to the aquifer would not occur. Creed (1994) presents spent nuclear fuel water quality
data for the analysis of the potential impacts resulting from a hypothetical leak of 20 liters (5 gallons)
per day from secondary containment around the SNF storage pools during operations. Arnett (1994)
addresses the effects that this leak could have on the quality of subsurface water resources.
Preliminary results indicate that there will be no contaminants above maximum contaminant levels at
the INEL boundary resulting from the postulated operational leak. Some storage pools have had
leakage in the past. However, based on the bounding accident scenario for high-level waste tank
failure, leakage during the implementation of the selected spent nuclear fuel management alternative
would cause negligible impacts to water resources (Bowman 1994). None of the proposed alternatives
for the management of spent nuclear fuel would result in any renewed discharges to infiltration ponds.
Section 5.15 discusses potential releases of hazardous or radioactive liquids as a result of accidents.
With respect to water usage, Alternative 4b(l) [Regionalization by Geography (INEL)] and
Alternative Sb (Centralization at the INEL) would consume the largest volume of water- 1.5 million
cubic meters (400 million gallons) over 40 years. The greatest water consumption rate for these
alternatives would be 50,000 cubic meters (13 million gallons) per year (Hendrickson 1995). This
incremental usage would represent approximately a 0.7 percent increase over the total average
withdrawal rate at the INEL of 7.4 million cubic meters (1.9 billion gallons) per year. The INEL's
consumptive use water right is 43 million cubic meters (11.4 billion gallons) per year. Therefore,
Alternatives 4b( I) and Sb would have negligible impact on the quantity of water in the Eastern Snake
River Plain Aquifer.
5.9 Ecology
DOE expects that construction impacts, which would include the loss of some wildlife habitat
due to land clearing and facility development, would be greatest under Alternative 4b(1)
[Regionalization by Geography (INEL)] and Alternative 5b (Centralization at the INEL). Because this
construction activity would take place either within the boundaries of heavily developed areas or
adjacent to those areas, it would have minimal impact on ecological resources. However, construction
activities could provide opportunities for the spread of exotic plant species (e.g., cheatgrass and
Russian thistle).
There would be no construction impacts to wetlands, which would be excluded from
development, and impacts to threatened and endangered species would be unlikely, given the location
(previously-developed areas) and the maximum size [approximately 31 acres (0.125 square
kilometers)] of the affected area. Construction activities at the INEL probably would not affect either
of the endangered species identified in Section 4.9.3 (the bald eagle and peregrine falcon). Both of
these birds of prey are associated with riparian areas, wetlands, and larger bodies of water (e.g.,
reservoirs) and inhabit dry upland areas only temporarily when migrating (National Geographic
Society 1987). Disturbance to other sensitive (but not Federally-listed) species identified in
Section 4.9.3 (e.g., the burrowing owl, northern goshawk, ferruginous hawk, Swainson's hawk,
gyrfalcon, Townsend's western big-eared bat, and pygmy rabbit) would be possible but unlikely, given
the scale of the planned construction. Any impacts would be negligible and short lived, lasting only
as long as the construction activities.
Representative impacts from operations would include the disturbance and displacement of
animals (such as the pronghorn) caused by the movement and noise of personnel, equipment, and
vehicles. Such impacts would be greatest under Alternative 4b(1) [Regionalization by Geography
(INEL)] and Alternative 5b (Centralization at INEL), which would involve a generally higher level of
operational activity; however, these impacts would be minor under all the proposed alternatives.
5.10 Noise
As discussed in Section 4.10, noises generated on the INEL do not travel off the site at levels
that affect the general population. Therefore, INEL noise impacts for each alternative would be
limited to those resulting from the transportation of personnel and materials to and from the site that
would affect nearby communities, and from onsite sources that could affect wildlife near those sources.
Transportation noises would be a function of the size of the workforce (e.g., an increased
workforce would result in increased employee traffic and corresponding increases in deliveries by
truck and rail; a decreased workforce would result in decreased employee traffic and corresponding
decreases in deliveries). This analysis of traffic noise considered railroad noise and noise from major
roadways that provide access to the INEL. 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. Rail shipments of
spent nuclear fuel, regardless of the alternative, would be a small fraction of the rail traffic on the
Blackfoot-to-Arco Branch of the Union Pacific System line that crosses the INEL. The vehicles that
transport employees and personnel on roads would be the principal source of community noise impacts
near the INEL.
This analysis used the day-night average sound level to assess community noise, as suggested by
the EPA (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 projected changes in employment and traffic levels. The analysis also
considers the combination of construction and operation employment. The baseline noise level is
comparable to that for the No-Action alternative. Section 4.10 discusses levels representative of the
No-Action alternative. The traffic noise analysis considered U.S. Highway 20, which employees use
to access the INEL from Idaho Falls. Changes in noise level below 3 decibels probably would not
result in a change in community reaction (FICON 1992).
The new employment associated with each alternative is a small percentage of the total onsite
workforce. The maximum new employment of about 375 INEL onsite jobs would occur with
Alternatives 3, 4a, 4b(1), and 5b during the peak construction period beginning in 2001 (see
Section 5.3, Socioeconomics). No new operations employment is projected for any of the alternatives
except Alternatives 4b(1) and 5b for which there would be 25 new jobs beginning in 2007. The
cumulative onsite workforce under each alternative would be greatest in 1995 and would decrease
thereafter. The peak cumulative onsite workforce for Alternatives 4b(2) and 5a would increase in
1995 by less than 1 percent compared to the No-Action baseline. There would be a corresponding
increase in private vehicle and truck trips to the site. The day-night sound level (DNL) at 15 meters
(50 feet) from the roads that provide access to the INEL probably would increase by less than
1 decibel. The peak cumulative onsite workforce for Alternative 2 in 1995 would be the same as that
for the No-Action baseline.
For any of the alternatives, truck activity would consist of a few trips per day to and from the
site carrying spent nuclear fuel. This increase in truck trips would not result in a perceptible increase
in traffic noise levels along the routes to the INEL. The day-night average sound level along U.S.
Highway 20 and other access routes probably would decrease slightly as a result of the anticipated
overall decrease in employment levels at the INEL. DOE expects no change in the community
reaction to noise along this route and other access routes. No mitigation efforts would be required.
5.11 Traffic and Transportation
5.11.1 Introduction
Spent nuclear fuel management activities involve the transportation of spent nuclear fuel inside
the boundaries of the INEL (onsite) and on highways and rail systems outside the boundaries of the
INEL (offsite). This section summarizes the methods of analysis used to determine the environmental
consequences of onsite transportation of nonnaval spent nuclear fuel under normal conditions
(incident-free) and of transportation accidents. The impacts include doses and health effects.
Appendices D and I of Volume 1 address consequences of shipments to or from the INEL that involve
other DOE sites and spent nuclear fuel-related locations.
5.11.2 Methodology
5.11.2.1 Incident-Free Transpodation. Radiological impacts were determined for two
groups of people during normal incident-free transportation: (1) crewmen (drivers) and (2) members
of the public. Members of the public are persons sharing the transport link (on-link). On-link doses
were determined for Onsite shipments because members of the public have access to the majority of
the roads on the INEL. Radiological impacts were calculated using the RADTRAN 4 (Neuhauser and
Kanipe 1992) and RISKIND (Yuan et al. 1993) computer codes.
The magnitude of the incident-free dose depends mainly on the Transport Index of the shipment
and the on-link vehicle densities. The Transport Index is defined as the dose rate at 1 meter
(3.28 feet) from the surface of a radioactive package; it is measured in millirem per hour. Spent
nuclear fuel was assigned a dose rate of 14 millirem per hour at 1 meter from the shipping container.
This dose rate yielded a dose rate of 10 millirem per hour at 2 meters (6.56 feet) from the edge of the
transport vehicle, which is the regulatory limit for an exclusive use vehicle (see Madsen et al. 1986).
Radiological doses were converted to cancer fatalities using risk conversion factors of
5.0 x lO~ fatal cancer per person-rem for members of the public and 4.0 x 10A fatal cancers per
person-rem for workers. These risk conversion factors are from Publication 60 of the International
Commission on Radiological Protection (ICRP 1991).
Because the onsite transportation of spent nuclear fuel at the INEL is considered rural, no
incident-free nonradiological risk (from exhaust emissions and dust resuspension) was calculated.
5.11.2.2 Accidents. The doses of the maximum reasonably foreseeable onsite spent nuclear
fuel transportation accident were calculated using the RISKIND computer code. Doses were analyzed
for generic rural and suburban population densities, assuming 6 persons per square kilometer for rural
areas and 719 persons per square kilometer for suburban areas. Areas within 80 kilometers (50 miles)
of INEL have population densities between rural and suburban but are closer to the generic rural
population density. Doses were also assessed under both neutral and stable atmospheric conditions.
Radiation doses calculated were used to estimate the potential for fatal cancers in the exposed
population using risk factors developed by the International Commission on Radiological Protection
(ICRP 1991).
The probability of the maximum reasonably foreseeable onsite spent nuclear fuel transportation
accident was estimated taking into account spent nuclear fuel handling procedures within the Advanced
Test Reactor facility as well as factors related to transportation of the spent nuclear fuel. For this
accident to occur, errors must occur in loading the wrong spent nuclear fuel into the shipping cask,
radiation surveys of the loaded cask fail to detect abnormally high radiation levels, the transport
vehicle must breakdown or rollover during the short transit between the Advanced Test Reactor and
the Idaho Chemical Processing Plant, and operators fail to ensure that adequate cooling water is
maintained inside the cask. The estimated probability of this accident is no greater than once in a
million years.
The risk of the onsite spent nuclear fuel transportation accident was estimated by multiplying the
accident doses by the accident probability, taking into account the probability of the atmospheric
conditions used. The resulting risk value gives a bounding estimate of the annual probability of fatal
cancers occurring in the local population due to onsite spent nuclear fuel transportation accidents.
5.11.3 Onsite Spent Nuclear Fuel Shipments
For each spent nuclear fuel management alternative, a small number of onsite DOE spent nuclear
fuel shipments would be likely each year as a result of continuing reactor operations at the Advanced
Test Reactor and the Experimental Breeder Reactor-li. The alternatives would not affect the operation
of these two facilities, thus the shipments be'tween these facilities and the Idaho Chemical Processing
Plant, integrated over 40 years, would be the same for each spent nuclear fuel management alternative.
Spent nuclear fuel shipments to the Idaho Chemical Processing Plant from four locations on the
INEL (including the Test Reactor Area, Argonne National Laboratory-West, Test Area North, and
Power Burst Facility) were evaluated. The number of shipments would not change with alternatives
because DOE plans to ship all spent nuclear fuel to the Idaho Chemical Processing Plant. Alternatives
that would ship spent nuclear fuel off the site under Regionalization [Alternatives 4a, 4b( 1) and 4b(2)]
and Centralization (Alterntives Sa and Sb) would ship it first to the Idaho Chemical Processing Plant
for canning or other stabilization prior to shipment. DOE estimated the total projected number of
shipments over 40 years of operation (1995-2035) from each facility from either historic records or
current inventories. DOE based the projected number of shipments for Test Reactor Area and
Argonne National Laboratory-West to the Idaho Chemical Processing Plant on historic records for
1987 through 1992, and the doses reflect shipments for 1995 through 2035. The projected number of
shipments from Test Area North would include Three Mile Island canisters, Loss of Fluid Test fuel,
special case commercial fuel, and non-fuel-bearing components stored in the Test Area North pool.
The projected number of shipments from the Power Burst Facility includes all spent nuclear fuel stored
at that facility.
Onsite shipments would include those that originated and ended on the INEL site. Shipments
that originate or terminate at non-INEL facilities are offsite shipments. Appendixes D and I describe
the consequences of naval and DOE offsite spent fuel shipments, respectively. Movements of spent
nuclear fuel inside (INEL) facility fences (e.g., from the CPP-603 Underwater Storage Facility to the
Fuel Storage Area) are operational transfers, not onsite shipments; therefore, this section does not
consider such shipments
5.11.4 Incident-Free Impacts
The occupational and general population collective doses from onsite spent nuclear fuel
shipments and the resulting incidence of latent cancer fatalities were calculated. The results are the
same regardless of alternative. Occupational radiation exposure would potentially be 3.4 person-rem,
resulting in 0.0014 latent cancer fatalities. General population exposure would potentially be 0.088
person-rem, resulting in 0.000044 latent cancer fatalities.
In addition to collective radiation exposure, the maximally exposed individual doses due to INEL
onsite SNF shipments were calculated for a driver (occupational exposure), a person following a single
shipment, and a person standing beside the road as a single shipment passes by (general member of
the public). The calculated dose to a driver would be 1.7 rem, assuming that person drove all
shipments over 40 years. The calculated maximally exposed individual dose to a person following a
single shipment covering the longest distance from Test Area North to the Idaho Chemical Processing
Plant would be 0.015 millirem, and to a person exposed to passing shipment at a distance of 1 meter
(3.28 feet), the dose would be 0.0014 millirem (Maheras 1995).
Traffic impacts for the spent nuclear fuel shipments were estimated from data in Heiselmann
(1994). The maximum number of spent nuclear fuel shipments of 691 per year would occur with
Alternative Sb, Centralization at the INEL. A maximum 23-percent increase in traffic volume per day
would occur with this alternative, based on the estimates of the number of trips required for the
transport of construction equipment, material, spent nuclear fuel, other wastes, and workers to and
from the INEL. Even if this average daily traffic volume were to occur for 1 hour, the maximum
traffic volume would increase to 145 vehicles per hour for US 20, US 26, Routes 33 and 22; this
would not change the baseline level of service, which is designated as "free flow."
5.11.5 Accident Impacts
An onsite spent nuclear fuel transportation accident involving the inadvertent shipment of a short-
cooled fuel element from the Advanced Test Reactor to the Idaho Chemical Processing Plant was
considered to be the maximum reasonably foreseeable accident. The melted spent nuclear fuel has
potential to relocate into a critical configuration. However, the probability of a criticality accident is
much less than 1 x l0(-7) per year and would be considered to be not reasonably foreseeable. Table
5.11-1 lists the calculated maximally exposed individual dose and collective dose to general population
in the maximally impacted sector and corresponding risk of fatal cancers. The dose to the maximally
exposed individual is considered an occupational exposure.
As listed in Table 5.11-1, the total number of fatal cancers expected in the suburban population
affected by the transportation for neutral and stable meteorological conditions would be 11 and 85,
respectively. For the neutral case, this would represent a 0.01-percent increase from the number of
fatal cancers that would be likely from normal incidence in the affected population. For the stable
case, this would represent a 0.20-percent increase from the number of fatal cancers that would be
likely from normal incidence in the affected population.
The total number of fatal cancers expected in the rural population affected by the transportation
for neutral and stable meteorological conditions would be 0.75 and 6.0, respectively. For the neutral
Table 5.11-1. Impacts from maximum reasonably foreseeable spent nuclear fuel transportation accident on INELa (using generic rural
and suburban population densities).
Population Meteorologyc Accident Dose to MEIe Offsite Risk of
density frequencyd (rem) population dose fatal cancer
categoryb (events/yr) (person-rem) per yearf
Rural Neutral 1.0y10-6 7.6y101 1.5y103 7.5y10-7
(7.5y10-1)
Rural Stable 1.0y10-7 2.5y102 1.2y104 6.0y10-7
(6.0y100)
Suburban Neutral 1.0y10-6 7.6y101 2.1y104 1.1y10-5
(1.1y101)
Suburban Stable 1.0y10-7 2.5y102 1.7y105 8.5y10-6
(8.5y101)
a. Source: Enyeart (1994).
b. Results are for generic rural and suburban population densities. The generic rural population density has an average population of 6
persons per square kilometer; the generic suburban population density has an average population of 719 persons per square kilometer. For
comparison, the sector with the highest population density within 80 kilometers (50 miles) is due east of the Idaho Chemical Processing
Plant and Test Reactor Area at the INEL with an average population density of 53 persons/km2.
c. Neutral meteorology is characterized by Stability Class D, 4 meters-per-second wind speed, and occuring approximately 50 percent of the
time. Stable meteorology is characterized by Stability Class F, 1 meter-per-second wind speed, and occuring approximately 5 percent of
the time.
d. Accident frequency includes both the event frequency and the frequency of the meteorology. The frequency of stable meteorology is
approximately one-tenth the frequency of neutral meteorology.
e. Maximally exposed individual located at the point of maximum exposure to the airborne release approximately 160 to 390 meters (525 to
1,280 feet) downwind, depending on meteorology. For onsite accidents the maximally exposed individual is assumed to be an INEL
worker.
f. Fatal cancer risk = dose times accident frequency times (ICRP 60 risk factor for fatal cancers). The ICRP 60 risk factor is 5.0 y 10-4 fatal
cancer per rem for public, 4.0 y 10-4 fatal cancer per rem for workers. For doses of 20 rem or more, the ICRP 60 conversion factor is
doubled. Numbers in parentheses indicate the total number of fatal cancers in the population if the accident occurs. The maximally
exposed individual dose is considered an occupational exposure.
case, this would represent a 0.09-percent increase from the number of fatal cancers that would be
likely from normal incidences in the affected population. For the stable case, this would represent a
1.7-percent increase from the number of fatal cancers that would he likely from normal incidence in
the affected population.
The estimated maximum nonradiological occupational and general population traffic fatalities
over 40 years due to any of the spent nuclear fuel management alternatives would be 7.1 x 10(-4) and
2.5 x 10(-3), respectively. These estimated fatalities were based on fatality risk factors for spent fuel
shipments (Cashwell et. al 1986).
5.11.6 Onsite Mitigative and Preventative Measures
All onsite shipments would be in compliance with DOE ID Directive 5480.3, "Hazardous
Materials Packaging and Transportation Safety Requirements." These requirements provide assurance
that, under normal conditions, the INEL would meet as-low-as-reasonably-achievable conditions,
reasonably foreseeable accident situations (those with a probability of occurrence greater than 1 x 10(-7)
per year) would not result in a loss of shielding or containment or a criticality, and an unintentional
release of radioactive maSerial would generate a timely response.
DOE would approve the type packages used for onsite shipments or would obtain a Nuclear
Regulatory Commission or DOE certificate of compliance. If the Type B onsite package did not have
Nuclear Regulatory Commission or DOE certification, the user of the package would have to establish
how administrative controls and site-mitigating circumstances would ensure that the package would
maintain containment and shielding integrity. The administrative and emergency response
considerations would provide sufficient control so that accidents would not result in loss of
containment or shielding, in criticality, or in an uncontrolled release of radioactive material that would
create a hazard to the health and safety of the public or workers.
In the event of an accident, each DOE site has an established emergency management program.
This program incorporates activities associated with emergency planning, preparedness, and response.
Participating government agencies with plans that are interrelated with the INEL Emergency Plan for
Action include the State of Idaho, Bingham County, Bonneville County, Butte County, Clark County,
Jefferson County, the Bureau of Indian Affairs, and Fort Hall Indian Reservation. When an
emergency condition exists at a facility, the Emergency Action Director is responsible for recognition,
classification, notification, and protective action recommendations. At INEL emergency preparedness
resources include fire protection, radiological and hazardous chemical material response, emergency
control center, the INEL Warning Communication Center, the INEL Site Emergency Operational
Center, and medical facilities.
5.12 Occupational and Public Health and Safety
This section presents DOE's estimates of the health effects from spent nuclear fuel-related
activities at the INEL for the following human receptor groups:
- Involved Workers - workers at the facilities involved with spent nuclear fuel alternatives,
including existing workers and new hires for selected alternative
- Maximally Exposed Individual (MEI) - person residing at the INEL site boundary
- Population - the general offsite population in the INEL region
- Construction Worker - labor force associated with construction activities
- Nonconstruction Worker - DOE labor force associated with nonconstruction activities
Radiological, chemical, and industrial safety hazards were considered in the estimates.
5.12.1 Radiological Exposure and Health Effects
The measure of impact used for evaluation of potential radiation exposures is risk of fatal
cancers. Worker and maximally exposed individual effects are reported as individual radiation dose
(in rem) and the estimated lifetime probability of fatal cancer. Population effects are reported as
collective radiation dose (in person-rem) and the estimated number of fatal cancers in the affected
population. Tables 5.12-1, 5.12-2, 5.12-3, and 5.12-4 summarize the radiological health effects
calculations for each alternative.
Activities that workers would perform under each of the alternatives would be similar to those
currently performed at the INEL. Therefore, the potential hazards encountered in the workplace would
be similar to those that currently exist at the INEL. Further, DOE would mitigate these hazards with
occupational and radiological safety programs operating under the same regulatory standards and limits
that currently apply at the INEL. For these reasons, DOE anticipates that the average radiation dose
Table 5.12-1. Annual occupational radiation exposure and employment summary.
