ML21096A293

NUREG/BR-0058, Rev. 5, Appendix G Regulatory Analysis Methods and Data for Nuclear Facilities Other than Power Reactors - Dfc

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NUREG/BR-0058, Rev. 5
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1 APPENDIX G 2 REGULATORY ANALYSIS METHODS AND DATA FOR NUCLEAR 3 FACILITIES OTHER THAN POWER REACTORS

1 TABLE OF CONTENTS 2

3 LIST OF FIGURES...................................................................................................... G-v 4 LIST OF TABLES ....................................................................................................... G-v 5 ABBREVIATIONS AND ACRONYMS ...................................................................... G-vii 6 G.1 PURPOSE ........................................................................................................ G-1 7 G.2 FUEL CYCLE ACTIVITIES .............................................................................. G-2 8 G.2.1 Mining and Milling ........................................................................................... G-4 9 G.2.2 Conversion...................................................................................................... G-6 10 G.2.3 Enrichment...................................................................................................... G-7 11 G.2.4 Fuel Fabrication .............................................................................................. G-9 12 G.2.5 Spent Nuclear Fuel Storage ......................................................................... G-10 13 G.2.6 Low-Level Waste Disposal ........................................................................... G-12 14 G.2.7 Geological Repository Risks ......................................................................... G-14 15 G.2.7.1 Transportation of Spent Nuclear Fuel to a Repository ................ G-14 16 G.2.7.2 Repository Construction, Operation, and Closure ...................... G-19 17 G.2.7.3 Repository Risks after Permanent Closure ................................. G-33 18 G.2.7.4 Summary of Risks of a Geological Repository............................ G-34 19 G.2.8 Decommissioning ......................................................................................... G-35 20 G.3 NON-FUEL CYCLE ACTIVITIES ................................................................... G-37 21 G.4 COMMON ACTIVITIES .................................................................................. G-40 22 G.4.1 Transportation............................................................................................... G-40 23 G.4.2 Security ......................................................................................................... G-42 24 G.4.3 Material Control and Accountability .............................................................. G-45 25 G.4.4 Emergency Planning and Preparedness ...................................................... G-46 26 G.5 REFERENCES ............................................................................................... G-48 27 G-iii NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 LIST OF FIGURES 2

3 Figure G-1 The Nuclear Fuel Cycle .................................................................................. G-2 4 Figure G-2 Schematic of a Single Gas Centrifuge Cylinder ............................................. G-8 5 Figure G-3 Typical Light-Water Reactor Fuel Fabrication Facility .................................. G-10 6

7 LIST OF TABLES 8

9 Table G-1 Data Used to Estimate Radiation Doses for Loading ................................... G-15 10 Table G-2 Estimated Radiological Impacts from Loading Operations ........................... G-16 11 Table G-3 Estimated Worker and Public Impacts from Shipping Commercial 12 SNF to the Repository over a 50-Year Period .............................................. G-17 13 Table G-4 Atmospheric Release of Radon through Ventilation of the Underground 14 Facility during Repository Development ....................................................... G-21 15 Table G-5 Internal Dose Conversion Factors for the Maximally Exposed Individuals 16 and the Offsite Population Present on the Land Surface Based on a 17 Release of 1 Ci of Radon ............................................................................. G-22 18 Table G-6 Internal and External Dose Conversion Factors for the Maximally 19 Exposed Worker in the Subsurface from Natural Sources of Radiation 20 throughout Repository Development ............................................................ G-23 21 Table G-7 Maximum Annual Individual Dose for Workers and the Public from 22 Naturally Occurring Radionuclides during the Development of a 23 Repository .................................................................................................... G-24 24 Table G-8 Population Size for Workers (Worker-Years) during Repository 25 Development ................................................................................................ G-25 26 Table G-9 Collective Population Dose for Workers and the Public in Person-Rem 27 from Natural Sources of Radiation during the Development of a 28 Repository .................................................................................................... G-25 29 Table G-10 Annual Individual Internal Dose for Atmospheric Releases of SNF 30 during Operations ......................................................................................... G-27 31 Table G-11 External Dose for Involved Workers for Specific Activities ........................... G-28 32 Table G-12 Largest Annual Individual Dose for Workers and the Maximally 33 Exposed Offsite Individual from SNF during the Operational Period ............ G-29 34 Table G-13 Collective Internal Dose for Workers and the Public from SNF during 35 Repository Development .............................................................................. G-31 36 Table G-14 Conditional Doses for Accident Scenarios Involving SNF Releases 37 during Repository Handling and Emplacement Activities ............................. G-32 38 Table G-15 Risk Associated with Accident Scenarios Involving SNF Releases 39 during Repository Handling and Emplacement Activities ............................. G-33 40 Table G-16 Collective Dose for Workers and the Public from the Transportation 41 Campaign and Repository Development ...................................................... G-35 42 G-v NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 ABBREVIATIONS AND ACRONYMS 2