No Action Decentralization 1992/1993 Regionalization Centralization Centralization at
(1) (2) Planning Basis by Fuel Type at Other DOE the INEL (5b)
(3) (4a)b Sites (5a)
Number of Workers 1 1 200 200 10 200
(annual average
over years 1995-
2004)c
Worker Collective 0.027 0.027 5.4 5.4 0.27 5.4
Dosed
(person-rem/year)
a. Source: Johnson (1995).
b. Alternative 4b(1), Regionalization by Geography (INEL), values are the same as those for Alternative 5b. Alternative 4b(2),
Regionalization by Geography (Elsewhere), values are the same as those for Alternative 5a.
c. This 10-year average yields conservatively high employment; the 40-year average would be lower but data do not exist.
d. Based on thermoluminescence dosimetry records.
Table 5.12-2. Annual nonoccupational radiation exposure summary.
No Action Decentralization 1992/1993 Regionalization Centralization Centralization at
(1) (2) Planning Basis by Fuel Type at Other DOE the INEL (5b)
(3) (4a)b Sites (5a)
MEI Dose 3.5y10-3 3.5y10-3 8.0y10-3 8.0y10-3 3.9y10-3 4.8y10-2
(mrem/year)
Population 1.0y10-1 1.0y10-1 1.9y10-1 1.9y10-1 8.3y10-2 3.9y10-1
Dosea
(person-
rem/year)
a. Population dose is calculated based on the projected population in 2000.
b. Alternative 4b(1), Regionalization by Geography (INEL), values are the same as those for Alternative 5b. Alternative 4b(2),
Regionalization by Geography (Elsewhere), values are the same as those for Alternative 5a.
Table 5.12-3. Annual fatal cancer incidence and probability summary from radiological exposure.
No Action Decentralization 1992/1993 Regionalization Centralization Centralization
(1) (2) Planning Basis by Fuel at Other DOE at the INEL
(3) Type(4a)b Sites (5a) (5b)
Worker
probability 1y10-5 1y10-5 1y10-5 1y10-5 1y10-5 1y10-5
incidence 1y10-5 1y10-5 2y10-3 2y10-3 1y10-4 2y10-3
Maximally
exposed member
of the public
probability 2y10-9 2y10-9 4y10-9 4y10-9 2y10-9 2y10-8
Population 5y10-5 5y10-5 1y10-4 1y10-4 4y10-5 2y10-4
incidence
a. Risk factors for the worker (4y10-4 probability of occurrence per rem) or offsite population (5y10-4 probability of occurrence per rem)
recommended by the International Commission on Radiological Protection (ICRP 1991).
b. Alternative 4b(1), Regionalization by Geography (INEL), values are the same as those for Alternative 5b. Alternative 4b(2),
Regionalization by Geography (Elsewhere), values are the same as those for Alternative 5a.
Table 5.12-4. 40-year fatal cancer incidence summary from radiological exposure.
No Action Decentralization 1992/1993 Regionalization by Centralization at Centralization at
(1) (2) Planning Fuel Type (4a) Other DOE the INEL (5b)
Basis (3) Sites (5a)
Workers
incidence 4y10-4 4y10-4 8y10-2 8y10-2 4y10-3 8y10-2
Population
incidence 2y10-3 2y10-3 4y10-3 4y10-3 2y10-3 8y10-3
a. Alternative 4b(1), Regionalization by Geography (INEL), values are the same as those for Alternative 5b. Alternative 4b(2),
Regionalization by Geography (Elsewhere), values are the same as those for Alternative 5a.
and the number of reportable cases of injury and illness would be proportional to the number of
workers at the INEL under each alternative.
Table 5.12-1 lists involved worker doses based on an historic annual average dose of 27 mrem
determined from thermoluminescent dosimeter data of workers involved in various INEL radiological
work over the period 1987 to 1991 (see Appendix F of Volume 2). As mentioned above, the hazards
associated with spent nuclear fuel activities are the same as the hazards associated with other INEL
activities. Table 5.12-2 lists the exposure summaries for the maximally exposed individual and offsite
population, based on radioactive emissions from normal operations and those resulting from startup of
proposed facilities for the various alternatives. Note that population collective dose is higher than
worker collective dose only under alternatives 1 and 2. For the alternatives, there is only 1 SNF
worker averaged over 40 years. The nonoccupational population has more people to be exposed.
When the worker population increases under Alternatives 3, 4, and 5, the worker dose becomes higher
than the population dose. Section 5.7 presents the exposure information. Dose calculations are based
on air emissions only, and not water pathways because none of the alternatives would involve the
discharge of pollutants to surface waters or to the subsurface. Section 5.8 summarizes water quality.
Table 5.12-3 summarizes the fatal cancer incidence and probability for workers, maximally
exposed individuals, and the offsite population based on the risk factors consistent with those
recommended by the International Commission on Radiological Protection (ICRP 1991). For all
alternatives, the probability of developing fatal cancer for any individual would be low, with the
maximum value of 1 y 10-5 for the involved worker. The calculated incidence of fatal cancer for the
total number of workers for each alternative and the offsite population would be less than 1.
Table 5.12-4 summarizes the 40-year projection of fatal cancer incidence associated with the
worker and offsite populations. The highest involved worker and offsite population incidence, 0.1 and
0.01, respectively, would be associated with Alternative 5b.
Radiation doses associated with construction activities would be as low as reasonably achievable
and no greater than 2 rem per year to any worker. Historical offsite doses associated with the INEL
are summarized in the Idaho National Engineering Laboratory Historical Dose Evaluation (DOE 1991).
The Centers for Disease Control and Prevention is conducting a more comprehensive reconstruction of
doses from INEL operations.
5.12.2 Nonradiological Exposure and Health Effects
The air quality data listed in Tables 5.7-1 and 5.7-2 were used to evaluate health impacts
associated with potential exposure to two compound classes, criteria pollutant and toxic. Table 5.7-1
lists five pollutant criteria and Table 5.7-2 lists six toxic air pollutant compounds. The toxic
compounds were classified as noncarcinogens or carcinogens, consistent with EPA designations
published in the Integrated Risk Information System (IRIS) data base. However, the IRIS data base
does not include sufficient data to perform a quantitative inhalation cancer risk assessment.
Nonradiological health effects (hazard indices) for the INEL worker or maximally exposed
individual were estimated by summing the ratios of the appropriate pollutant concentrations and their
applicable standards presented in Table 5.7-1 and Table 5.7-2. Table 5.7-1 presents criteria pollutant
concentrations at public access roads, which are the maximum of those calculated at the INEL site
boundary, public access roads inside the INEL site boundary, and the Craters of the Moon Wilderness
Area. The hazard index for the five criteria pollutants is less than 1 (0.2) for the workers or the
maximally exposed individual, based on concentrations for the longest averaging times presented in
Table 5.7-1. Table 5.7-2 presents toxic air pollutant concentrations at the public access roads, which
are the maximum when compared with concentrations at the INEL site boundary and the Craters of the
Moon Wilderness Area. The hazard index for the toxic air pollutants is also less than 1 (0.8) for the
workers or the maximally exposed individual, based on concentrations with annual averaging time
consideration. Accordingly, health effects are unlikely for either the criteria pollutants or the toxic air
pollutants from spent nuclear fuel-related activities. The hazard index is not a statistical probability;
therefore, it cannot be interpreted as such.
5.12.3 Industrial Safety
This section describes the following measures of impact for workplace hazards: (1) total
reportable injuries and illness and (2) fatalities in the work force. This analysis considered injury and
fatality rates for construction workers only since the alternatives do not result in incremental changes
in operations employment. Table 5.12-5 lists the maximum annual number of projected injuries and
illnesses and fatalities for construction workers by alternatives based on the maximum employment
levels for any year between 1995-2035.
Table 5.12-5. Annual industrial safety health effects incidence summary. ,b
No Decentralization 1992/1993 Regionalization Centralization at Centralization at
Action (2) Planning Basis by Fuel Type other DOE Sites the INEL (5b)
(1) (3) (4a)c (5a)
Construction workers
Injury/illness 0 0 23 23 3 23
Fatality 0 0 <1 <1 <1 <1
a. 1988-1992 averages for occupational injury/illness and fatality rates for DOE and contractor employees.
b. Sources: DOE (1993b) and Section 5.3 of this appendix.
c. Alternative 4b(1) values are the same as those for Alternative 5b. Alternative 4b(2) values are the same as those for Alternative 5a.
5.13 Idaho National Engineering Laboratory Services
This section discusses the potential impacts from spent nuclear fuel management on utilities and
energy at the INEL. It considers the consumption of water, electrical energy, fossil-based fuels, and
wastewater discharge at the INEL site.
5.13.1 Construction
Table 5.13-1 summarizes estimates of annual requirements for electricity, water, wastewater, and
diesel fuel for construction activities associated with each alternative and compares them to projected
1995 use levels for these resources. In general, the smallest increase in the demand for site services
would result from Alternatives 4b(2) and 5a [Regionalization by Geography (Elsewhere) and
Centralization at Other DOE Sites] and the largest increase would be associated with Alternatives
4b(1) and 5b [Regionalization by Geography (INEL) and Centralization at INEL].
Table 5.13-1. Estimated increase in annual electricity, water, wastewater treatment, and fuel
requirements for construction activities associated with each alternative.
Service Projected Estimated additional demand
1995 usage construction
w/o
Alternative
Alternatives Alternatives Alternatives Alternatives
1 and 2 3 and 4a 4b(1) and 5b 4b(2) and 5a
Electricity (MWHa per year) 208,000 71 150 2,100 10
Water (millions of liters per year)b 6,450 No increase 2.1 2.2 0.5
Sanitary wastewater (millions of 540 No increase 1.5 4.5 0.5
liters per year)
Diesel fuel (liters per year) 5,830,000 6,400 8,500 14,000 1,500
a. MWH = megawatt hours.
b. To convert liters to gallons, multiply by 0.264.
Source: Hendrickson (1995).
Under Alternatives 4b(1) and 5b, the estimated annual increases in utility and energy usage rates
from construction activities would be 2,100 megawatt-hours of electricity, 2.2 million liters
(580,000 gallons) of water, 4.5 million liters (1,200,000 gallons) of wastewater discharge, and
14,000 liters (3,700 gallons) of diesel fuel. These changes represent modest increases ranging from
near zero percent to 1.0 percent above projected 1995 usage levels and are well within current system
capabilities and usage limits (see Section 4.13). The other alternatives would result in smaller
increases in energy usage and would have no adverse impact on utility services at the INEL.
5.13.2 Operations
Table 5.13-2 summarizes estimates of annual requirements for electricity, water, wastewater, and
fuel for operations activities associated with each alternative and compares them to project 1995 INEL
usage of these resources. In general, the smallest increase in the demand for site services would result
from Alternatives 1 and 2 (No-Action and Decentralization) and the largest would be associated with
Alternatives 4b(1) and 5b [Regionalization by Geography (INEL) and Centralization at INEL].
Table 5.13-2. Estimated increase in annual electricity, water, wastewater treatment, and fuel
requirements for operations activities associated with each alternative.
Service Projected Estimated additional demand
1995 usage operation
w/o
Alternative
Alternatives Alternatives Alternatives Alternatives
1 and 2 3 and 4a 4b(1) and 5b 4b(2) and 5a
Electricity (MWHa per year) 208,000 180 2,200 11,000 2,000
Water (millions of liters per year)b 6,450 No increase No increase 48 No increase
Sanitary wastewater (millions of 540 No increase No increase 0.3 No increase
liters per year)c
Fuel oil (liters per year) 11,100,000 28,000 330,000 1,100,000 300,000
a. MWH = megawatt hours.
b. To convert liters to gallons, multiply by 0.264.
c. Some industrial wastewater, such as steam condensate, is also discharged to evaporation ponds and injection wells.
Sources: Hendrickson (1995).
Under Alternatives 4b(1) and 5b, the estimated annual increases in utility and energy usage rates
from operations activities would be 11,000 megawatt-hours of electricity, 48 million liters (13 million
gallons) of water, 0.3 million liters (79,000 gallons) of wastewater, and 1,100,000 liters
(290,000 gallons) of fuel oil. These changes represent modest increases ranging from near zero
percent to 10 percent and are well within current system capabilities and usage limits (see
Section 4.13). The other alternatives would result in smaller increases in energy usage and would
have no adverse impact on utility services at the INEL.
5.14 Materials and Waste Management
This section discusses the impacts to the management of materials and wastes at the INEL site
and Idaho Falls facilities as a result of the implementation of the spent nuclear fuel management
alternatives. Alternatives 4b(1), and 5b, both with the spent fuel processing option, each establish the
upper bound of potential impacts on projected rates of generation, treatment, storage, and disposal
inventories of materials and wastes. Table 5.14-1 and 5.14-2 summarize waste generation projections
for each alternative. The tables present average generating rates over the life cycle of each alternative
and maximum annual increments over peak generation periods.
5.14.1 Alternative 1 - No Action
Under the No Action Alternative, 9 cubic meters of industrial solid waste would be generated
during construction of the Alternate Fuel Storage Facility for the TAN Pool Fuel Transfer Project at
the Idaho Chemical Processing Plant. At the completion of this project in 1998, there would be
485 cubic meters of non-fuel solid low-level waste consisting of Three Mile Island hardware and
metals that would be removed and dispositioned in a separate project. These impacts apply also to the
description of impacts for the other spent nuclear fuel management alternatives with the exception of
Alternatives 4b(2) and 5a. The non-fuel solid low-level waste is already existing; therefore, it is not
included in Table 5.14-1 as an increase in low-level waste generation.
5.14.2 Alternative 2 - Decentralization
In general, the character of the impacts to materials and waste management would be similar to
those under the No Action Alternative.
5.14.3 Alternative 3 - 1992/1993 Planning Basis
Industrial solid waste would be generated from construction and operation of the various SNF
projects under Alternative 3. This nonradioactive waste would be disposed of in the Central Facilities
Area landfill. Landfill space is nonrestrictive for industrial solid waste disposal. Construction phase
activities would generate a cumulative total of 620 cubic meters of industrial and commercial solid
Table 5.14-1. Average annual waste generation projections for selected SNF management alternatives at INEL.
Average annual increment over 1995 baseline
Alternative Waste type Phase Period Increase Annual rate
(years) (percent) (cubic meters per year)
No Action (Alternative 1) and Decentralization Industrial Construction 1995-1996 0.02 9
(Alternative 2)
1992/1993 Planning Basis Industrial Construction 1995-2005 0.1 62
(Alternative 3) and Regionalization by Fuel Operation 1996-2035 1.2 600
Type (Alternative 4a) Low-Levelb,c Construction 1995-1999 8.6 370
Operation 1996-2035 4.6 200
High-Level Operation 1996-2024 0.1 3
Mixed Low-Level Operation 1996-2024 <0.1 <1
Transuranic Operation 1996-2024 530 32
Regionalization by Geography (INEL) Industrial Construction 1995-2008 0.6 290
[Alternative 4b(1)] and Centralization at INEL Operation 1996-2035 5.0 2,600
(Alternative 5b) Low-Levelb,c Construction 1995-1999 8.6 370
Operation 1996-2035 9.6 410
High-Level Operation 1996-2035 15.7 120
Mixed Low-Level Operation 1996-2024 <0.1 <1
Transuranic Operation 1996-2024 530 32
Regionalization by Geography (Elsewhere) Industrial Construction 1995-1996 <0.1 50
[Alternative 4b(2)] and Centralization at Other Operation 1996-2024 0.4 210
DOE Sites (Alternative 5a) Low-Level Operation 1996-2024 1.9 83
High-Level Operation 1996-2024 0.1 3
Mixed Low-Level Operation 1996-2024 <0.1 <1
Transuranic Operation 1996-2024 530 32
a. Source: Appendix C of Volume 2 of this EIS.
b. Low-level waste from TAN Pool Fuel Transfer Project to be removed and dispositioned in a separate project not included for any alternatives.
c. Low-level waste generated from dispositioning and decontamination of fuel racks not included in any alternatives.
Table 5.14-2. Peak waste generation highlights for selected SNF management alternatives at INEL.
Maximum increment over 1995 baseline
Alternative Waste type Phase Period Increase Annual rate
(years) (percent) (cubic meters per year)
No Action (Alternative 1) and Decentralization Industrial Construction 1995-1996 0.02 9
(Alternative 2)
1992/1993 Planning Basis Industrial Construction 1995-1996 0.4 220
(Alternative 3) and Regionalization by Fuel Operation 2005-2021 1.6 810
Type (Alternative 4a) Low-Levelb,c Construction 1995-1997 13.4 570
Operation 2005-2024 6.1 260
Concurrent Activityd 1996-1997 14.2 610
High-Level Operation 1997-1998 0.2 6
Mixed Low-Level Operation 1997-1998 <0.1 <1
Transuranic Operation 1997-1998 600 36
Regionalization by Geography (INEL) Industrial Construction 1999-2006 0.9 450
[Alternative 4b(1)] and Centralization at INEL Operation 2008-2021 6.8 3,500
(Alternative 5b) Low-Levelb,c Construction 1995-1997 13.4 570
Operation 2008-2024 13.3 570
Concurrent Activityd 1996-1997 14.2 610
High-Level Operation 2005-2024 21.1 160
Mixed Low-Level Operation 1997-1998 <0.1 <1
Transuranic Operation 1997-1998 600 36
Regionalization by Geography (Elsewhere) Industrial Construction 1995-1996 <0.1 50
[Alternative 4b(2)] and Centralization at Other Operation 1996-2024 0.4 210
DOE Sites (Alternative 5a) Low-Level Operation 1996-2010 3.1 130
High-Level Operation 1996-2024 0.1 3
Mixed Low-Level Operation 1996-2024 <0.1 <1
Transuranic Operation 1996-2024 530 32
a. Source: Appendix C of Volume 2 of this EIS.
b. Low-level waste from TAN Pool Fuel Transfer Project to be removed and dispositioned in a separate project not included for any alternatives.
c. Low-level waste generated from dispositioning and decontamination of fuel racks not included in any alternatives.
d. Construction and operations occurring simultaneously.
waste. The Fuel Receiving, Canning, Characterization, and Shipping Facility will generate the most
industrial waste of any of the projects, 490 cubic meters per year from 2005 through 2035.
In addition, the Fuel Receiving, Canning, Characterization, and Shipping Facility will generate
220 cubic meters per year of low-level waste during the same period. The Dry Storage Facility would
generate an additional 5 cubic meters of low-level waste annually from 2005 through 2035. Including
liquid low-level waste, the Increased Rack Capacity and Additional Increased Rack Capacity projects
would increase generation rates by 570 cubic meters annually during construction from 1995 through
1997. Low-level waste would decrease to approximately 160 cubic meters per year from 1997 through
1999 with the completion of the Increased Rack Capacity project. Liquid low-level waste would be
disposed in existing liquid waste processing systems at the Idaho Chemical Processing Plant. Solid
radioactive wastes would be packaged and disposed of at the Radioactive Waste Management
Complex, or incinerated at the Waste Experimental Reduction Facility, whichever is appropriate.
Low-level waste from reracking fuel racks for the Increased Rack Capacity Project will be
decontaminated and dispositioned by a licensed commercial vendor.
Experimental Breeder Reactor-II Blanket Treatment will generate 7 cubic meters of low-level
waste for 1 year from 1997 to 1998.
The storage of low-level waste for incineration is not considered to be restrictive between 1995
through 2005. However, beyond 2005, low-level waste storage capacity may become strained. Use of
commercial facilities to incinerate the backlog of low-level waste is under consideration in order to
reduce or prevent the accumulation of low-level waste, but no firm commitment or contract has yet
been established (EG&G 1993a).
The Radioactive Waste Management Complex appears to have adequate disposal capacity for
low-level waste between 1995 and 2005. However, beyond 2005, additional capacity may be required.
Excess capacity would be provided with the development of the proposed Low-Level Waste/Mixed
Low-Level Waste Disposal Facility (EG&G 1993a).
The Electrometallurgical Process Demonstration Project will generate high-level, mixed low-
level, low-level, transuranic, and industrial wastes from the demonstration and testing of new spent
fuel management processes from 1996 through 2024.
Experimental Breeder Reactor-II Blanket Treatment will also generate high-level, mixed low-
level, and transuranic wastes.
High-level waste would be immobilized after 2005, and may eventually be transported to a
Federal high-level waste and spent nuclear fuel repository for disposal. Transuranic waste meeting
waste acceptance criteria to be developed could be shipped to a potential Federal repository for
disposal should one be selected (EG&G 1993a).
5.14.4 Alternative 4a - Regionalization by Fuel Type
In general, the character of the impacts to materials and waste management would be similar to
those under Alternative 3.
5.14.5 Alternative 4b(1) - Regionalization by Geography (INEL)
The character and intensity of impacts on waste management activities at the INEL are similar to
those under Alternatives 3 and 4a for some of the SNF management projects including the TAN Pool
Fuel Transfer Project at the Idaho Chemical Processing Plant; the Increased Rack Capacity and
Additional Increased Rack Capacity projects; the Experimental Breeder Reactor-II Blanket Treatment
facility; and the Electrometallurgical Process Demonstration Project. Under Alternative 4b(1), the Dry
Fuel Storage Facility is expanded and Fuel Receiving, Canning/Characterization, and Shipping Facility
waste streams decrease relative to Alternatives 3 and 4a; however, the net effect of these differences
on industrial/commercial solid waste generation and low-level waste generation for both construction
and operation results in waste generation rates similar to those under Alternatives 3 and 4a.
The increase in average and peak generation rates over Alternatives 3 and 4a (Tables 5.14-1 and
5.14-2) is due to the Spent Fuel Processing option included under Alternative 4b(1), which accounts
for the relative increase in generation rates over Alternatives 3 and 4a. Fuel processing would be done
in order to stabilize the spent nuclear fuel and remove risks associated with storage and disposal, and
to manage the resultant high-level waste in a cost-effective manner. If this alternative were pursued
aggressively, the generated high-level waste residual resulting from segregating fissile material from
the spent nuclear fuel may require additional high-level waste tankage. This increase in capacity
would be covered by the High-Level Tank Farm New Tanks project described in Volume 2 of the EIS.
Capacity discussions for industrial/commercial solid waste and low-level waste under
Alternative 3 apply to Alternative 4b(1).
5.14.6 Alternative 4b(2) - Regionalization by Geography (Elsewhere)
Construction phase activities would generate a cumulative total of 50 cubic meters of industrial
and commercial solid waste. Overall, waste generation would be lower than all of the SNF
management alternatives, with the exceptions of the No Action and Decentralization Alternatives.
5.14.7 Alternative 5a - Centralization at Other DOE Sites
In general, the character of the impacts to materials and waste management would be similar to
those under Alternative 4b(2).
5.14.8 Alternative 5b - Centralization at the INEL
In general, the character of the impacts to materials and waste management would be similar to
those under Alternative 4b(1).
5.15 Accidents
5.15.1 Introduction
Activities associated with the transportation, receipt, handling, stabilization, and storage of spent
nuclear fuel at the INEL involve substantial quantities of radioactive materials and limited quantities of
toxic chemicals. Under certain circumstances, the potential exists for accidents involving these
materials to occur, which would result in exposure to INEL workers or members of the public, or
contamination of the surrounding environment. Accidents can be categorized as follows:
- Abnormal events such as minor spills
- Design-basis events, which a facility is designed to withstand
- Beyond-design-basis events, which a facility is not designed to withstand (but whose
consequences it may nevertheless mitigate)
This section summarizes postulated radiological and toxic material accidents in each accident
category and describes their estimated consequences to workers, members of the public, and the
environment. The scope of this section is limited to accidents within facilities; transportation
accidents between facilities are addressed in Section 5.11. [Further information on the accidents
summarized in this section, as well as information on other "lower consequence" accidents analyzed, is
provided in Slaughterbeck et al. (1995)].
An accident is a series of unexpected or undesirable "initiating" events that lead to a release of
radioactive or toxic materials within a facility or to the environment. This analysis defines initiating
events that can lead to a spent nuclear fuel-related facility accident in three broad categories: external
initiators, internal initiators, and natural phenomena initiators. External initiators (e.g., aircraft crashes,
and nearby explosions or toxic material releases) originate outside the facility and can affect the ability
of the facility to maintain confinement of radioactive or hazardous material. Internal initiators
originate within a facility (e.g., equipment failures or human error) and are usually the result of facility
operation. Sabotage and terrorist activities (i.e., intentional human initiators) might be either external
or internal initiators. Natural phenomena initiators include weather-related (e.g., floods and tornadoes)
and seismic events. This analysis defines initiators in terms of events that cause, directly or indirectly,
a release of radioactive or hazardous materials within a facility or to the environment by failure or
bypass of confinement.
Tables 5.15-1 through 5.15-4 summarize the radiological results of the analyses described in this
section. Section 5.15.2 summarizes historic accidents at the INEL associated with spent nuclear
fuel-related activities. Section 5.15.3 describes the methodology used to identify and evaluate potential
radiological accidents associated with spent nuclear fuel receipt, handling, storage, and intra-area
transportation activities. Sections 5.15.4 and 5.15.5 evaluate the postulated maximum reasonably
foreseeable radiological and toxic material accidents, respectively.
5.15.2 Historic Perspective
Many of the actions proposed under the different spent nuclear fuel management alternatives
considered in this EIS are continuations or variations of past practices at the INEL. DOE has analyzed
consequences to the public from historic INEL accidents in detail and has determined them to be low
(DOE 1991).
Consequences of accidents can involve fatalities, injuries, or illness. Fatalities can be prompt
(immediate), such as in construction accidents, or latent (delayed), such as cancer caused by radiation
exposure. While public comments received in scoping meetings for this EIS included many concerns
about potential accidents at the INEL, the historic record demonstrates that DOE facilities, including
the INEL, have a very good safety record, particularly in comparison to commercial industries
(e.g., agriculture and construction). Figure 5.15-1 shows the rate of worker fatalities at the INEL and
other DOE sites (DOE 1993b) compared to national-average rates that the National Safety Council
compiled over a 10-year period for various industry groups (NSC 1993) and State of Idaho average
rates (Hendrix 1994). While past accident occurrence rates are not necessarily indicative of future
rates, the historic record reflects the DOE emphasis on safe operations.
There have been no prompt fatalities and no known latent fatalities to members of the public
from accidental releases of radioactive or hazardous materials associated with spent nuclear fuel
management activities in the 40-year history of INEL facilities, although some accidents associated
Table 5.15-1. Summary of radiological accidents for worker located 100 meters downwind from the point of release.
Accident Attribute Alternative 1 Alternative 2 Alternative 3 Alternative 4aa Alternative 5a Alternative 5b
Description No Action Decentralization 1992/1993 Regionalization Centralization Centralization at
Planning Basis by Fuel Type at Other Sites the INEL
1. Fuel handling accident, fuel Consequencesc (d) (d) (d) (d) (d) (d)
pin breach, venting of noble
gases and iodine at HFEFb
Adjusted annual 1.0y10-2 1.2y10-2 3.1y10-2 4.8y10-2 8.6y10-2 2.0y10-1
frequency
Adjusted point (d) (d) (d) (d) (d) (d)
estimate of riske
2. Uncontrolled chain reaction Consequencesc 3.9y10-5 3.9y10-5 3.9y10-5 3.9y10-5 3.9y10-5 3.9y10-5
(criticality) at ICPPf
Adjusted annual 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3
frequency
Adjusted point 4.0y10-8 4.0y10-8 4.0y10-8 4.0y10-8 4.0y10-8 4.0y10-8
estimate of riske
3. Fuel melting of small Consequencesc 2.5y10-4 2.5y10-4 2.5y10-4 2.5y10-4 2.5y10-4 2.5y10-4
number of assemblies at
HFEF resulting from
seismic event and cell breach
Adjusted annual 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5
frequency
Adjusted point 2.5y10-9 2.5y10-9 2.5y10-9 2.5y10-9 2.5y10-9 2.5y10-9
estimate of riske
4. Material release from HFEF Consequencesc 1.8y10-3 1.8y10-3 1.8y10-3 1.8y10-3 1.8y10-3 1.8y10-3
resulting from aircraft crash
and ensuing fire
Adjusted annual 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g
frequency
Adjusted point 1.8y10-10 1.8y10-10 1.8y10-10 1.8y10-10 1.8y10-10 1.8y10-10
estimate of riske
5. Inadvertent nuclear criticality Consequencesc (h) (h) (h) (h) (h) 3.6y10-3
at ICPPf CPP-666 during
processing
Adjusted annual (h) (h) (h) (h) (h) 1.0y10-3
frequency
Adjusted point (h) (h) (h) (h) (h) 3.6y10-6
estimate of riske
6. Hydrogen explosion in ICPPf Consequencesc (h) (h) (h) (h) (h) (d)
CPP-666 dissolver
Adjusted annual (h) (h) (h) (h) (h) (d)
frequency
Adjusted point (h) (h) (h) (h) (h) (d)
estimate of riske
7. Inadvertent dissolution of Consequencesc (h) (h) (h) (h) (h) (d)
30-day cooled fuel at ICPPf
CPP-666
Adjusted annual (h) (h) (h) (h) (h) (d)
frequency
Adjusted point (h) (h) (h) (h) (h) (d)
estimate of riske
a. The radiological accident results for Alternative 4b(1), "Regionalization by Geography (INEL)," are conservatively assumed to be the same as those presented for
Alternative 5b, as discussed in Section 5.15.4.4. The radiological accident results for Alternative 4b(2), "Regionalization by Geography (Elsewhere)," are identical to those
presented for Alternative 5a, as discussed in Section 5.15.4.4.
b. HFEF = Hot Fuel Examination Facility.
c. Consequences are presented in terms of latent fatal cancers based on conservative (95 percentile) meteorological conditions. Consequences are calculated by multiplying the
estimated exposure (i.e., dose) by an International Commission on Radiological Protection conversion factor of 4.0 y 10-4 cancer per rem for an adult worker (or 8.0 y 10-4
cancer per rem if the estimated exposure is greater than 20 rem).
d. The safety analysis report utilized for this accident analysis does not provide this information because it was developed prior to DOE Order 5480.23 requiring this information.
As demonstrated by the dose to the maximally exposed individual, consequences to the public from Accident 1 could be less than the consequences from Accidents 2 through
4. However, given the high frequency for Accident 1 compared to Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
e. This attribute is equal to consequences y frequency (events per year). The information is based on conservative (95 percentile) meteorological conditions.
f. ICPP = Idaho Chemical Processing Plant.
g. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in Section 5.15.6.4.
h. Resuming processing at the INEL under this alternative is not considered.
Table 5.15-2. Summary of radiological accidents for individual located at the nearest point of public access within the site boundary.
Accident Attribute Alternative 1 Alternative 2 Alternative 3 Alternative 4aa Alternative 5a Alternative 5b
Description No Action Decentralization 1992/1993 Regionalization Centralization Centralization at
Planning Basis by Fuel Type at Other Sites the INEL
1. Fuel handling accident, fuel Consequencesc (d) (d) (d) (d) (d) (d)
pin breach, venting of noble
gases and iodine at HFEFb
Adjusted annual 1.0y10-2 1.2y10-2 3.1y10-2 4.8y10-2 8.6y10-2 2.0y10-1
frequency
Adjusted point (d) (d) (d) (d) (d) (d)
estimate of riske
2. Uncontrolled chain reaction Consequencesc 7.0y10-7 7.0y10-7 7.0y10-7 7.0y10-7 7.0y10-7 7.0y10-7
(criticality) at ICPPf
Adjusted annual 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3
frequency
Adjusted point 7.0y10-10 7.0y10-10 7.0y10-10 7.0y10-10 7.0y10-10 7.0y10-10
estimate of riske
3. Fuel melting of small Consequencesc 3.3y10-4 3.3y10-4 3.3y10-4 3.3y10-4 3.3y10-4 3.3y10-4
number of assemblies at
HFEF resulting from
seismic event and cell breach
Adjusted annual 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5
frequency
Adjusted point 3.3y10-9 3.3y10-9 3.3y10-9 3.3y10-9 3.3y10-9 3.3y10-9
estimate of riske
4. Material release from HFEF Consequencesc 1.6y10-4 1.6y10-4 1.6y10-4 1.6y10-4 1.6y10-4 1.6y10-4
resulting from aircraft crash
and ensuing fire
Adjusted annual 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g
frequency
Adjusted point 1.6y10-11 1.6y10-11 1.6y10-11 1.6y10-11 1.6y10-11 1.6y10-11
estimate of riske
5. Inadvertent nuclear criticality Consequencesc (h) (h) (h) (h) (h) 2.5y10-5
ICPPf CPP-666 during
processing
Adjusted annual (h) (h) (h) (h) (h) 1.0y10-3
frequency
Adjusted point (h) (h) (h) (h) (h) 2.5y10-8
estimate of riske
6. Hydrogen explosion in ICPPf Consequencesc (h) (h) (h) (h) (h) (d)
CPP-666 dissolver
Adjusted annual (h) (h) (h) (h) (h) (d)
frequency
Adjusted point (h) (h) (h) (h) (h) (d)
estimate of riske
7. Inadvertent dissolution of Consequencesc (h) (h) (h) (h) (h) (d)
30-day cooled fuel at ICPPf
CPP-666
Adjusted annual (h) (h) (h) (h) (h) (d)
frequency
Adjusted point (h) (h) (h) (h) (h) (d)
estimate of riske
a. The radiological accident results for Alternative 4b(1), "Regionalization by Geography (INEL)," are conservatively assumed to be the
same as those presented for Alternative 5b, as discussed in Section 5.15.4.4. The radiological accident results for Alternative 4b(2),
"Regionalization by Geography (Elsewhere)," are identical to those presented for Alternative 5a, as discussed in Section 5.15.4.4.
b. HFEF = Hot Fuel Examination Facility.
c. Consequences are presented in terms of latent fatal cancers based on conservative (95 percentile) meteorological conditions.
Consequences are calculated by multiplying the estimated exposure (i.e., dose) by an International Commission on Radiological
Protection conversion factor of 5.0 y 10-4 cancer per person-rem for the offsite population (or 1.0 y 10-3 cancer per rem if
the estimated population exposure is greater than 20 rem for any individual member of the public).
d. The safety analysis report utilized for this accident analysis does not provide this information because it was developed prior
to DOE Order 5480.23 requiring this information. As demonstrated by the dose to the maximally exposed individual, consequences to the
public from this accident could be less than the consequences from Accidents 2 through 4. However, given the high frequency for
this accident compared to Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
e. This attribute is equal to consequences y frequency (events per year). The information is based on conservative (95 percentile)
meteorological conditions.
f. ICPP = Idaho Chemical Processing Plant.
g. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in Section 5.15.6.4.
h. Resuming processing at the INEL under this alternative is not considered.
Table 5.15-3. Summary of radiological accidents for maximally exposed hypothetical individual located at the nearest site boundary.
Accident Attribute Alternative 1 Alternative 2 Alternative 3 Alternative 4aa Alternative 5a Alternative 5b
Description No Action Decentralization 1992/1993 Regionalization Centralization Centralization at
Planning Basis by Fuel Type at Other Sites the INEL
1. Fuel handling accident, fuel Consequencesc 1.0y10-6 1.0y10-6 1.0y10-6 1.0y10-6 1.0y10-6 1.0y10-6
pin breach, venting of noble
gases and iodine at HFEFb
Adjusted annual 1.0y10-2 1.2y10-2 3.1y10-2 4.8y10-2 8.6y10-2 2.0y10-1
frequency
Adjusted point 1.0y10-8 1.2y10-8 3.1y10-8 4.8y10-8 8.6y10-8 2.0y10-7
estimate of riskd
2. Uncontrolled chain reaction Consequencesc 5.0y10-7 5.0y10-7 5.0y10-7 5.0y10-7 5.0y10-7 5.0y10-7
(criticality) at ICPPe
Adjusted annual 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3
frequency
Adjusted point 5.0y10-10 5.0y10-10 5.0y10-10 5.0y10-10 5.0y10-10 5.0y10-10
estimate of riskd
3. Fuel melting of small Consequencesc 2.5y10-3 2.5y10-3 2.5y10-3 2.5y10-3 2.5y10-3 2.5y10-3
number of assemblies at
HFEF resulting from
seismic event and cell breach
Adjusted annual 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5
frequency
Adjusted point 2.5y10-8 2.5y10-8 2.5y10-8 2.5y10-8 2.5y10-8 2.5y10-8
estimate of riskd
4. Material release from HFEF Consequencesc 2.5y10-3 2.5y10-3 2.5y10-3 2.5y10-3 2.5y10-3 2.5y10-3
resulting from aircraft crash
and ensuing fire
Adjusted annual 1.0y10-7f 1.0y10-7f 1.0y10-7f 1.0y10-7f 1.0y10-7f 1.0y10-7f
frequency
Adjusted point 2.5y10-10 2.5y10-10 2.5y10-10 2.5y10-10 2.5y10-10 2.5y10-10
estimate of riskd
5. Inadvertent nuclear criticality Consequencesc (g) (g) (g) (g) (g) 1.4y10-5
ICPPe CPP-666 during
processing
Adjusted annual (g) (g) (g) (g) (g) 1.0y10-3
frequency
Adjusted point (g) (g) (g) (g) (g) 1.4y10-8
estimate of riskd
6. Hydrogen explosion in ICPPe Consequencesc (g) (g) (g) (g) (g) 3.2y10-7
CPP-666 dissolver
Adjusted annual (g) (g) (g) (g) (g) 1.0y10-5
frequency
Adjusted point (g) (g) (g) (g) (g) 3.2y10-12
estimate of riskd
7. Inadvertent dissolution of Consequencesc (g) (g) (g) (g) (g) 1.5y10-5
30-day cooled fuel at ICPPe
CPP-666
Adjusted annual (g) (g) (g) (g) (g) 1.0y10-6
frequency
Adjusted point (g) (g) (g) (g) (g) 1.5y10-11
estimate of riskd
a. The radiological accident results for Alternative 4b(1), "Regionalization by Geography (INEL)," are conservatively assumed to be the same as those presented for
Alternative 5b, as discussed in Section 5.15.4.4. The radiological accident results for Alternative 4b(2), "Regionalization by Geography (Elsewhere)," are identical to those
presented for Alternative 5a, as discussed in Section 5.15.4.4.
b. HFEF = Hot Fuel Examination Facility.
c. Consequences are presented in terms of latent fatal cancers based on conservative (95 percentile) meteorological conditions. Consequences are calculated by multiplying the
estimated exposure (i.e., dose) by an International Commission on Radiological Protection conversion factor of 5.0 y 10-4 cancer per person-rem for the offsite population
(or 1.0 y 10-3 cancer per rem if the estimated population exposure is greater than 20 rem for any individual member of the public).
d. This is equal to consequences y frequency (events per year). The information is based on conservative (95 percentile) meteorological conditions.
e. ICPP = Idaho Chemical Processing Plant.
f. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in Section 5.15.6.4.
g. Resuming processing at the INEL under this alternative is not considered.
Table 5.15-4. Summary of radiological accidents for offsite population within 80 kilometers (50 miles) from the point of release.
Accident Attribute Alternative 1 Alternative 2 Alternative 3 Alternative 4aa Alternative 5a Alternative 5b
Description No Action Decentralization 1992/1993 Regionalization Centralization Centralization at
Planning Basis by Fuel Type at Other Sites the INEL
1. Fuel handling accident, fuel Consequencesc (d) (d) (d) (d) (d) (d)
pin breach, venting of noble
gases and iodine at HFEFb
Adjusted annual 1.0y10-2 1.2y10-2 3.1y10-2 4.8y10-2 8.6y10-2 2.0y10-1
frequency
Adjusted point (d) (d) (d) (d) (d) (d)
estimate of riske
2. Uncontrolled chain reaction Consequencesc 3.0y10-4 3.0y10-4 3.0y10-4 3.0y10-4 3.0y10-4 3.0y10-4
(criticality) at ICPPf
Adjusted annual 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3 1.0y10-3
frequency
Adjusted point 3.0y10-7 3.0y10-7 3.0y10-7 3.0y10-7 3.0y10-7 3.0y10-7
estimate of riske
3. Fuel melting of small Consequencesc 7.0y100 7.0y100 7.0y100 7.0y100 7.0y100 7.0y100
number of assemblies at
HFEF resulting from
seismic event and cell breach
Adjusted annual 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5 1.0y10-5
frequency
Adjusted point 7.0y10-5 7.0y10-5 7.0y10-5 7.0y10-5 7.0y10-5 7.0y10-5
estimate of riske
4. Material release from HFEF Consequencesc 1.0y100 1.0y100 1.0y100 1.0y100 1.0y100 1.0y100
resulting from aircraft crash
and ensuing fire
Adjusted annual 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g 1.0y10-7g
frequency
Adjusted point 1.0y10-7 1.0y10-7 1.0y10-7 1.0y10-7 1.0y10-7 1.0y10-7
estimate of riske
5. Inadvertent nuclear criticality Consequencesc (h) (h) (h) (h) (h) 2.8y10-3
ICPPf CPP-666 during
processing
Adjusted annual (h) (h) (h) (h) (h) 1.0y10-3
frequency
Adjusted point (h) (h) (h) (h) (h) 2.8y10-6
estimate of riske
6. Hydrogen explosion in ICPPf Consequencesc (h) (h) (h) (h) (h) 4.1y10-4
CPP-666 dissolver
Adjusted annual (h) (h) (h) (h) (h) 1.0y10-5
frequency
Adjusted point (h) (h) (h) (h) (h) 4.1y10-9
estimate of riske
7. Inadvertent dissolution of Consequencesc (h) (h) (h) (h) (h) 1.5y10-2
30-day cooled fuel at ICPPf
CPP-666
Adjusted annual (h) (h) (h) (h) (h) 1.0y10-6
frequency
Adjusted point (h) (h) (h) (h) (h) 1.5y10-8
estimate of riske
a. The radiological accident results for Alternative 4b(1), "Regionalization by Geography (INEL)," are conservatively assumed to be the same as those presented for
Alternative 5b, as discussed in Section 5.15.4.4. The radiological accident results for Alternative 4b(2), "Regionalization by Geography (Elsewhere)," are identical to those
presented for Alternative 5a, as discussed in Section 5.15.4.4.
b. HFEF = Hot Fuel Examination Facility.
c. Consequences are presented in terms of latent fatal cancers based on conservative (95 percentile) meteorological conditions. Consequences are calculated by multiplying the
estimated exposure (i.e., dose) by an International Commission on Radiological Protection conversion factor of 5.0 y 10-4 cancer per person-rem for the offsite population
(or 1.0 y 10-3 cancer per rem if the estimated population exposure is greater than 20 rem for any individual member of the public).
d. The safety analysis report utilized for this accident analysis does not provide this information because it was developed prior to DOE Order 5480.23 requiring this information.
As demonstrated by the dose to the maximally exposed individual, consequences to the public from this accident could be less than the consequences from Accidents 2 through
4. However, given the high frequency for this accident compared to Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
e. This attribute is equal to consequences y frequency (events per year). The information is based on conservative (95 percentile) meteorological conditions.
f. ICPP = Idaho Chemical Processing Plant.
g. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in Section 5.15.6.4.
h. Resuming processing at the INEL under this alternative is not considered.
Figure 5.15-1. Comparison of fatality rates among workers in various industry groups. with spent nuclear fuel management activities have occurred. In 1958, filters in the Idaho Chemical
Processing Plant CPP-601 Fuel Element Cutting Facility failed during decontamination operations. An
estimated 100 curies of particulate radioactivity were released over an area of approximately 200 acres
(0.809 square kilometers) in the vicinity of the Idaho Chemical Processing Plant. Approximately
39 curies became airborne, resulting in an estimated dose of 0.11 millirem to a hypothetical offsite
individual located at the nearest site boundary (DOE 1991).
Three inadvertent nuclear chain reactions (i.e., nuclear criticalities) occurred at the Idaho
Chemical Processing Plant in 1959, 1961, and 1978. The 1959 criticality occurred in a process waste
and cell floor drain collection tank. Available evidence indicates that the critical solution resulted
from an accidental transfer of concentrated uranyl nitrate solution to the waste collection tank through
a line normally used to transfer decontaminating solutions to the waste tank. The estimated airborne
release from this incident was 3,700 curies, and the estimated dose to the maximally exposed
hypothetical individual located at the nearest site boundary was 1.1 millirem (DOE 1991). The 1961
and 1978 nuclear criticalities resulted from spent nuclear fuel dissolution and reprocessing activities.
Estimated releases to the environment as a result of these accidents were 120 curies and 620 curies for
the 1961 and 1978 accidents, respectively, and the calculated radiation doses at the nearest site
boundary were less than 0.1 millirem for both releases (DOE 1991).
The INEL Fluorinel and Storage (FAST) facility (CPP-666), which historically performed spent
nuclear fuel-related reprocessing activities, is currently shut down. Activities are under way to place
this facility in a permanent shutdown mode. Restart of this facility and the potential for an inadvertent
nuclear criticality resulting from operating this facility are considered in Sections 5.15.4.4 and 5.15.4.5
[Alternatives 4b(1) and 5b, respectively]. Because DOE has no current plans to resume spent nuclear
fuel reprocessing activities at the Idaho Chemical Processing Plant, events similar to the three historic
nuclear criticalities discussed above will be unlikely in future INEL spent nuclear fuel-related
activities. Additional information regarding the historical accidents summarized above is provided in
Slaughterbeck et al. (1995).
In the site's 40-year history, three prompt fatalities of INEL workers have occurred by accidents
involving radiation exposure. In 1961, a steam explosion resulting from an unplanned nuclear
criticality in an experimental reactor (Stationary Low-Power Reactor No. 1) killed these workers, who
were manually moving reactor control elements. The estimated dose from this accident to a
hypothetical individual located at the nearest site boundary was approximately 3 millirem (DOE 1991).
All the accidents discussed above have caused contamination that has led to secondary impacts, such
as the contamination of facility equipment and land inside the site boundary, and have required
cleanup.
Twenty workers at the Argonne National Laboratory-West facility area were injured in early
1994 when, in an accident involving toxic material exposure, approximately 9 kilograms (20 pounds)
of chlorine gas used to treat potable (i.e., drinking) water were accidently released to the environment.
Although an investigation into this incident by the DOE was still ongoing at the time this analysis was
performed, the accident is presumed to have occurred while a vendor was removing and replacing a
nearly empty chlorine cylinder. A maintenance employee assisting in the activity apparently
disconnected the nearly empty in-service chlorine gas cylinder from the potable water system with the
cylinder valve in the open position, resulting in the remaining tank contents being discharged to the
environment. As a result of the accidental release, 20 workers were sent to a local hospital. Eighteen
workers reported for treatment of minor respiratory distress, one worker reported symptoms of more
serious respiratory problems, and one worker reported back injuries as a result of falling while
responding to the accident. (ANL 1994 and DOE 1994b).
5.15.3 Methodology for Determining the Maximum Reasonably Foreseeable Radiological Accidents
5.15.3.1 Selection of Spent Nuclear Fuel Facilities and Operations Requiring
Accident Analyses. The accident analyses performed to support this EIS considered all INEL
nonreactor nuclear facilities that support spent nuclear fuel-related activities with the exception of
those at the Naval Reactors Facility (NRF) area. Appendix D of this EIS discusses each of the spent
nuclear fuel management alternatives and postulated accident scenarios associated with the Naval
Reactors Facility and other naval spent nuclear fuel facilities.
DOE Order 5480.23 (DOE 1992a) defines nonreactor nuclear facilities as those activities or
operations that involve radioactive or fissionable materials in such form and quantity that a nuclear
hazard potentially exists to the workers or the general public. This analysis considered spent nuclear
fuel facilities designed and constructed as direct support to reactor facilities (e.g., Advanced Test
Reactor Storage Canal, which stores spent nuclear fuel and irradiated fuels) as nonreactor spent nuclear
fuel facilities.
DOE manages spent nuclear fuel at the following INEL facility areas: Idaho Chemical
Processing Plant, Naval Reactors Facility, Test Reactor Area, Auxiliary Reactor Area/Power Burst
Facility, Argonne National Laboratory-West, and Test Area North. For further information regarding
the activities conducted in these areas, refer to Chapter 2. After identifying all the nonreactor nuclear
facilities within these facility areas that stabilize, handle, or store spent nuclear fuel, this analysis
ranked the facilities according to potential hazards using preexisting facility "hazard classifications."
DOE Order 5480.23 requires contractors operating nonreactor nuclear facilities to perform a hazard
classification of a facility to assess the consequences of an unmitigated release of radioactive or
hazardous material in one of the following categories(1):
- Category 1. The hazard analysis shows the potential for significant offsite consequences.
- Category 2. The hazard analysis shows the potential for significant onsite consequences.
- Category 3. The hazard analysis shows the potential for only significant localized
consequences.
The classification of nonreactor nuclear facilities in one of these three categories was in
accordance with DOE Standard DOE-STD-1027-92 (DOE 1992b). This standard provides guidance
for the hazard categorization of nuclear facilities based on facility inventories of radionuclides and the
potential for those radionuclides to affect workers or the public if released to the environment.
This analysis used these categories as a screening threshold to identify those facilities of interest
(i.e., those spent nuclear fuel-related facilities with sufficient quantities of radionuclides to present the
potential for significant impacts to workers or the public if released to the environment). The analysis
excluded (screened out) Category 3 (low hazard) facilities if they present possible worker
consequences enveloped by postulated accidents at Category 2 facilities. Facilities with a hazard
classification of 2 or greater (or Category 3 facilities that were not screened out) were evaluated
further, as discussed in the next section.
5.15.3.2 Determination of Maximum Reasonably Foreseeable Radiological
Accidents. After determining spent nuclear fuel-related facilities with sufficient quantities of
radionuclides to present radiological consequences to workers or the public (as discussed in
-----------------------------------------------------------------------------------------------
1. These categories were formerly labeled "high", "moderate," and "low" in accordance
with DOE Order 5480.23 for nonreactor nuclear facilities.
-----------------------------------------------------------------------------------------------
Section 5.15.3.1), the analysis generated potential accident scenarios for each of these INEL facilities
by performing the following activities:
- Reviewing historic spent nuclear fuel-related accidents that have occurred during the 40-year
history of the INEL.
- Reviewing existing accident analyses and safety analysis reports for spent nuclear
fuel-related activities and facilities.
- Identifying potential internal, external, and natural phenomena events that could initiate
spent nuclear fuel-related accidents other than those previously analyzed.
- Performing additional accident analyses for those accidents considered to present the greatest
consequences to workers or the public, as necessary.
The analysis considered internal and external initiators associated with a wide range of activities
(e.g., research and development and construction or modification of facilities) not necessarily covered
in existing safety analyses. For example, potential radiological accident scenarios initiated by
construction activities associated with constructing new spent nuclear fuel-related facilities or
modifying existing spent nuclear fuel-related facilities (as proposed under the various alternatives)
were postulated. Typically, events involved in the construction of new spent nuclear fuel-related
facilities would act as external initiators to existing facilities, while events involved in modifying
existing spent nuclear fuel facilities would act as internal initiators. Examples of construction or
industrial-type events that could initiate a radiological accident included fires, confinement impacts or
puncture events, equipment failure, and human error.
Additional considerations used to determine potential internal and external initiators that could
lead to spent nuclear fuel-related radiological accidents included vulnerabilities associated with
handling, stabilizing, and storing severely degraded spent nuclear fuel and equipment. For example, in
November 1993, DOE issued a report (DOE 1993c) discussing vulnerabilities associated with various
spent nuclear fuel-related facilities across the DOE complex. The report identified one INEL facility,
the CPP-603 Underwater Fuel Storage Facility, as requiring immediate management attention to avoid
unnecessary increases in worker exposures, cleanup costs, and postulated accident frequencies.
Activities have begun to stabilize spent nuclear fuel inventories in the CPP-603 facility and relocate
them to another facility (CPP-666); these activities will continue for several years after the scheduled
1995 Record of Decision for this EIS. Therefore, the analysis considered postulated accident scenarios
associated with stabilizing and relocating CPP-603 spent nuclear fuel inventories to be potential
accident initiators in developing the radiological accidents summarized in this EIS. Examples of
accident scenarios considered as a result of degraded spent nuclear fuel or facility equipment included
inadvertent nuclear criticalities, physical damage of spent nuclear fuel and spent nuclear fuel facilities,
and radionuclide releases resulting from handling and stabilizing degraded spent nuclear fuel. For
postulated accident scenarios at facilities other than the CPP-603 Underwater Fuel Storage Facility, the
analysis also considered the potential for long-term degradation of facility structures, equipment, and
spent nuclear fuel inventories that could lead to an increased probability for radiological accidents.
To compare the various possible spent nuclear fuel-related accident scenarios and to identify
those maximum reasonably foreseeable accidents that present the greatest consequences to workers and
the public, the analysis divided each postulated spent nuclear fuel-related accident into the appropriate
frequency category (abnormal events, design-basis accidents(2), or beyond-design-basis accidents),
according to its estimated frequency of occurrence. Table 5.15-5 lists the frequency ranges associated
with the abnormal event, design-basis accident, and beyond-design-basis accident categories discussed
in Section 5.15.1.
The estimated frequency of each postulated accident was based on an identification of the
physical basis for the accident and the events required for the accident to occur. Because many of the
postulated accidents or their constituent events (initiators or precursors) have rarely or never occurred,
frequency data based on historic experience were not available. Therefore, in many instances, it was
necessary to develop a frequency estimate on the basis of events for which experience existed and
engineering judgment. More than 40 sources of frequency data for the accident events postulated were
reviewed, including analyses and reports prepared for the DOE, U.S. Nuclear Regulatory Commission
(NRC), Electric Power Research Institute, and private industry. [For further information regarding the
development of estimated accident frequencies, refer to Slaughterbeck et al. (1995).]
After the division of the postulated spent nuclear fuel-related accidents into the frequency ranges
defined in Table 5.15-5, the analysis identified the postulated nonprocessing-related accident within
each frequency range determined to present the maximum offsite consequences as a maximum
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2. For facilities where design-basis accident analyses were unavailable, evaluation basis
accident scenarios (postulated accident scenarios used where documented design basis accident
analyses do not exist) were considered in accordance with DOE-DP-STD-3005-YR (DOE 1994a).
---------------------------------------------------------------------------------------------------
Table 5.15-5. Accident frequency categories.
Frequency Category Accident Frequency Range
(accidents per year)
Abnormal events frequency > 1y10-3 per year
Design-basis accidents 1y10-3 per year > frequency > 1y10-6 per year
Beyond-design-basis accidents 1y10-6 per year > frequency > 1y10-7 per year
reasonably foreseeable radiological accident to be further analyzed for this EIS. Potential
nonprocessing-related accident scenarios were chosen as maximum reasonably foreseeable accidents
because of the shutdown status of the INEL facility (CPP-666) that historically processed spent nuclear
fuel. However, because existing inventories of spent nuclear fuel at the INEL would substantially
increase under Alternatives 4b(1) and 5b [Regionalization by Geography (INEL) and Centralization at
the INEL, respectively], there could be a need to resume processing operations to stabilize degraded
spent nuclear fuel operations and assure adequate storage space for spent nuclear fuel received from
other sites(3). Therefore, in addition to the maximum reasonably foreseeable nonprocessing-related
accident scenarios, this analysis considers the three postulated processing-related accidents that present
the maximum offsite consequences as additional maximum reasonably foreseeable accidents under
Alternatives 4b(1) and 5b.
In addition, a postulated inadvertent nuclear criticality accident at the CPP-603 Underwater
Storage Facility was considered for further analysis because significant vulnerabilities associated with
its spent nuclear fuel inventories have been identified (DOE 1993b) and postulated criticality accidents
have been addressed in virtually all nonreactor DOE EISs and safety analysis reports where the
accidents are reasonably foreseeable because of public concerns regarding their potential. As a result,
the seven radiological accidents summarized in Section 5.15.4 were determined to be the maximum
reasonably foreseeable radiological accidents (i.e., greatest consequences). Further discussion and
analysis information for each of these accidents, as well as other accidents analyzed, is provided in
Slaughterbeck et al. (1995). Appendix D identifies maximum reasonably foreseeable accidents
associated with transporting, receiving, handling, and storing naval spent nuclear fuel at the INEL.
The postulated accidents summarized in this section considered with the INEL facilities analyzed in
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3. Processing would be performed in the Flourinel and Storage (FAST) facility (CPP-666) and
a new facility to be constructed, the Fuel Processing Restoration (FPR) facility (CPP-691).
Processing would consist of dissolving spent nuclear fuel to immobilize radionuclides for
final waste disposal.
----------------------------------------------------------------------------------------------------
Appendix D provide a basis for characterizing the potential risks and consequences associated with
managing spent nuclear fuel at the INEL over the next 40 years.
Seismic events were the only identified common-cause initiators with the potential to initiate
radioactive material releases to the environment at more than one spent nuclear fuel-related facility at
the INEL. However, a seismic event resulting in significant damage and radioactive releases from
facilities in more than one facility area (e.g., Idaho Chemical Processing Plant and Test Area North) is
considered beyond reasonably foreseeable (frequency less than one in ten million years), because of
the physical distance and isolation between facility areas. In accordance with DOE guidance (DOE
1994a), a seismic event initiating multiple-facility releases in more than one facility area on the site
was screened from further consideration because of its extremely low frequency of occurrence.
Analyses were performed that evaluated the potential consequences and risks associated with
multiple-facility releases within a single INEL facility area resulting from a severe seismic event
(Slaughterbeck et al. 1995). For example, within a 500-meter radius in the Idaho Chemical Processing
Plant facility area, there are several spent nuclear fuel facilities, the primary facilities being the CPP-
749 dry storage facilities and the CPP-666 and CPP-603 underwater fuel storage facilities. An
analysis was performed (Slaughterbeck et al. 1995) to determine whether simultaneous releases from
these facilities could result from a severe seismic event. Because the CPP-666 and CPP-749 facilities
were designed and qualified to withstand a severe seismic event, they are not expected to contribute to
the consequences and risks resulting from a severe seismic event impacting the Idaho Chemical
Processing Plant. However, because of known structural deficiencies and vulnerabilities with the spent
nuclear fuel at the CPP-603 facility, the CPP-603 facility is expected to be significantly damaged
following a severe seismic event, resulting in one or more criticalities and the leakage of contaminated
basin water to the surrounding environment. While the consequences from these simultaneous
multiple-release mechanisms (one or more criticalities and water drainage) would be greater than the
single criticality analyzed for CPP-603 facility (Section 5.15.3.3.2), the consequences and risk of such
releases are expected to be bounded by the other accidents analyzed in the EIS--primarily, a seismic
event that causes fuel melting at the Argonne National Laboratory-West Hot Fuel Examination Facility
(highest consequence accident), and a fuel handling accident in the same facility (highest risk accident,
where risk = consequence x frequency). Similar analyses (DOE 1993a) for the Test Area North and
Argonne National Laboratory-West also demonstrate that potential multiple-facility releases or
multiple-release mechanisms from a single facility resulting from a severe seismic event would also be
bounded by accidents postulated for the Hot Fuel Examination Facility. Based on this conclusion and
the accident selection methodology described 5.15.3.1, the consequences and risks associated with
multiple-facility releases were screened from further consideration since they do not represent the
bounding accident scenarios within the frequency categories defined in Table 5.15-5.
In addition, the screening methodology did not specifically include potential accident scenarios
associated with operating new spent nuclear fuel handling and storage facilities proposed under the
various alternatives considered in this EIS because postulated accident scenarios for existing facilities
would bound the consequences associated with potential accidents at new facilities. This assumption
is appropriate for two primary reasons. First, the missions of new spent nuclear fuel facilities would
be similar to the missions of existing spent nuclear fuel-related DOE facilities, which implies that
DOE would consider the same types of accident scenarios for the new facilities it considered for the
existing facilities. Second, DOE would design and build new facilities that would incorporate modern
preventive and mitigative features to reduce the frequency and potential consequences associated with
postulated accidents.
To compare the consequences of the same accident scenario at an identical hypothetical facility
constructed at each DOE site included in this EIS (based on local geological and meteorological
conditions), Appendix D summarizes postulated accident scenarios for a new Expended Core Facility
at Oak Ridge, Hanford Site, Savannah River Site, or Nevada Test Site.
To determine the radiological and toxicological consequences presented throughout Section 5.15
associated with the postulated accidents and with spent nuclear fuel-related activities, the analysis used
the following definitions:
- Worker. An individual 100 meters (328 feet) downwind of the facility location where the
release occurs.4
- Nearest Public Access. The nearest point of public access to the location where the release
occurs, sometimes inside the site boundary.
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4. The worker is defined as the individual located at 100 meters because reliable safety analyses
quantifying the impacts (e.g., dose and health effects) to workers at distances less than 100
(i.e., "close-in" workers) meters fram an accidental release of radionuclides are unavailable.
The effects on and risks to workers closer in than 100 meters are recognized and discussed in
Section 5.15.3.3. Each of the maximum reasonably forseeable accidents considered in this EIS,
particularly the design-basis and beyond-design-basis accidents, contains some risk of worker injury
or death at distances closer than 100 meters.
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- Maximally Exposed Offsite Individual. A hypothetical resident at the site boundary nearest
to the facility where the release occurs.
- Offsite Population. The collective total of individuals within an 80-kilometer (50-mile)
radius of the INEL.
- Environment. The area outward from 100 meters (328 feet) downwind of the facility where
the release occurs.
5.15.3.3 Impact of Accidents on Close-In Workers. An evaluation has been made on the
radiological impact to close-in workers from the selected accident scenarios. Injuries or fatalities that
might occur due to an external event, such as a severe seismic disturbance or airplane crash into the
structure, are not considered in this evaluation since they are not attributable to direct radiological
consequences. Seven accident scenarios for nonprocessing-related and processing-related activities are
considered maximum reasonably foreseeable accidents.
5.15.3.3.1 Mechanical Handling Accident at the Argonne National Laboratory
West Hot Fuel Examination Facility - This accident is assumed to result in fuel pin breach and
venting of noble gases and iodine.
No fatalities to workers are expected from this event. However, a
substantial iodine dose to the thyroid could cause radiation-induced hypothyroidism or a similar
disorder.
5.15.3.3.2 Criticality Accident at the Idaho Chemical Processing Plant -
CPP-603 - This event is an unplanned nuclear criticality associated with underwater spent nuclear
fuel storage at the CPP-603 facility.
Based on shielding provided by the pool water, it is likely that
no fatalities would occur. To the extent water is expelled due to the energy of the event, close-in
workers could receive substantial radiation exposure. Worker presence in the area above the pool or
very close to the edge of the pool is not routine. The impact of the event would likely be isolated to
nearby equipment operators if the criticality were initiated by a handling error.
5.15.3.3.3 Seismic Event Leading to Fuel Melt at the Argonne National
Laboratory West Hot Fuel Examination Facility - A seismic event is postulated to result in a
breech of the main cell used for examination of the fuel, which is assumed to lead to a failure of the
fuel cooling system.
It is likely that the release of radioactive materials from fuel melting would occur
slowly enough to allow evacuation of all workers before any appreciable exposure. Therefore, no
radiation-induced fatalities would be expected.
5.15.3.3.4 Airplane Crash and Fire at Argonne National Laboratory West Hot
Fuel Examination Facility - An airplane crash and subsequent fire sustained by airplane fuel
could result in a major breach of the confinement barriers and could lead to a substantial atmospheric
release of radionuclides.
Workers unaffected by the airplane crash or fire would not be expected to
remain in the area long enough to receive substantial radiation exposure. It is assumed the buoyancy
of the radioactive material due to the fire would mitigate the direct radiological impacts to close-in
workers, substantially reducing the likelihood of radiation induced worker fatalities.
5.15.3.3.5 Criticality Accident During Processing at the Idaho Chemical
Processing Plant - CPP-666 - This is the first of three evaluated accidents that could occur only
if processing were resumed at the Fluorinel and Storage Facility (FAST).
Three inadvertent nuclear
criticalities have occurred in INEL processing facilities and none has resulted in worker fatalities. In
each event, radioactive material was released to the atmosphere and close-in workers received direct
exposure. If processing were resumed, the techniques and controls implemented to prevent recurrence
of processing-related criticalities would be employed again. Due to the cell wall shielding provided by
concrete walls that are several feet thick, it is expected that no workers would receive substantial
radiation exposure.
5.15.3.3.6 Hydrogen Explosion at the Idaho Chemical Processing Plant - A
hydrogen explosion in the dissolver off-gas system of the Flourinel and Storage (FAST) Facility would
result in release of radioactive material to the facility.
If workers were near the dissolver off-gas
system, they could receive substantial radiation exposure from the explosion. No fatalities would be
expected, but radiation-induced health detriments could occur.
5.15.3.3.7 Dissolution of Short-Cooled Fuel at the Idaho Chemical Processing
Plant - An explosion in the dissolver tank could occur if fuel that has not cooled for at least 30 days
was inadvertently shipped to the dissolver at the Flourinel and Storage Facility (FAST).
This energetic
event would likely breach the dissolver off gas system and could breach the dissolver tank. Workers
in the areas closely associated with the dissolver tank could receive substantial radiation exposure, but
it is likely that no radiation-induced fatalities would occur.
5.15.3.4 Analysis of Radiological Accident Consequences. The quantities of
radioactive materials and the ways these materials interact with human beings are important factors in
determining health effects. The ways in which radioactive materials reach human beings, their
absorption and retention in the body, and the resulting health effects have been studied in great detail.
The International Commission on Radiological Protection (ICRP) has made specific recommendations
for quantifying 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 can be classified into two categories: prompt (also referred to as acute) and
latent. Prompt health effects are those experienced immediately after exposure and include damage to
the body up to and including death. Latent health effects are those experienced some time after
exposure and include cancers and hereditary symptoms. An INEL-developed computer code,
Radiological Safety Analysis Computer Program-5 (RSAC-5), estimates potential radiation doses to
maximally exposed individuals or population groups from accidental releases of radionuclides. This
code, which is customized to specific INEL conditions, uses well-established and generally accepted
scientific engineering principles as the basis for its various calculational steps. The code is based on
guidance provided in NRC Guide 1.145 (NRC 1983) and has been validated to comply with accepted
standards for such software. [For a detailed description of RSAC-5, refer to Slaughterbeck et al.
(1995).]
The RSAC-5 code determined estimated consequences to the worker, an individual assumed to
be stranded at the nearest point of public access, the maximally exposed hypothetical individual at the
nearest site boundary, and the offsite population within 80 kilometers (50 miles) of the radiological
accidents postulated under Alternative 1, No Action. Postulated frequencies and consequences
analyzed under Alternative 1 are based on (1) the approximate amount of spent nuclear fuel currently
at the INEL [measured in Metric Tons Heavy Metal (MTHM)], (2) the estimated increases in
inventories resulting from spent nuclear fuel generated by operating INEL reactors (i.e., fuel recently
removed from a reactor that has not had sufficient time to cool), and (3) the estimated number of fuel
handling activities associated with stabilizing or relocating spent fuel inventories inside the INEL site
boundary. Although the four nonprocessing-related maximum reasonably foreseeable radiological
accident scenarios identified for Alternative 1 are also considered under Alternatives 2 through 5,
proposed changes in INEL spent nuclear fuel inventories and the number of fuel handling activities
associated with these changes could affect the estimated frequencies and consequences expected for
Alternatives 2 through 5. Therefore, to reasonably estimate the frequencies and consequences
associated with activities proposed under Alternatives 2 through 5, the frequencies and consequences
for the accidents presented under Alternative 1 require appropriate "adjustment" or "scaling."
To be conservative, the analysis assumed that the increase in the annual frequency of mechanical
handling accidents would be equal to the estimated increase in the annual number of handling events
proposed under Alternatives 2 through 5. However, the consequences associated with a mechanical
handling accident would not vary with a change in the number of handling events because the amount
of material involved in each event would not change. To determine potential changes in annual
mechanical handling accident frequencies between the different spent nuclear fuel management
alternatives, the analysis based its estimates of the annual number of fuel handling events under each
alternative on spent fuel shipment rates anticipated for the next 40 years, as discussed in Appendix I.
Estimates of long-term (40-year) and short-term (5-year) shipments at the INEL were considered in
determining the annual shipment rates for each alternative. The basis for the number of long-term
shipments include spent nuclear fuel the INEL will continue to receive from operating reactors such as
DOE, Naval Nuclear Propulsion Program, university, and research reactors. Short-term shipments
consist of shipments that would be required to relocate existing spent fuel inventories between sites
under the various alternatives. Table 5.15-6 summarizes the estimated annual shipment rate to and
from the INEL under each alternative, and within INEL site boundaries. The estimates provided in
Table 5.15-6 consider both onsite and offsite shipments.
Table 5.15-6. Determination of accident frequency adjustment factors for Alternatives 2 through 5
based on estimated number of annual spent nuclear fuel shipments under each alternative.
Alternative Estimated Shipment Adjustment Factor
Rate (per year)a (shipment
rate/baseline)
1. No Action 41 Baseline
2. Decentralization 50 1.2
3. 1992/1993 Planning Basis 128 3.1
4a. Regionalization by Fuel Type 195 4.8
4b(1) Regionalization by Geography (INEL) 824 20.0
4b(2) Regionalization by Geography 351 8.6
(Elsewhere)
5a. Centralization at Other DOE Sites 351 8.6
5b. Centralization at the INEL 824 20.0
a. Data presented for the estimated annual shipment rate is based on information tabulated in
Appendix I. The annual shipment rate for the No-Action Alternative (baseline) is derived from
Table 3 of Wichmann 1994.
Based on the number of annual shipments estimated for Alternatives 2 through 5, as listed in
Table 5.15-6, the analysis calculated multiplication factors by dividing the estimated shipment rates
under Alternatives 2 through 5 by the baseline (Alternative 1) shipment rate. To determine the
estimated frequency for the maximum reasonably foreseeable mechanical handling accidents under
each alternative, the frequency identified for Alternative 1 was multiplied by the appropriate
adjustment factor. The same approach determined estimated frequencies for Accident 1 (fuel pin
breach and noble gases and iodine release from the Hot Fuel Examination Facility) under
Alternatives 2 through 5. For Accident 2 (inadvertent criticality in the CPP-603 Underwater Fuel
Storage Facility resulting from a handling accident associated with degraded spent nuclear fuel), the
estimated frequency considered under Alternative 1 (1 y 10-3 event per year) is based on the number of
handling activities associated with relocation of the CPP-603 spent nuclear fuel inventories to the
CPP-666 facility. Because proposed changes in INEL inventories under the different alternatives
would not affect handling events associated with relocating spent fuel from the CPP-603 facility to the
CPP-666 facility, the estimated frequency for this mechanical handling event would not change. As a
result of this approach and the fact that 3 of the 4 accident scenarios that present the greatest
consequences are not handling accidents, Accident 1 is the only accident requiring "adjustment" for
each alternative.
Variable source-term-sensitive accidents would have consequences that depended on the amount of
spent nuclear fuel in storage. One example is the accidental drainage of a spent fuel storage canal that
results in the release of corrosion products in the canal to the environment. The larger the spent fuel
inventory in the canal, the larger the release of corrosion products to the environment resulting from
draining the canal. (Drainage of a water canal completely filled with spent nuclear fuel was
considered in the determination of the maximum reasonably foreseeable accidents and was determined
to present lower consequences than other accident scenarios analyzed.) Variable source-term sensitive
accidents depend only on spent nuclear fuel inventories and do not require adjustment of their
estimated frequencies of occurrence. Because none of the postulated accidents summarized under
Alternative 1 is source-term sensitive (e.g., spent nuclear fuel inventories in the Hot Fuel Examination
Facility are not likely to increase), adjustment of the estimated consequences calculated under
Alternative 1 is not required for Alternatives 2 through 5.
5.15.4 Impacts from Postulated Maximum Reasonably Foreseeable Radiological Accidents
Section 5.15.4.1 summarizes impacts (e.g., exposures and health effects) from the four
nonprocessing-related maximum reasonably foreseeable radiological accidents postulated under
Alternative 1 (No Action). Sections 5.15.4.4.2.1 through 5.15.4.5.2 describe changes in these
postulated accident impacts resulting from changes in spent nuclear fuel inventories and handling
activities under the other alternatives. Sections 5.15.4.4.2.1 and 5.15.4.5.2 also summarize impacts
from three additional maximum reasonably foreseeable accidents associated with resumption of
processing activities at the INEL. Section 5.15.6 provides more information about the assumptions
and analyses performed for each of the radiological accidents discussed under each alternative.
5.15.4.1 Alternative 1: No Action. Based on the quantity of spent nuclear fuel at the INEL
(excluding naval fuel at Naval Reactors Facility, which is analyzed in Appendix D), its storage
configuration (wet versus dry), the amount of time the spent fuel has been allowed to cool, and
consideration of various internal, external, and natural phenomena initiators (as discussed in
Section 5.15.3), the postulated accidents listed in Table 5.15-7 would have the greatest radiological
consequences within the abnormal event, design-basis accident, and beyond-design-accident categories
under this alternative. For each accident, Table 5.15-7 also lists estimated accident frequencies;
radiation exposures to the offsite population within 80 kilometers (50 miles), a member of the public
stranded at the nearest point of public access inside the INEL site boundary, a hypothetical maximally
exposed individual (MEI) at the nearest site boundary, and a worker; point estimates of the annualized
risk of the maximally exposed individual contracting a fatal cancer during his/her lifetime as a result
of the radiation exposure; and point estimates of risk of the expected number of fatal cancers
(annualized and total) in the offsite population. The estimates of the consequences and risk to the
offsite population are based on conservative (95 percentile) and average (50 percentile) meteorological
conditions(5). The estimates of the consequences and risk to the maximally exposed individual are
based on conservative (95 percentile) meteorological conditions. The postulated accidents listed in
Table 5.15-7, in conjunction with the maximum reasonably foreseeable spent nuclear fuel accidents
identified for the INEL Naval Reactors Facility in Appendix D, characterize the potential consequences
and risks associated with the proposed spent fuel management activities at the INEL under this
alternative.
Atmospheric transport of radionuclides from the postulated accidents could result in some
secondary impacts, such as contamination of the environment or impacts to national defense. To
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5. Conservative (95 percentile) meteorological conditions are defined as the
meteorlogical conditions that, for a given release, the concentration at a fixed
receptor location will not be exceeded 95 percent of the time. Average (50 percentile)
meteorological conditions are defined as the meteorological conditions that, for a
given release, the concentration at a fixed receptor location will not be exceeded
50 percent of the time.
--------------------------------------------------------------------------------------------
Table 5.15-7. Impacts from selected maximum reasonably foreseeable radiological accidents -
Alternative 1, No Action (50 and 95 percentile meteorological conditions).
Accident Frequency Worker Nearest Dose to Offsite Point estimates of risk of fatal cancers
(events per Dosea Public MEIc Population (per year)
year) (rem) Accessb (rem) Dose (95%)
(rem) (person-rem)
MEI Offsite Population
95%d 50% 95%
1. Fuel handling
accident, fuel pin
breach, venting of 1.0y10-2 (f) (f) 2.0y10-3 (f) 1.0y10-8 (f) (f)
noble gases and
iodine at HFEFe
2. Inadvertent criticality 6.5y10-9 3.0y10-7
in ICPPg CPP-603 1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 (6.5y10-6)d (3.0y10-4)d
storage facilityh
3. Fuel melting of small
number of assemblies 4.5y10-7 7.0y10-5
at HFEF resulting 1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 (4.5y10-2)d (7.0y100)d
from seismic event
and cell breach
4. Material release from 3.6y10-8 1.0y10-7
HFEF resulting from1.0y10-7(i) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 (3.6y10-1)d (1.0y100)d
aircraft crash and
ensuing fire
a. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
b. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
c. MEI = Maximally exposed hypothetical offsite individual, located at the nearest site boundary.
d. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y
5.0 y 10-4 fatal cancer per rem (ICRP-60 conversion factor) if dose is less than 20 rem. For doses 20 rem or
more the ICRP-60 conversion factor is doubled, or 1.0 y 10-3. Numbers in parentheses indicate the total
number of fatal cancers in the population if the accident occurred.
e. HFEF - Hot Fuel Examination Facility.
f. The safety analysis report utilized for this accident analysis does not provide this information because it was
developed prior to DOE Order 5480.23 requiring this information. As demonstrated by the dose to the
maximally exposed individual, consequences to the public from this accident could be less than the
consequences from Accidents 2 through 4.
g. ICPP = Idaho Chemical Processing Plant.
h. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred at
the INEL during its 40-year operating history, the estimated frequency for an inadvertent criticality is not
based on historic reprocessing data because reprocessing is not considered under this alternative. Nominal
frequency estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y 10-3 (CPP-603
underwater storage facility) event per year.
i. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in
Section 5.15.6.4.
prevent these radionuclides from increasing any potential safety concerns, DOE would initiate cleanup
activities if an accident occurred, and no irreversible environmental impacts would be likely.
Table 5.15-8 summarizes postulated secondary impacts resulting from the postulated radiological
accidents listed in Table 5.15-7.
This analysis takes limited credit for emergency response actions in determining the consequences
listed in Table 5.15-7. DOE would initiate INEL emergency response programs, as appropriate,
following the occurrence of an accident to prevent or mitigate potential consequences. These
emergency response programs, implemented in accordance with 5500-DOE series Orders, typically
involve emergency planning, emergency preparedness, and emergency response actions. Each
emergency response plan utilizes resources specifically dedicated to assist a facility in emergency
management. These resources include but are not limited to the following:
- INEL Warning Communications Center
- INEL Fire Department
- Facility Emergency Command Centers
- DOE Emergency Operations Centers
- County and State Emergency Command Centers
- Medical, health physics, and industrial hygiene specialists
- Protective clothing and equipment (respirators, breathing air supplies, etc.)
- Periodic training exercises and drills within and between the organizations involved in
implementing the response plans
5.15.4.2 Alternative 2: Decentralization. Adjustments in estimated accident frequencies
and point estimates of risk presented for Alternative 1 would be related to (1) the receipt, handling,
and storage activities associated with the additional spent nuclear fuel inventories; and (2) the increase
in overall spent nuclear fuel-related storage, relocation, and handling activities not allowed under
Alternative 1. Because no changes in the accident consequences estimated for Alternative 1 are likely
to occur under this alternative from increased fuel inventories (i.e., the same amount of radioactive
material would accidentally be released to the environment as discussed in Section 5.15.3.3), no
changes are likely in the postulated secondary impacts listed in Table 5-15-8. Table 5.15-9
summarizes the four postulated accidents with the greatest radiological impacts under this alternative.
Table 5.15-8. Estimated secondary impacts resulting from the maximum reasonably foreseeable accidents postulated under Alternative 1, No
Action, assuming conservative (95 percentile) meteorological conditions.
Environmental or Social Impacts
Radiological (Assuming 88 millirem per year limit with 24-hour-per-day exposure)a
Accident
Summary
Biotic Water Economic National Environmental Endangered Land Treaty Rights &
Resources Resources Impacts Defense Contamination Species Use Tribal Resources
1. Fuel handling Limited adverse Limited adverse Limited economic No effects on Local No impacts No change in land No irreversible
accident, fuel effects expected effects expected to impacts expected. national defense contamination exptected to use or irreversible impacts to Native
pin breach, vegetation or surface water or Any cleanup expected. requiring cleanup endangered or impacts expected. Americans or
venting of wildlife. groundwater. required would be expected around threatened species. public lands
noble gases and localized and site accident. expected.
iodine at could be
HFEFb (1x10-2 accomplished with
per year) existing workforce
and equipment.
2. Uncontrolled Limited adverse Limited adverse No economic No effects on Local No impacts No change in land No irreversible
chain reaction effects expected effects expected to impacts expected. national defense contamination exptected to use or irreversible impacts to Native
(criticality) at vegetation or surface water or Any cleanup expected. requiring cleanup endangered or impacts expected. American or
ICPPc (1x10-3 wildlife. groundwater. required would be expected around threatened species. public lands
per year) localized and site accident. expected.
could be
accomplished with
existing workforce
and equipment.
3. Fuel melting of Limited adverse Limited adverse Potential No effects on Local No impacts Potential for Potential for
small number effects expected effects expected to interdiction of national defense contamination exptected to 1 year of temporary
of assemblies at vegetation or surface water or affected expected. requiring cleanup endangered or agricultural land restricted access
HFEF resulting wildlife. groundwater. agricultural expected around threatened species. withdrawal of up to affected public
from seismic products on site accident. to 10,000 acresd land (less than
event and cell nearby lands. (on and off the 10,000 acres).d
breach (1x10-5 Local cleanup in INEL site).
per year) the vicinity of
HFEF.
4. Material release Limited adverse Limited adverse Potential No effects on Local No impacts Potential for Potential for
from HFEF effects expected effects expected to interdiction of national defense contamination exptected to 1 year of temporary
resulting from vegetation or surface water or affected expected. requiring cleanup endangered or agricultural restricted access
aircraft crash wildlife. groundwater. agricultural expected around threatened species. withdrawal of up to affected public
and ensuing products on site accident. to 10,000 acresd land (less than
fire (1x10-7 per nearby lands. (on and off the 10,000 acres).d
year) Local cleanup in INEL site).
the vicinity of
HFEF.
a. Postulated secondary impacts based on 10-microrem-per-hour exposure (88 millirem per year with 24-hour-per-day exposure) from ground contamination resulting from radionuclide deposition
from the plume. This approach in estimated secondary impacts is conservative because DOE Order 5400.5 states that the public dose limit for exposure to residual contamination and natural
background radiation is 100 millirem per year.
b. HFEF = Hot Fuel Examination Facility.
c. ICPP = Idaho Chemical Processing Plant.
d. To convert acres to square kilometers, multiply by 0.004.
Table 5.15-9. Impacts from selected maximum reasonably foreseeable accidents - Alternative 2,
Decentralization (50 and 95 percentile meteorological conditions).
Accident Adjusted Worker Nearest Dose to Offsite Adjusted point estimates of risk of fatal
Frequencya Doseb Public MEId Population cancers (per year)
(events per (rem) Accessc (rem) Dose
year) (rem) (95%)
(person-
rem)
MEI Offsite Population
95%e 50% 95%
1. Fuel handling accident,
fuel pin breach, 1.2y10-2 (g) (g) 2.0y10-3 (g) 1.2y10-8 (g) (g)
venting of noble gas(1.2)
and iodine at HFEFf
2. Inadvertent criticality 6.5y10-9 3.0y10-7
in ICPPh CPP-603 1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 (6.5y10-6)e (3.0y10-4)e
storage facilityi (1.0)j
3. Fuel melting of small
number of assemblies 4.5y10-7 7.0y10-5
at HFEF resulting fr1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 (4.5y10-2)e (7.0y100)e
seismic event and ce(1.0)
breach
4. Material release from
HFEF resulting from 1.0y10-7(k) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 3.6y10-8 1.0y10-7
aircraft crash and (1.0) (3.6y10-1)e (1.0y100)e
ensuing fire
a. Numbers in parentheses indicate multiplication factor used to scale or adjust estimated accident frequencies
under Alternative 1, as described in Section 5.15.3.3.
b. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
c. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
d. MEI = Maximally exposed hypothetical offsite individual located at the nearest site boundary.
e. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y
5.0 y 10-4 fatal cancer per rem (ICRP-60 conversion factor) if dose is less than 20 rem. For doses of 20 rem
or more, the ICRP-60 conversion factor is doubled, or 1.0 y 10-3. Numbers in parentheses indicate total
number of fatal cancers in the population if the accident occurs.
f. HFEF = Hot Fuel Examination Facility.
g. The safety analysis report utilized for this accident analysis does not provide this information because it was
developed prior to DOE Order 5480.23 requiring this information. As demonstrated by the dose to the
maximally exposed individual, consequences to the public from this accident could be less than the
consequences from Accidents 2 through 4.
h. ICPP = Idaho Chemical Processing Plant.
i. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred at
the INEL during its 40-year operating history, the estimated frequency for an inadvertent criticality is not
based on historic reprocessing data since reprocessing is not considered under this alternative. Nominal
frequency estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y 10-3 (CPP-603
underwater storage facility) events per year.
j. Refer to Sections 5.15.3.3 and 5.15.6.2 for details on why this frequency was not adjusted under this
alternative.
k. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in
Section 5.15.6.4.
5.15.4.3 Alternative 3: 1992/1993 Planning Basis. Under this alternative, the INEL could
receive the following spent nuclear fuel:
. Spent nuclear fuel from domestic DOE and university reactors and foreign research test
reactors
- All Training Reactor Isotopics General Atomics (TRIGA) spent nuclear fuel from foreign
and Hanford reactors
- Fort St. Vrain spent nuclear fuel from Public Service Company of Colorado
- Special case commercial pressurized water reactor and boiling water reactor spent nuclear
fuel from West Valley, New York
- Naval spent nuclear fuel from sites such as the Norfolk or Puget Sound Naval Shipyard.
Adjustments in estimated accident frequencies and point estimates of risk presented for
Alternative 1 would be related to (1) the receipt, handling, and storage activities associated with the
additional spent nuclear fuel inventories; and (2) the increase in overall spent fuel-related storage,
relocation, and handling activities not allowed under Alternative 1. Because no changes in the
accident consequences estimated for Alternative 1 are likely to occur under this alternative from
increased fuel inventories (i.e., the same amount of radioactive material would accidentally be released
to the environment as discussed in Section 5.15.3.3), no changes are likely in the postulated secondary
impacts listed in Table 5.15-8. Table 5.15-10 summarizes the postulated accidents with the greatest
radiological impacts under this alternative.
5.15.4.4 Alternative 4: Regionalization. Under this alternative, there are two primary
Regionalization alternatives: (1) Alternative 4a (Regionalization by Fuel Type), where existing and
spent nuclear fuel inventories will be distributed between the DOE sites based primarily on the
similarity of fuel types, although DOE would also consider transportation distances, available
stabilization capabilities, available storage capacities, or a combination of these factors; or
(2) Alternative 4b (Regionalization by Geography), where existing and new spent nuclear fuel
inventories in the western region of the country will be centralized at a single western site, and
existing and new spent nuclear fuel inventories in the eastern region of the country will be centralized
at a single eastern site.
Table 5.15-10. Impacts from selected maximum reasonably foreseeable accidents - Alternative 3,
Planning Basis (50 and 95 percentile meteorological conditions).
Accident Adjusted Worker Nearest Dose to Offsite Adjusted point estimates of risk of fatal
Frequencya Doseb Public MEId Population cancers (per year)
(events per (rem) Accessc (rem) Dose (95%)
year) (rem) (person-rem)
MEI Offsite Population
95%e 50% 95%
1. Fuel handling
accident, fuel pin
breach, venting of 3.1y10-2 (g) (g) 2.0y10-3 (g) 3.1y10-8 (g) (g)
noble gases and (3.1)
iodine at HFEFf
2. Inadvertent critica1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 6.5y10-9 3.0y10-7
in ICPPh CPP-603 (1.0)j (6.5y10-6)e (3.0y10-4)e
storage facilityi
3. Fuel melting of small
number of assemblies 4.5y10-7 7.0y10-5
at HFEF resulting 1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 (4.5y10-2)e (7.0y100)e
from seismic event (1.0)
and cell breach
4. Material release from
HFEF resulting from1.0y10-7(k) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 3.6y10-8 1.0y10-7
aircraft crash and (1.0) (3.6y10-1)e (1.0y100)e
ensuing fire
a. Numbers in parentheses indicate multiplication factor used to scale or adjust estimated accident frequencies
under Alternative 1, as described in Section 5.15.3.3.
b. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
c. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
d. MEI = Maximally exposed hypothetical offsite individual located at the nearest site boundary.
e. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y
5.0 y 10-4 fatal cancer per rem (ICRP-60 conversion factor) if dose is less than 20 rem. For doses of 20 rem
or more, the ICRP-60 conversion factor is doubled, or 1.0 y 10-3. Numbers in parentheses indicate total
number of fatal cancers in the population if the accident occurs.
f. HFEF = Hot Fuel Examination Facility.
g. The safety analysis report utilized for this accident analysis does not provide this information because it was
developed prior to DOE Order 5480.23 requiring this information. As demonstrated by the dose to the
maximally exposed individual, consequences to the public from this accident could be less than the
consequences from Accidents 2 through 4. However, given the high frequency for this accident compared to
Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
h. ICPP = Idaho Chemical Processing Plant.
i. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred at
the INEL during its 40-year operating history, the estimated frequency for an inadvertent criticality is not
based on historic reprocessing data since reprocessing is not considered under this alternative. Nominal
frequency estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y 10-3 (CPP-603
underwater storage facility) events per year.
j. Refer to Sections 5.15.3.3 and 5.15.6.2 for details on why this frequency was not adjusted under this
alternative.
k. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in
Section 5.15.6.4.
5.15.4.4.1 Alternative 4a - Regionalization By Fuel Type - Adjustments in the estimated
accident frequencies and point estimates of risk presented for Alternative 1 would be related to (1) the
receipt, handling, and storage activities associated with the additional spent nuclear fuel inventories;
and (2) the increase in overall spent nuclear fuel-related storage, relocation, and handling activities not
allowed under Alternative 1.
Because no changes in the accident consequences estimated for
Alternative 1 are likely to occur under this alternative from increased fuel inventories (i.e., the same
amount of radioactive material would accidentally be released to the environment as discussed in
Section 5.15.3.3), no changes are likely in the postulated secondary impacts listed in Table 5.15-8.
Table 5.15-11 summarizes the postulated accidents with the greatest radiological impacts under this
alternative.
5.15.4.4.2 Alternative 4b - Regionalization by Geography - Under this alternative, spent
nuclear fuel inventories in the western region of the country would be centralized at either the INEL,
Hanford Site, or Nevada Test Site.
Alternative 4b(1) considers regionalization at the INEL.
Alternative 4b(2) considers regionalization at the Hanford Site or Nevada Test Site.
5.15.4.4.2.1 Alternative 4b(1) - Regionalization by Geography (INEL) - Under
this alternative, existing and new spent nuclear fuel inventories in the western region of the country
would be centralized at the INEL. Fuel stabilization would be performed in the Fluorinel and Storage
(FAST) facility (CPP-666) and a new facility to be constructed, the Fuel Processing Restoration
facility (CPP-691), to dissolve spent nuclear fuel and stabilize (i.e., immobilize) radionuclides.
Because the volume of spent nuclear fuel considered under this alternative is only slightly lower than
that considered under Alternative 5b, adjustments in the estimated accident frequencies and point
estimates of risk for the four accidents presented under Alternative 1 were conservatively considered
equivalent to the adjustments required under Alternative 5b (i.e., centralization of all the DOE, Naval
Nuclear Propulsion Program, university, and research reactor spent nuclear fuel in the country at the
INEL). Adjustments in the estimated accident frequencies and point estimates of risk for the four
accidents presented under Alternative 1 would be related to (1) the receipt, handling, and storage
activities associated with the additional spent nuclear fuel inventories; and (2) the increase in overall
spent nuclear fuel-related storage, relocation, and handling activities not allowed under Alternative 1.
Because no changes in the accident consequences estimated for Alternative 1 are likely to occur under
this alternative from increased fuel inventories (i.e., the same amount of radioactive material would
accidentally be released to the environment as discussed in Section 5.15.3.3), no changes are likely in
the postulated secondary impacts listed in Table 5.15-8.
Table 5.15-11. Impacts from selected maximum reasonably foreseeable accidents - Alternative 4a,
Regionalization by Fuel Type (50 and 95 percentile meteorological conditions).
Accident Adjusted Worker Nearest Dose to Offsite Adjusted point estimates of risk of fatal
Frequencya Doseb Public MEId Population cancers (per year)
(events per (rem) Accessc (rem) Dose (95%)
year) (rem) (person-rem)
MEI Offsite Population
95%e 50% 95%
1. Fuel handling
accident, fuel pin
breach, venting of4.8y10-2 (g) (g) 2.0y10-3 (g) 4.8y10-8 (g) (g)
noble gases and (4.8)
iodine at HFEFf
2. Inadvertent 6.5y10-9 3.0y10-7
criticality in ICP1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 (6.5y10-6)e (3.0y10-4)e
CPP-603 storage (1.0)j
facilityi
3. Fuel melting of
small number of 4.5y10-7 7.0y10-5
assemblies at HFEF1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 (4.5y10-2)e (7.0y100)e
resulting from (1.0)
seismic event and
cell breach
4. Material release
from HFEF resultin1.0y10-7(k) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 3.6y10-8 1.0y10-7
from aircraft cras(1.0) (3.6y10-1)e (1.0y100)e
and ensuing fire
a. Numbers in parentheses indicate multiplication factor used to scale or adjust estimated accident frequencies
under Alternative 1, as described in Section 5.15.3.3.
b. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
c. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
d. MEI = Maximally exposed hypothetical offsite individual located at the nearest site boundary.
e. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y
5.0 y 10-4 fatal cancer per rem (ICRP-60 conversion factor) if dose is less than 20 rem. For doses of 20 rem
or more, the ICRP-60 conversion factor is doubled, or 1.0 y 10-3. Numbers in parentheses indicate total
number of fatal cancers in the population if the accident occurs.
f. HFEF = Hot Fuel Examination Facility.
g. The safety analysis report utilized for this accident analysis does not provide this information because it was
developed prior to DOE Order 5480.23 requiring this information. As demonstrated by the dose to the
maximally exposed individual, consequences to the public from this accident could be less than the
consequences from Accidents 2 through 4. However, given the high frequency for this accident compared to
Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
h. ICPP = Idaho Chemical Processing Plant.
i. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred at
the INEL during its 40-year operating history, the estimated frequency for an inadvertent criticality is not
based on historic reprocessing data since reprocessing is not considered under this alternative. Nominal
frequency estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y 10-3 (CPP-603
underwater storage facility) events per year.
j. Refer to Sections 5.15.3.3 and 5.15.6.2 for details on why this frequency was not adjusted under this
alternative.
k. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in
Section 5.15.6.4.
Because the option exists to restart processing activities, three additional processing-related
maximum reasonably foreseeable accidents are considered under this alternative (as discussed in
Section 5.15.3.2). Since the amount of radioactive material that would accidentally be released to the
environment from these accidents is expected to be lower than in Accidents 3 and 4 (i.e., small fuel
melt and aircraft crash at the Hot Fuel Examination Facility, respectively), potential secondary impacts
associated with these additional processing-related accidents would be less severe than those presented
for the nonprocessing-related accidents in Table 5.15-8.
Table 5.15-12 summarizes the postulated accidents with the greatest radiological impacts under this
alternative.
5.15.4.4.2.2 Alternative 4b(2) - Regionalization by Geography (Elsewhere) - Under this
alternative, existing and new spent nuclear fuel inventories in the western region of the country would
be centralized at either the Hanford Site or Nevada Test Site. Similar to Alternative 5a, which
considers centralization of existing INEL spent nuclear fuel inventories at another DOE site, the
inventory of spent nuclear fuel at the INEL would be reduced substantially so that the only spent
nuclear fuel at the INEL would consist of fresh fuel generated from operating INEL reactors that had
not cooled sufficiently for relocation to the regionalized or centralized site. Therefore, this alternative
considers the same amount of material considered under Alternative 1 until the regionalized site could
accept existing inventories of INEL spent nuclear fuel and freshly generated spent nuclear fuel that has
sufficiently cooled.
Table 5.15-13 summarizes the postulated accidents with the greatest radiological impacts under this
alternative.
5.15.4.5 Alternative 5: Centralization. Under this alternative, DOE would collect all
current and future spent nuclear fuel inventories from both DOE and the Naval Nuclear Propulsion
Program at one site. For the INEL, there are two possibilities: (1) Alternative 5a, in which most
spent fuel inventories and activities would take place at the Hanford Site, Savannah River Site, Nevada
Test Site, or Oak Ridge Reservation; or (2) Alternative 5b, in which all spent fuel inventories and
activities would be centralized at the INEL.
5.15.4.5.1 Alternative 5a: Centralization at Other DOE Sites - This alternative
would consider approximately the same amount of material considered under Alternative 1 until the
centralized site could accept existing INEL spent nuclear fuel inventories and freshly generated spent
Table 5.
15-12. Impacts from selected maximum reasonably foreseeable accidents - Alternative 4b(1),
Regionalization by Geography (INEL) (50 and 95 percentile meteorological conditions).
Accident Adjusted Worker Nearest Dose to Offsite Adjusted point estimates of risk of fatal
Frequencya Doseb Public MEId Population cancers (per year)
(events per (rem) Accessc (rem) Dose
year) (rem) (95%)
(person-
rem)
MEI Offsite Population
95%e 50% 95%
1. Fuel handling
accident, fuel pin 2.0y10-1 (g) (g) 2.0y10-3 (g) 2.0y10-7 (g) (g)
breach, venting of (20.0)
noble gases and
iodine at HFEFf
2. Inadvertent critica1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 6.5y10-9 3.0y10-7
in ICPPh CPP-603 (1.0)j (6.5y10-6)e (3.0y10-4)e
storage facilityi
3. Fuel melting of small
number of assemblie1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 4.5y10-7 7.0y10-5
at HFEF resulting (1.0) (4.5y10-2)e (7.0y100)e
from seismic event
and cell breach
4. Material release from
HFEF resulting from1.0y10-7(k) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 3.6y10-8 1.0y10-7
aircraft crash and (1.0) (3.6y10-1)e (1.0y100)e
ensuing fire
5. Inadvertent nuclear
criticality ICPPh 1.0y10-3 9.1y10+ 4.9y10-2 2.8y10-2 5.6y10+0 1.4y10-8 3.1y10-6 2.8y10-6
CPP-666 during 0 (3.1y10-3) (2.8y10-3)
processingl
6. Hydrogen in ICPPh 1.0y10-5 (m) (m) 6.3y10-4 8.1y10-1 3.2y10-12 (m) 4.1y10-9
CPP-666 dissolver (4.1y10-4)
7. Inadvertent
dissolution of 30-d1.0y10-6 (m) (m) 3.0y10-2 2.9y10+1 1.5y10-11 (m) 1.5y10-8
cooled fuel at ICPPh (1.5y10-8)
CPP-666
a. Numbers in parentheses indicate multiplication factor used to scale or adjust estimated accident frequencies under Alternative 1, as
described in Section 5.15.3.3.
b. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
c. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
d. MEI = Maximally exposed hypothetical offsite individual located at the nearest site boundary.
e. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y 5.0 y 10-4 fatal cancer per rem
(ICRP-60 conversion factor) if dose is less than 20 rem. For doses of 20 rem or more, the ICRP-60 conversion factor is doubled, or
1.0 y 10-3. Numbers in parentheses indicate total number of fatal cancers in the population if the accident occurs.
f. HFEF = Hot Fuel Examination Facility.
g. The safety analysis report utilized for this accident analysis does not provide this information because it was developed prior to DOE
Order 5480.23 requiring this information. As demonstrated by the dose to the maximally exposed individual, consequences to the public
from Accident 1 could be less than the consequences from Accidents 2 through 4. However, given the high frequency for Accident 1
compared to Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
h. ICPP = Idaho Chemical Processing Plant.
i. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred during the 40-year
operating history of CPP-666, the estimated frequency for an inadvertent criticality in this facility is based on existing spent
nuclear conditions and fuel vulnerabilities. Nominal estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y
10-3 (CPP-603 underwater storage facility) events per year.
j. Refer to Sections 5.15.3.3 and 5.15.6.2 for details on why this frequency was not adjusted under this alternative.
k. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in Section 5.15.6.4.
l. The Idaho Chemical Processing Plant has experienced three inadvertent nuclear criticalities during its operating history, the last one
14 years ago. This frequency is based on modern facility conditions and safeguards that exist at CPP-666.
m. The safety analysis report utilized for this accident does not provide this information because it was developed prior to DOE
Order 5480.23 requiring this information. However, a comparison of the data presented for this accident to the other accidents provides
a relative measure of the impacts to this receptor.
Table 5.15-13. Impacts from selected maximum reasonably foreseeable accidents - Alternative 4b(2),
Regionalization by Geography (Elsewhere) (50 and 95 percentile meteorological conditions).
Accident Adjusted Worker Nearest Dose to Offsite Adjusted point estimates of risk of fatal
Frequencya Doseb Public MEId Population cancers (per year)
(events per (rem) Accessc (rem) Dose (95%)
year) (rem) (person-rem)
MEI Offsite Population
95%e 50% 95%
1. Fuel handling
accident, fuel pin 8.6y10-2 (g) (g) 2.0y10-3 (g) 8.6y10-8 (g) (g)
breach, venting of (8.6)
noble gases and
iodine at HFEFf
2. Inadvertent critica1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 6.5y10-9 3.0y10-7
in ICPPh CPP-603 (1.0)j (6.5y10-6)e (3.0y10-4)e
storage facilityi
3. Fuel melting of small
number of assemblie1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 4.5y10-7 7.0y10-5
at HFEF resulting (1.0) (4.5y10-2)e (7.0y100)e
from seismic event
and cell breach
4. Material release from
HFEF resulting from1.0y10-7(k) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 3.6y10-8 1.0y10-7
aircraft crash and (1.0) (3.6y10-1)e (1.0y100)e
ensuing fire
a. Numbers in parentheses indicate multiplication factor used to scale or adjust estimated accident frequencies
under Alternative 1, as described in Section 5.15.3.3.
b. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
c. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
d. MEI = Maximally exposed hypothetical offsite individual located at the nearest site boundary.
e. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y
5.0 y 10-4 fatal cancer per rem (ICRP-60 conversion factor) if dose is less than 20 rem. For doses of 20 rem
or more, the ICRP-60 conversion factor is doubled, or 1.0 y 10-3. Numbers in parentheses indicate total
number of fatal cancers in the population if the accident occurs.
f. HFEF = Hot Fuel Examination Facility.
g. The safety analysis report utilized for this accident analysis does not provide this information because it was
developed prior to DOE Order 5480.23 requiring this information. As demonstrated by the dose to the
maximally exposed individual, consequences to the public from this accident could be less than the
consequences from Accidents 2 through 4. However, given the high frequency for this accident compared to
Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
h. ICPP = Idaho Chemical Processing Plant.
i. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred at
the INEL during its 40-year operating history, the estimated frequency for an inadvertent criticality is not
based on historic reprocessing data since reprocessing is not considered under this alternative. Nominal
frequency estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y 10-3 (CPP-603
underwater storage facility) events per year.
j. Refer to Sections 5.15.3.3 and 5.15.6.2 for details on why this frequency was not adjusted under this
alternative.
k. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in
Section 5.15.6.4.
fuel that had cooled sufficiently. On demonstration of the centralized site's capability to receive INEL
spent nuclear fuel, the inventory of spent fuel at the INEL would be reduced substantially so that the
only spent nuclear fuel at the INEL would consist of fresh fuel generated from operating INEL
reactors that had not cooled sufficiently for relocation to the centralized site.
Adjustments in estimated accident frequencies and point estimates of risk presented for
Alternative 1 would be related to (1) the receipt, handling, and storage activities associated with the
additional spent nuclear fuel inventories; and (2) the increase in overall spent fuel-related storage,
relocation, and handling activities not allowed under Alternative 1. Because no changes in the
accident consequences estimated for Alternative 1 are likely to occur under this alternative from
increased fuel inventories (i.e., the same amount of radioactive material would accidentally be released
to the environment as discussed in Section 5.15.3.3), no changes are likely in the postulated secondary
impacts presented in Table 5.15-8. Table 5.15-14 summarizes the postulated accidents with the
greatest radiological impacts under these alternatives.
5.15.4.5.2 Alternative 5b: Centralization at the INEL - Adjustments in estimated
accident frequencies and point estimates of risk presented for Alternative 1 would be related to (1) the
receipt, handling, and storage activities associated with the additional spent nuclear fuel inventories;
and (2) the increase in overall spent nuclear fuel-related storage, relocation, and handling activities not
allowed under Alternative 1.
Because no changes in the accident consequences estimated for
Alternative 1 are likely to occur under this alternative from increased fuel inventories (i.e., the same
amount of radioactive material would accidentally be released to the environment as discussed in
Section 5.15.3.3), no changes are likely in the postulated secondary impacts presented in Table 5.15-8.
Table 5.15-15 summarizes the postulated accidents with the greatest radiological impacts under this
alternative.
Because the option exists to restart processing activities, three additional processing-related
maximum reasonably foreseeable accidents are considered under this alternative (as discussed in
Section 5.15.3.2). Since the amount of radioactive material that would accidentally be released to the
environment from these accidents is expected to be lower than Accidents 3 and 4 (i.e., small fuel melt
and aircraft crash at the Hot Fuel Examination Facility, respectively), potential secondary impacts
associated with these additional processing-related accidents would be less severe than those presented
for the nonprocessing-related accidents in Table 5.15-8.
Table 5.15-14. Impacts from selected maximum reasonably foreseeable accidents - Alternative 5a,
Centralization at Other DOE Sites (50 and 95 percentile meteorological conditions).
Accident Adjusted Worker Nearest Dose to Offsite Adjusted point estimates of risk of fatal
Frequencya Doseb Public MEId Population cancers (per year)
(events per (rem) Accessc (rem) Dose (95%)
year) (rem) (person-rem)
MEI Offsite Population
95%e 50% 95%
1. Fuel handling
accident, fuel pin 8.6y10-2 (g) (g) 2.0y10-3 (g) 8.6y10-8 (g) (g)
breach, venting of (8.6)
noble gases and
iodine at HFEFf
2. Inadvertent critica1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 6.5y10-9 3.0y10-7
in ICPPh CPP-603 (1.0)j (6.5y10-6)e (3.0y10-4)e
storage facilityi
3. Fuel melting of small
number of assemblie1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 4.5y10-7 7.0y10-5
at HFEF resulting (1.0) (4.5y10-2)e (7.0y100)e
from seismic event
and cell breach
4. Material release from
HFEF resulting from1.0y10-7(k) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 3.6y10-8 1.0y10-7
aircraft crash and (1.0) (3.6y10-1)e (1.0y100)e
ensuing fire
a. Numbers in parentheses indicate multiplication factor used to scale or adjust estimated accident frequencies
under Alternative 1, as described in Section 5.15.3.3.
b. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
c. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
d. MEI = Maximally exposed hypothetical offsite individual located at the nearest site boundary.
e. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y
5.0 y 10-4 fatal cancer per rem (ICRP-60 conversion factor) if dose is less than 20 rem. For doses of 20 rem
or more, the ICRP-60 conversion factor is doubled, or 1.0 y 10-3. Numbers in parentheses indicate total
number of fatal cancers in the population if the accident occurs.
f. HFEF = Hot Fuel Examination Facility.
g. The safety analysis report utilized for this accident analysis does not provide this information because it was
developed prior to DOE Order 5480.23 requiring this information. As demonstrated by the dose to the
maximally exposed individual, consequences to the public from this accident could be less than the
consequences from Accidents 2 through 4. However, given the high frequency for this accident compared to
Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
h. ICPP = Idaho Chemical Processing Plant.
i. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred at
the INEL during its 40-year operating history, the estimated frequency for an inadvertent criticality is not
based on historic reprocessing data since reprocessing is not considered under this alternative. Nominal
frequency estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y 10-3 (CPP-603
underwater storage facility) events per year.
j. Refer to Sections 5.15.3.3 and 5.15.6.2 for details on why this frequency was not adjusted under this
alternative.
k. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in
Section 5.15.6.4.
Table 5.15-15. Impacts from selected maximum reasonably foreseeable accidents - Alternative 5b,
Centralization at the INEL (50 and 95 percentile meteorological conditions).
Accident Adjusted Worker Nearest Dose to Offsite Adjusted point estimates of risk of fatal
Frequencya Doseb Public MEId Population cancers (per year)
(events per (rem) Accessc (rem) Dose
year) (rem) (95%)
(person-
rem)
MEI Offsite Population
95%e 50% 95%
1. Fuel handling
accident, fuel pin (g)
breach, venting of2.0y10-1 (g) (g) 2.0y10-3 (g) 2.0y10-7 (g)
noble gases and (20.0)
iodine at HFEFf
2. Inadvertent 6.5y10-9 3.0y10-7
criticality in ICP1.0y10-3 9.7y10-2 1.4y10-3 1.0y10-3 5.9y10-1 5.0y10-10 (6.5y10-6)e (3.0y10-4)e
storage facilityi (1.0)j
3. Fuel melting of
small number of 4.5y10-7 7.0y10-5
assemblies at HFEF1.0y10-5 6.2y10-1 6.5y10-1 5.0y100 1.4y104 2.5y10-8 (4.5y10-2)e (7.0y100)e
resulting from (1.0)
seismic event and
cell breach
4. Material release
from HFEF resultin1.0y10-7(k) 4.6y100 3.2y10-1 5.0y100 2.0y103 2.5y10-10 3.6y10-8 1.0y10-7
from aircraft cras(1.0) (3.6y10-1)e (1.0y100)e
and ensuing fire
5. Inadvertent nuclear
criticality ICPPh 1.0y10-3 9.1y10+0 4.9y10-2 2.8y10-2 5.6y10+0 1.4y10-8 3.1y10-6 2.8y10-6
CPP-666 during (3.1y10-3) (2.8y10-3)
processingl
6. Hydrogen in ICPPh 1.0y10-5 (m) (m) 6.3y10-4 8.1y10-1 3.2y10-12 (m) 4.1y10-9
CPP-666 dissolver (4.1y10-4)
7. Inadvertent
dissolution of 30-1.0y10-6 (m) (m) 3.0y10-2 2.9y10+1 1.5y10-11 (m) 1.5y10-8
day cooled fuel at (1.5y10-2)
ICPPh CPP-666
a. Numbers in parentheses indicate multiplication factor used to scale or adjust estimated accident frequencies under Alternative 1, as
described in Section 5.15.3.3.
b. A worker is defined as a worker located 100 meters (328 feet) from the point of release.
c. Public individual assumed to be stranded at the nearest point of public access inside the site boundary.
d. MEI = Maximally exposed hypothetical offsite individual located at the nearest site boundary.
e. Maximally exposed individual and offsite population fatal cancer risk = dose y accident frequency y 5.0 y 10-4 fatal cancer per rem
(ICRP-60 conversion factor) if dose is less than 20 rem. For doses of 20 rem or more, the ICRP-60 conversion factor is doubled, or
1.0 y 10-3. Numbers in parentheses indicate total number of fatal cancers in the population if the accident occurs.
f. HFEF = Hot Fuel Examination Facility.
g. The safety analysis report utilized for this accident analysis does not provide this information because it was developed prior to DOE
Orders requiring this information. As demonstrated by the dose to the maximally exposed individual, consequences to the public from
this accident could be less than the consequences from Accidents 2 through 4. However, given the high frequency for this accident
compared to Accidents 2 through 4, the risk could actually be greater than for Accidents 2 through 4.
h. ICPP = Idaho Chemical Processing Plant.
i. Although three nuclear criticalities associated with spent nuclear fuel reprocessing activities have occurred during the 40-year
operating history of CPP-666, the estimated frequency for an inadvertent criticality in this facility is based on existing spent
nuclear conditions and fuel vulnerabilities. Nominal estimates vary from 1.0 y 10-4 (CPP-666 underwater storage facility) to 1.0 y
10-3 (CPP-603 underwater storage facility) events per year.
j. Refer to Sections 5.15.3.3 and 5.15.6.2 for details on why this frequency was not adjusted under this alternative.
k. This frequency is a qualitative bounding estimate for a potential aircraft crash, as discussed in Section 5.15.6.4.
l. The Idaho Chemical Processing Plant has experienced three inadvertent nuclear criticalities during its operating history, the last one
14 years ago. This frequency is based on modern facility conditions and safeguards that exist at CPP-666.
m. The safety analysis report utilized for this accident does not provide this information because it was developed prior to DOE
Order 5480.23 requiring this information. However, a comparison of the data presented for this accident to the other accidents
provides a relative measure of the impacts to this receptor.
5.15.5 Impacts from Postulated Maximum Reasonably Foreseeable Toxic Material Accidents
Like radioactive materials, toxic materials (e.g., chemicals) are involved in a variety of
operations, including spent nuclear fuel-related activities, at the INEL. As a result of these operations
and activities, the potential exists for releases of toxic materials to the environment from the same
types of initiators considered in determining the radiological accident scenarios discussed in
Section 5.15.4. This section summarizes analyses of postulated accident scenarios associated with
spent nuclear fuel activities that could result in the release of toxic materials from their confinements.
5.15.5.1 Identification of Toxic Chemicals at the INEL. The facilities at the INEL use
many types and quantities of chemically toxic materials. To determine the spent fuel-related chemicals
that exist in sufficient quantities to present health effects to workers or the offsite population, DOE
performed an initial screening of the chemical inventories at the INEL. This screening consisted of
identifying those hazardous chemicals at the INEL listed in the Superfund Amendments and
Reauthorization Act of 1986 (SARA) 312 Report for 1992 (Priestly 1992) that (1) exist in bulk
quantities [assumed to be greater than 227 kilograms (500 pounds)]; or (2) exceed reportable quantities
[usually 0.45 kilogram (1 pound)] on the EPA Title III List of Lists (EPA 1990), which includes
hazardous chemicals defined in the following:
- SARA Section 302, Extremely Hazardous Substances (40 CFR Part 355, Appendixes A and
B, List of Extremely Hazardous Substances and Their Threshold Planning Quantities)
(CFR 1993)
- Comprehensive Environmental Response, Compensation, and Liability Act Hazardous
Substances (40 CFR Part 302, Table 302.4, Lists of Hazardous Substances and Reportable
Quantities) (CFR 1992a)
- SARA Section 313, Toxic Chemicals (CFR 1992b)
- Federal Register list of 100 extremely hazardous chemicals (FR 1994)
5.15.5.2 Selection of Spent Nuclear Fuel-Related Toxic Chemicals Requiring
Accident Analysis. As indicated by the screening methodology discussed above, toxic chemical
inventories are located throughout INEL facilities in varying quantities and are involved in nearly all
operations and activities performed by INEL facilities, including spent nuclear fuel-related activities.
The screening identified no toxic chemicals associated with the dry storage of spent nuclear fuel.
Except for processing-related activities that could be performed under the Regionalization and
Centralization at INEL alternatives [i.e., Alternatives 4b(1) and 5b, respectively], the screening
identified activities associated with the underwater storage of spent nuclear fuel (e.g., maintaining
water chemistry) as the only spent nuclear-fuel related activities that might utilize toxic chemicals in
sufficient quantities to present a potential for health effects to workers or the offsite population, or
potential contamination of the environment. For Alternatives 4b(2) and 5a, in which DOE would
relocate INEL spent nuclear fuel inventories and related activities to other DOE sites, the existing toxic
chemical inventories at the INEL would be expected to slightly decrease. For Alternatives 4b(1) and
5b, in which the INEL could potentially resume processing activities, a substantial increase in existing
chemical inventories, primarily hydrofluoric acid and anhydrous ammonia, would be expected. No
substantial changes in existing spent nuclear fuel-related toxic chemical inventories would be expected
under Alternatives 1, 2, or 3.
To demonstrate how the consequences of the same accident at an identical hypothetical facility
constructed at the Hanford Site or the Savannah River Site under this alternative would compare to the
INEL (based on local geological and meteorological conditions), Appendix D summarizes postulated
accident scenarios for a new Expended Core Facility that DOE could construct at any of the sites
considered in this EIS.
To determine potential accident scenarios associated with handling or storing toxic chemicals at
the various spent nuclear fuel-related facilities, DOE performed an extensive review of existing safety
analyses and walkdowns of various facilities. This review identified two nonprocessing-related toxic
chemicals at the Idaho Chemical Processing Plant - nitric acid and chlorine - as requiring further
evaluation to determine potential health effects to workers and the offsite population. Additionally,
two toxic chemicals that would be required to support the resumption of processing activities at the
Idaho Chemical Processing Plant - hydrofluoric acid and anhydrous ammonia - were identified as
requiring further evaluation(6). Although spent fuel-related facilities at the Idaho Chemical Processing
Plant use several other toxic chemicals (e.g., oxalic acid), the quantities of these chemicals are not
sufficient to present an impact to workers or the environment from accidental releases to the
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6. Although bulk quantities of nitric acid would be required to perform processing activities that
could be resumed Alternatives 4b(1) and 5b, the consequences of processing-related accidents
involving nitric acid would be bounded by the hydrofluoric acid and anhydrous accidents analyzed in
Sections 5.15.3.3. and 5.15.3.4., respectively. Therefore, this analysis focuses on a potential
nitric acid accident resulting from the nonprocessing spent nuclear fuel-related activities
considered under the other alternatives.
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environment. (For postulated accident scenarios involving Naval spent nuclear fuel-related activities at
the INEL, refer to Appendix D.)
Because DOE determined that it needed to evaluate postulated toxic chemical accidents at the
Idaho Chemical Processing Plant as part of this EIS, it did not consider postulated toxic chemical
accidents at the Advanced Test Reactor Storage Canal and the Hot Fuel Examination Facility that
could be involved in spent fuel-related activities(7) for further evaluation in this EIS for the following
reasons:
- In general, quantities of spent nuclear fuel-related chemicals at the Idaho Chemical
Processing Plant are substantially greater than those at the Advanced Test Reactor Storage
Canal and Hot Fuel Examination Facility.
- The Idaho Chemical Processing Plant is located approximately 1,000 meters (1,094 yards)
closer to the nearest site boundary than the Advanced Test Reactor.
Based on a review of safety documentation for the Test Area North spent nuclear fuel underwater
storage facility and discussions with facility personnel, DOE determined that none of the toxic
chemicals identified in the screening (Section 5.15.5.1) is related to spent fuel handling or storage
activities.
5.15.5.3 Toxic Chemical Accident Analysis. For chemically toxic materials, several
government agencies recommend quantifying health effects that cause short-term effects as threshold
values of concentrations in air or water. The long-term health consequences of human exposure to
toxic materials are not as well understood as the long-term health consequences related to radiation
exposure. Thus, the potential health effects for exposures to toxic chemicals are more subjective than
those for radioactive materials. Factors such as receptor locations, terrain, meteorological conditions,
release conditions, and characteristics of chemical inventories are required parameters for
determinations of airborne concentrations of toxic chemicals at various distances from a postulated
point of release.
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7. The scope of this analysis has been restricted to the Advanced Test Reactor fuel storage
canal. Everything inside the reactor gas-tight boundary and associated with reactor operations
has been excluded from consideration because reactor operations are not related to the spent
nuclear fual activities considered in this EIS.
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EPICodeTM was used to estimate airborne concentrations resulting from spent nuclear fuel-related
toxic chemical releases at the INEL. [For a detailed description of EPICodeTM, refer to Slaughterbeck
et al. (1995).]
To determine the potential health effects from accidental releases of toxic chemicals, this analysis
compared the concentrations determined by EPICodeTM against Emergency Response Planning
Guideline values, where available. These values, which are specific for each substance, are related to
three general severity levels:
- Exposure to concentrations greater than Emergency Response Planning Guideline-1 values
for a period of time greater than 1 hour results in an unacceptable likelihood that a person
would experience mild transient adverse health effects, or perception of a clearly defined
objectionable odor.
- Exposure to concentrations greater than Emergency Response Planning Guideline-2 values
for a period of time greater than 1 hour results in an unacceptable likelihood that a person
would experience or develop irreversible or other serious health effects, or symptoms that
could impair one's ability to take protective action.
- Exposure to concentrations greater than Emergency Response Planning Guideline-3 values
for a period of time greater than 1 hour results in an unacceptable likelihood that a person
would experience or develop life-threatening health effects.
If there were no Emergency Response Planning Guideline values for a toxic substance, the
analysis substituted other chemical toxicity values, as follows:
- Threshold limit values/time-weighted average values (ACGIH 1988) substituted for
Emergency Response Planning Guideline-1. This is the time-weighted average concentration
for a normal 8-hour workday and a 40-hour workweek to which nearly all workers could be
repeatedly exposed, day after day, without adverse effect.
- Level of concern values (equal to 0.1 of the immediately dangerous to life or health values -
see below) substituted for Emergency Response Planning Guideline-2. The level of concern
value is the concentration of a hazardous substance in the air above which there might be
serious irreversible health effects or death as a result of a single exposure for a relatively
short period of time.
- Immediately dangerous to life or health values are substituted for Emergency Response
Planning Guideline-3. The immediately dangerous to life or health value is the maximum
concentration from which a person could escape within 30 minutes without a respirator and
without experiencing any impairment of escape or irreversible side effects (NIOSH 1990).
As stated in the above section, four toxic chemicals - chlorine, nitric acid, hydrofluoric acid,
and anhydrous ammonia - at the Idaho Chemical Processing Plant were identified as requiring further
evaluation to estimate potential health effects to workers and the public. The following sections
summarize the analyses performed for these chemicals.
5.15.5.3.1 Accidental Chlorine Release - Chlorine, while not directly associated with
spent nuclear fuel-related activities at the INEL, is used to treat drinking water supplies at the various
spent fuel facilities.
Therefore, an analysis of a postulated accidental chlorine release at the Idaho
Chemical Processing Plant was performed to determine potential impacts on workers operating the
spent fuel-related facilities.
At the Idaho Chemical Processing Plant, chlorine is contained in two pressurized bottles
[65 atmospheres at 20yC (68yF)], a 68-kilogram (150-pound) bottle and a 55-kilogram
(120-pound) bottle, totaling 123 kilograms (270 pounds). To be conservative, DOE assumed that a
breach of the drain line causes an instantaneous release of the total inventory of both tanks. The
highest chlorine concentrations at the receptor locations would result from the largest release over the
shortest time period. Therefore, the release duration was assumed to be approximately 5 minutes.
An accidental chlorine release from one of the chlorine tanks could be initiated by one of several
events, such as a handling event, piping or valve rupture, or human error. Because the two tanks are
physically separated, an accidental simultaneous release from both tanks would require a common
initiator such as a delivery accident, a common maintenance failure, or a natural phenomena event
(e.g., seismic) that damaged or punctured both tanks. The frequency of an accidental release from one
pressurized tank is 1.0 y 10-4 event per year (EPA/FEMA/DOT 1987). A common cause failure
resulting in the release of chlorine from two separated tanks is assumed to be no greater than 5 percent
of the time given for the first tank failure. Therefore, the estimated frequency of an accidental release
from both tanks is 5.0 y 10-6 events per year (with no credit taken for pressure vessel management and
training).
Table 5.15-16 summarizes the concentrations of the subject chlorine release at the following
receptor locations: a facility worker, a member of the public stranded at the nearest point of public
access inside the INEL boundary, and a maximally exposed hypothetical member of the public located
at the nearest site boundary. As listed in Table 5.15-10, the peak chlorine concentrations for facility
workers could result in life-threatening health effects (i.e., Emergency Response Planning Guideline-3
values are exceeded) for both conservative (95 percentile) and average (50 percentile) meteorological
conditions.
Table 5.15-16. Summary of chemical concentrations for postulated nonprocessing-related accidental
releases at the Idaho Chemical Processing Plant under Alternatives 1 through 5.
Chemical Concentrations
Receptor Location (milligrams per cubic meter)a
95% Meteorologyb 50% Meteorologyc
Chlorine Nitric Acide Chlorine Nitric Acide
ERPG-1d = 3 (1) TWA = 5.2 (2) ERPG-1 = 3 (1) TWA = 5.2 (2)
ERPG-2 = 9 (3) LOC = 25.5 (10) ERPG-2 = 9 (3) LOC = 25.5 (10)
ERPG-3 = 60 (20) IDLH = 255 (100) ERPG-3 = 60 (20) IDLH = 255 (100)
1. Worker located at 84,000 250 1,620 33
100 meters (325 feet). (28,000) (95) (540) (13)
2. Nearest point of public
access where a member 19.5 0.32 1.89 0.049
of the public is (6.5) (0.12) (0.63) (0.019)
assumed stranded at the
time of the release.f
3. Maximally exposed
hypothetical individual 4.2 0.12 0.42 0.016
located at the nearest (1.4) (0.047) (0.14) (0.006)
site boundary.g
a. Numbers in parentheses reflect concentrations in parts per million.
b. The 95 percentile meteorology is based on Class F (unfavorable) meteorological conditions with 0.5 meter per
second (1.1 miles per hour) wind speed for receptors located within 2 kilometers (1.2 miles) of the release
and 2 meters per second (4.5 miles per hour) for receptors beyond 2 kilometers of the release.
c. The 50 percentile meteorology is based on Class D (typical) meteorological conditions with 4.5 meters per
second (10 miles per hour) wind speed for all receptors.
d. ERPG = Emergency Response Planning Guidelines.
e. Because Emergency Response Planning Guideline values are not available for nitric acid, time-weighted
average values are substituted for ERPG-1 values, level of concern values are substituted for ERPG-2 values,
and immediately dangerous to life or health values are substituted for Emergency Response Planning
Guideline-3 values. Refer to Section 5.15.5.3 for further information regarding the use of these values.
f. The nearest point of public access from this postulated release is 5,870 meters (6,419 yards).
g. The nearest site boundary is located at 14,000 meters (15,310 yards).
Peak chlorine concentrations estimated at the nearest point of public access can exceed the
Emergency Response Planning Guideline-2 value assuming 95 percentile meteorological conditions, as
listed in Table 5.15-10. Symptoms associated with exposure to these concentrations could include
burning of the eyes, nose, and throat, coughing, choking, and possibly skin burns.
As listed in Table 5.15-16, the estimated peak averaged chlorine concentration at the nearest site
boundary would be above the Emergency Response Planning Guideline-1 value for 95 percentile
meteorological conditions. However, due to the nature of the release, this concentration probably
would not last for more than a few minutes. Therefore, it would be likely that individuals at this
distance would experience no more than mild transient adverse health effects.
This analysis took limited credit for emergency response actions following a chlorine release in
calculating the concentrations listed in Table 5.15-16. To mitigate the consequences of a chlorine
release to the environment, the same emergency response programs and actions described for
radiological accident scenarios (Section 5.15.4.1) would be initiated following the release. Therefore,
actual health effects experienced by persons inside the site boundary would realistically be less than
the values listed in Table 5.15-16.
Because the estimated airborne concentration of chlorine at 100 meters (328 feet) substantially
exceeds the guidelines listed in Table 5.15-16, workers could be fatally injured or could receive
long-term or permanent health effects. Potential secondary impacts associated with the chlorine
accident scenario would involve economic impacts such as workers' compensation, medical bills, and
potential lawsuits. No other secondary impacts, such as impacts on national defense or biotic
resources, were identified.
5.15.5.3.2 Accidental Nitric Acid Release - Nitric acid is used at various spent
nuclear fuel-related storage facilities for maintaining the chemistry of the water used in underwater
storage facilities(8).
Based on the toxic chemical screening discussed in Section 5.15.5.1, review of
existing safety analyses, walkdowns of spent nuclear fuel-related facilities, and interviews with INEL
------------------------------------------------------------------------------------------------------
8. Although bulk quantities of nitric acid would be required to perform processing activities that
could be resumed under Alternatives 4b(1) and 5b, the consequences of processing-related accidents
involving nitric acid would be bounded by the hydrfluoric acid and anhydrous accidents analyzed
in Sections 5.15.5.3.3. and 5.15.5.3.4., respectively. Therefore, this analysis focuses on a
potential nitric acid accident resulting from the non-processing spent nuclear fuel-related
activities considered under the other alternatives.
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personnel, DOE determined that the potential exists for an accidental release of nitric acid from one of
two 1,135 liters (300-gallon) storage tanks used to support spent nuclear fuel-related water treatment
activities at the Idaho Chemical Processing Plant. Because one of the tanks is usually empty, the two
tanks have separate valves, and they are physically separated, DOE could not identify a reasonably
likely initiator that could cause an accidental simultaneous release from both tanks.
The quantity of nitric acid assumed available for release from a single initiator would be
(1,135 liters) 300 gallons. The following assumptions were made for this analysis:
- An initiating event causes severe structural damage (e.g., large puncture) to one of the tanks.
- The entire inventory of nitric acid is released into the containment wall surrounding the
storage tank.
- The area of the containment wall is approximately 28 square meters (300 square feet).
- The total release of nitric acid [i.e., 1.135 liters (300 gallons)] evaporates into the
atmosphere before the implementation of emergency response procedures can recover the
nitric acid.
Table 5.15-16 summarizes the concentrations of the nitric acid release at the following receptor
locations for both conservative (95 percentile) and average (50 percentile) meteorological conditions:
a facility worker, a member of the public stranded at the nearest point of public access inside the
INEL boundary, and a maximally exposed hypothetical member of the public at the nearest site
boundary. The estimated frequency for this event is 1 y 10-5 events per year.
This analysis took limited credit for emergency response actions following a nitric acid release in
calculating the concentrations listed in Table 5.15-16. To mitigate the consequences of a release to the
environment, the same emergency response programs and actions described for radiological accident
scenarios (Section 5.15.4.1) would be initiated following a nitric acid release. Therefore, actual health
effects experienced by persons inside the site boundary would realistically be less than the values
listed in Table 5.15-16.
Other than limited economic secondary impacts, no other secondary impacts would be likely if
this accident occurred.
5.15.5.3.3 Accidental Hydrofluoric Acid Release - To resume spent nuclear fuel
processing activities at the Fluorinel and Storage (FAST) facility (CPP-666), which is currently
shutdown and being placed in a permanent shutdown mode, bulk quantities of hydrofluoric acid would
be required to support the dissolution process.
A hydrofluoric acid storage tank with an operating
capacity of approximately 30,283 liters (8,000 gallons) is located in the Idaho Chemical Processing
Plant facility area to support processing activities, although only 11,356 liters (3,000 gallons) of
hydrofluoric acid remain in the tank, and efforts are currently underway to remove the remaining
hydrofluoric acid in the tank from the INEL site.
Table 5.15-17 summarizes the potential impacts upon a maximally exposed hypothetically offsite
individual located at the nearest site boundary [14,000 meters (15,310 yards)] resulting from a
potential hydrofluoric acid release at the Idaho Chemical Processing Plant assuming 95 percentile
meteorological conditions. Slaughterbeck et al. (1995) provides further details and discussion
regarding this postulated accident scenario. Although Slaughterbeck et al. (1995) presents impacts to
only the maximally exposed offsite hypothetical individual resulting from this postulated accident for
95 percentile meteorological conditions, a comparison of the airborne concentration of hydrofluoric
acid at 14,000 meters (15,310 yards) to the airborne concentrations from other postulated chemical
accident scenarios (as presented in Table 5.15-16) at the same receptor distance provides meaningful
perspective on the significance of this accident.
Table 5.15-17. Summary of chemical concentrations for postulated processing-related accidental
releases at the Idaho Chemical Processing Plant under Alternatives 4b(1) and 5b.
Chemical Concentrations
(milligrams per cubic meter)a
95% Meteorologyb
Hydrofluoric Acid Anhydrous Ammonia
ERPG-1c = 4 (5) ERPG-1 = 17 (25)
ERPG-2 = 17 (20) ERPG-2 = 136 (200)
Receptor Location ERPG-3 = 43 (50) ERPG-3 = 680 (1000)
Maximally exposed hypothetical individual 0.078 82
located at the nearest boundaryd (0.09) (120.6)
a. Numbers in parentheses reflect concentrations in parts per million.
b. The 95 percentile meteorology is based on Class F (unfavorable) meteorological conditions with
0.5 meter per second (1.1 miles per hour) wind speed for receptors located within 2 kilometers
(1.2 miles) of the release and 2 meters per second (4.5 miles per hour) for receptors beyond
2 kilometers of the release.
c. ERPG = Emergency Response Planning Guidelines.
d. The nearest site boundary is located at 14,000 meters (15,310 yards).
The estimated frequency for this event is 1 y 10-5 events per year. It should be noted that this
potential accident applies only to Alternatives 4b(1) and 5b, and is in addition to the potential chlorine
and nitric acid release accidents described in Sections 5.15.5.3.1 and 5.15.5.3.2, respectively.
This analysis took limited credit for emergency response actions following a hydrofluoric acid
release in calculating the concentrations listed in Table 5.15-17. To mitigate the consequences of a
release to the environment, the same emergency response programs and actions described for
radiological accident scenarios (Section 5.15.4.1) would be initiated following a hydrofluoric acid
release. Therefore, actual health effects experienced by persons inside the site boundary would
realistically be less than the values listed in Table 5.15-17.
Other than limited economic secondary impacts, no other secondary impacts would be likely if
this accident occurred.
5.15.5.3.4 Accidental Anhydrous Ammonia Release - To resume spent nuclear
fuel processing activities at the Fluorinel and Storage (FAST) facility (CPP-666), bulk quantities of
anhydrous ammonia would be required to support operation of the NOx-Abatement Facility
(CPP-1670), a facility that would be constructed to treat airborne effluents from the INEL processing
facilities before being released to the environment.
The NOx-Abatement Facility would be expected to utilize two anhydrous ammonia tanks, each
with a storage capacity of 68,000 liters (18,000 gallons). Table 5.15-17 summarizes the potential
impacts upon the maximally exposed hypothetical offsite individual located at the nearest site
boundary [14,000 meters (15,310 yards)] resulting from a short-term release of the contents of both
storage tanks [i.e., 136,000 liters (36,000 gallons)] at the Idaho Chemical Processing Plant assuming
95 percentile meteorological conditions. Slaughterbeck et al. (1995) provides further details and
discussion regarding this postulated accident scenario. Although Slaughterbeck et al. (1995) presents
only impacts to the maximally exposed offsite hypothetical individual resulting from this postulated
accident for 95 percentile meteorological conditions, a comparison of the airborne concentration of
anhydrous ammonia at 14,000 meters (15,310 yards) to the airborne concentrations from other
postulated chemical accident scenarios (as presented in Table 5.15-16) at the same distance provides
meaningful perspective on the significance of this accident.
The estimated frequency for this event is 5 y 10-6 events per year. The basis for this estimated
frequency is identical to that described for an accidental chlorine release from two separate tanks, as
described in Section 5.15.5.3.1. It should be noted that this potential accident applies only to
Alternatives 4b(1) and 5b, and is in addition to the potential chlorine and nitric acid release accidents
described in Sections 5.15.5.3.1 and 5.15.5.3.2, respectively.
This analysis took limited credit for emergency response actions following an anhydrous
ammonia release in calculating the concentrations listed in Table 5.15-17. To mitigate the
consequences of a release to the environment, the same emergency response programs and actions
described for radiological accident scenarios (Section 5.15.4.1) would be initiated following a
hydrofluoric acid release. Therefore, actual health effects experienced by persons inside the site
boundary would realistically be less than the values listed in Table 5.15-17.
Other than limited economic secondary impacts, no other secondary impacts would be likely if
this accident occurred.
5.15.6 Maximum Reasonably Foreseeable Radiological Accident Scenario Descriptions
The purpose of this section is to summarize the different accident scenarios identified in
Section 5.15.4. The Facility Safety Report for the Argonne National Laboratory-West Hot Fuel
Examination Facility (ANL 1975) contains further details and discussions for Accident 1, discussed
below. Slaughterbeck et al. (1995) provides further details, discussions, and references for Accidents 2
through 7, discussed below. Additional discussions and references regarding the processing-related
accidents summarized in this section are also provided in a study performed to determine the potential
impacts spent nuclear fuel processing-related accidents could have on the siting of a new production
reactor at the INEL (EG&G 1993b). These documents contain additional information, such as release
fractions, source terms, and other assumptions used in the accident analyses. Appendix D describes
postulated accident scenarios associated with Naval spent nuclear fuel-related facilities and activities at
the INEL.
5.15.6.1 Accident 1: Fuel Pin Breach and Venting of Noble Gases and Iodine to
the Environment from a Mechanical Handling Accident at the Argonne National
Laboratory-West Hot Fuel Examination Facility. The accident screening methodology discussed
in Section 5.15.3 identified a mechanical handling event at the Argonne National Laboratory-West Hot
Fuel Examination Facility as an initiator to the maximum reasonably foreseeable accident within the
abnormal event frequency range. This event would result in a fuel pin breach and venting of noble
gases and iodine to the environment. The identification of this accident as a maximum reasonably
foreseeable accident is based on the estimated radiological consequences to the maximally exposed
hypothetical offsite individual at the nearest site boundary presented in the Hot Fuel Examination
Facility Safety Report (ANL 1975). Other postulated accidents associated with handling spent nuclear
fuel in the Hot Fuel Examination Facility before the identification of the fuel pin breach accident as
the maximum reasonably foreseeable accident included an inadvertent criticality and a sodium fire. A
fuel pin breach accident was chosen as the maximum reasonably foreseeable accident because the
estimated frequencies for an inadvertent criticality and a sodium fire in the facility are extremely low
(ANL 1975).
The analyses defined in the Facility Safety Report (ANL 1975) made the following assumptions:
- The fuel subassemblies and experimental capsules being examined in the facility were
cooled for at least 15 days to ensure that the short-lived fission products had decayed.
- The noble gases and iodines that could be released from this accident scenario were
immediately released.
- One hundred percent of the noble gases, 25 percent of the iodines, and 1 percent of
particulates were available for escape to the atmosphere.
- The building containment structure, including the building ventilation system, and the Main
Cell, including the argon ventilation system, remained operational following the handling
accident. This assumption is considered appropriate because the mechanical handling
accident scenario under consideration would not initiate a failure in these systems.
(Accident 3 considers the simultaneous failure of all these systems in conjunction with the
melting of fuel assemblies stored in the facility).
The Facility Safety Report (ANL 1975) contains specific information on the source terms
associated with breaching the fuel section of a pin. Because that report does not provide an estimated
frequency of occurrence for the subject mechanical handling accident scenario, the analysis used
historic information and engineering judgment to determine the conservatively estimated frequency for
this accident of 1.0 y 10-2 event per year.
For determining the impacts from this postulated accident scenario, the nearest point of public
access is equivalent to the nearest site boundary, which is 5,240 meters (5,730 yards) from the point of
the release. Although the Facility Safety Report (ANL 1975) does not estimate consequences to the
offsite population resulting from this accident scenario, this analysis reasonably estimated that the
exposures (i.e., dose) to the offsite population would be less than the offsite population dose calculated
for Accidents 2 through 4 because the dose to the maximally exposed hypothetical individual at the
nearest site boundary from this accident would be less than that estimated for Accidents 2 through 4.
5.15.6.2 Accident 2: Inadvertent Nuclear Chain Reaction in Wet Spent Nuclear
Fuel Storage (1 y 1019 fissions, 8-hour release) at the Idaho Chemical Processing Plant
CPP-603 Underwater Fuel Storage Facility. The accident screening methodology discussed in
Section 5.15.3 identified an inadvertent nuclear criticality associated with underwater spent nuclear fuel
storage at the CPP-603 Underwater Fuel Storage Facility as an accident requiring further evaluation.
Other postulated accidents that were considered before the identification of an inadvertent criticality
accident as a maximum reasonably foreseeable accident included pool leaks, fuel damage events, and
loss of cooling events. This analysis selected an inadvertent nuclear criticality for evaluation in this
EIS over the other accidents for the following reasons:
- Postulated inadvertent nuclear criticality accidents have been addressed in virtually all DOE
nonreactor EISs and safety analysis reports in which such accidents were reasonably
foreseeable because of public concerns regarding the potential for these accidents.
- The Idaho Chemical Processing Plant has experienced three inadvertent nuclear criticality
accidents. Although none of these accidents involved a fuel storage facility, they
demonstrate the potential and concern for such events.
- The consequences of water leakage from a pool-draining event would present lower prompt
consequences to workers than a criticality because the INEL could implement emergency
response plans to evacuate workers before the risk to these workers could substantially
increase. In addition, a pool drain was considered to be an initiator to a criticality accident.
- Mechanical fuel damage events are less impacting than a nuclear chain reaction scenario
because some degree of fuel damage is part of the criticality accident scenario and analysis.
Of the different Idaho Chemical Processing Plant facility areas that store spent nuclear fuel, the
CPP-603 Underwater Fuel Storage Facility was selected for analysis of a criticality accident for the
following reasons:
- CPP-603 facility storage includes most types of spent nuclear fuel stored elsewhere on the
site. Fuel stored at reactor basins is an exception (but was considered in the determination
of other reasonably foreseeable accident scenarios) because of its much shorter cooling times
after removal from a reactor.
- CPP-603 facility spent nuclear fuel storage quantities are comparable to or exceed the spent
nuclear fuel inventories stored elsewhere on the site.
- The CPP-603 facility is an older facility that does not contain all the preventive or
mitigative design features found in more modern facilities, such as the CPP-666 Fuel
Storage Area.
The analysis selected the underwater fuel storage portion of the CPP-603 facility rather than the
Irradiated Fuels Storage Facility portion of the CPP-603 facility because accidents involving graphite
fuels in dry storage probably would have less severe potential consequences because they had been
removed from reactors for a much longer period of time and, because of their design, would prevent
most of the remaining fission products from being released if a criticality accident occurred.
Initiating events that the analysis considered possible to lead to an inadvertent nuclear criticality
included operator error, hanger corrosion, equipment failure, an earthquake, pool drain, and an aircraft
crash. The scenario discussed in this EIS assumes a postulated criticality scenario that could be
initiated by human error, equipment failure, or earthquake. Heat generated from the chain reaction
would easily dissipate and thereby avoid fuel melting but would still cause the release of fission
products associated with 1 y 1019 fissions over an 8-hour period.
Between 1945 and 1980, 40 known inadvertent criticalities occurred worldwide, none of which
involved the handling or storage of spent nuclear fuel in an underwater fuel storage facilities. In
addition, between 1975 and 1980, there were 160 nuclear power reactor facilities with underwater fuel
storage facilities worldwide. None of these facilities ever had a nuclear criticality associated with its
underwater storage facilities. Therefore, it is generally assumed that the likelihood for such an event
in a modern underwater storage facility is unlikely, with a frequency estimated at 1 y 10-4 event per
year. This estimated frequency is supported by information in the safety analysis report for the
CPP-666 underwater storage facility, which is a modern facility (e.g., 1980s vintage) at the INEL used
to store various types of spent nuclear fuel. In the CPP-603 Underwater Fuel Storage Facility,
however, where spent nuclear fuel inventories have substantially corroded or degraded (DOE 1993c),
and where the design of the facility and its supporting equipment do not meet current design
specifications, activities associated with handling and storing spent nuclear fuel present an increase in
the likelihood for an inadvertent nuclear criticality accident by as much as an order of magnitude.
Therefore, this analysis conservatively assumes the estimated frequency for an inadvertent nuclear
criticality associated with handling spent nuclear fuel in the CPP-603 Underwater Fuel Storage Facility
to be 1 y 10-3 event per year for this analysis.
The handling activities associated with stabilizing CPP-603 facility spent nuclear fuel inventories
would occur under each of the five alternatives considered in this EIS. The estimated frequency for an
inadvertent criticality at the CPP-603 facility is an order of magnitude larger than that of any other
INEL facility (e.g., 1 y 10-3 event per year), and is considered a "worst-case" frequency that bounds
changes in estimated criticality frequencies at other INEL facilities resulting from increased handling
activities associated with changes in spent nuclear fuel inventories. Therefore, using the estimated
criticality frequency related to the CPP-603 as the estimated frequency under each alternative provides
a conservative bound on the estimated criticality frequencies for other spent nuclear fuel-related
handling and storage facilities.
To determine the accident impacts from this postulated accident scenario, the analysis assumed
the worker to be located 100 meters (328 feet) from the event, the nearest point of public access (U.S.
Route 20/26) is 5,870 meters (6,420 yards), and the nearest site boundary is located at 14,000 meters
(15,310 yards).
5.15.6.3 Accident 3: Earthquake-Induced Breach and Fuel Melt at the Argonne
National Laboratory-West Hot Fuel Examination Facility. The accident screening
methodology discussed in Section 5.15.3 identified an earthquake-induced breach and fuel melt at the
Argonne National Laboratory-West Hot Fuel Examination Facility as a maximum reasonably
foreseeable accident that would present higher radiological consequences to facility workers or the
offsite population than other postulated accidents analyzed in the same accident frequency range. The
postulated events leading to atmospheric release of radionuclides are as follows:
- The earthquake results in a peak horizontal ground acceleration of sufficient magnitude to
cause structural damage to the building structure and a large breach in the main cell.9
- Coincident with the breach, a failure of the fuel subassembly cooling system occurs,
resulting in the melting of fresh assemblies.
- Radionuclides from the melting fuel subassemblies are released to the atmosphere.
The estimated probability of an earthquake in the Argonne National Laboratory-West facility area
resulting in a peak horizontal acceleration of sufficient magnitude to damage the facility structure and
breach the cell is 1 y 10-5 event per year. This analysis conservatively assumes the probability of
failure of the building structure, Main Cell, and subassembly cooling to be 1.0, given that the
earthquake has occurred. A preliminary assessment of the seismic integrity of the Hot Fuel
Examination Facility, as discussed in Slaughterbeck et al. (1995), indicates that, given the current state
of analysis, significant failures could result at the Hot Fuel Examination Facility from this earthquake.
In determining the number of fuel assemblies that would be affected during this scenario, the
analysis assumed that 20 fuel subassemblies would melt due to failure of the forced cooling in this
accident. Although 40 storage positions are available for fuel that would require forced cooling,
current plans do not estimate the need to use more than 20 of these positions. The release duration for
this scenario is 30 days. To prevent doses greater than 5 rem to the public from this scenario,