3 ACNW Advisory Committee on Nuclear Waste 4 BLS Bureau of Labor Statistics 5 BSC Bechtel SAIC Company 6 CFR Code of Federal Regulations 7 Ci curie(s) 8 DOE U.S. Department of Energy 9 DOL U.S. Department of Labor 10 DPC dual-purpose canister 11 EIS environmental impact statement 12 EPA U.S. Environmental Protection Agency 13 FNMC Fundamental Nuclear Material Control 14 FR Federal Register 15 GTCC greater-than-Class C 16 HEPA high-efficiency particulate air 17 HF hydrogen fluoride 18 HLW high-level waste 19 ISA integrated safety analysis 20 ISFSI independent spent fuel storage installation 21 ISL/ISR in situ leach/in situ recovery 22 LLW low-level waste 23 mrem millirem 24 MTHM metric ton(s) of heavy metal 25 NA not available or not applicable 26 NEPA National Environmental Policy Act 27 NMED Nuclear Materials Events Database 28 NRC U.S. Nuclear Regulatory Commission 29 OMB U.S. Office of Management and Budget 30 OpE operating experience 31 PRA probabilistic risk assessment 32 PuO2 plutonium dioxide 33 REIRS Radiation Exposure Information Reporting System 34 RF receipt facility 35 SFP spent fuel pool 36 SNF spent nuclear fuel 37 TAD transportation, aging, and disposal 38 U uranium 39 UF6 uranium hexafluoride 40 U3O8 uranium (V, VI) oxide, yellowcake 41 UO2 uranium dioxide 42 UO2F2 uranyl fluoride 43 WHF wet handling facility 44 yr year G-vii NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 REGULATORY ANALYSIS METHODS AND DATA FOR NUCLEAR 2 FACILITIES OTHER THAN POWER REACTORS 3

4 G.1 PURPOSE 5

6 This appendix documents established approaches and data considerations for use in 7 performing non-power reactor regulatory analyses. The information presented in this appendix 8 supplements the basic concepts for conducting a regulatory analysis described in Sections 2 9 through 5 of this NUREG. This appendix is for use by the analyst preparing a regulatory 10 decision-making document for non-power reactor facilities and activities, including fuel 11 fabrication facilities, independent spent fuel storage installations (ISFSIs), irradiators, uses of 12 byproduct material, and high-level waste (HLW) repositories.

13 14 The analyst should strive to use quantitative measures when performing a regulatory analysis 15 for regulatory changes affecting non-power reactor facilities and activities. However, many 16 benefits of improved regulation for these facilities and activities do not lend themselves to 17 quantification. As discussed in Appendix A, Qualitative Factors Assessment Tools to this 18 NUREG, nonquantifiable costs and benefits can be significant elements of a regulatory analysis, 19 and the analyst and decisionmaker should consider them as appropriate. The analyst should 20 refer to Risk-Informed Decisionmaking for Nuclear Material and Waste Applications, as a 21 supplement to this NUREG for guidance on making risk-informed regulatory decisions for 22 nonpower reactors.

23 24 The analyst will need to consider NRCs backfitting regulations when evaluating changes to 25 requirements for non-power reactor activities. These requirements are found in 10 CFR 70.76, 26 Backfitting, for specific licenses involving special nuclear material; 10 CFR 72.62, Backfitting, 27 for ISFSIs and monitored retrievable storage facilities; and 10 CFR 76.76, Backfitting, for 28 gaseous diffusion plants. Management Directive 8.4, Management of Facility-Specific 29 Backfitting and Information Collection, and NUREG-1409, Backfitting Guidelines, provide 30 guidance to NRC staff on the application of the substantial increase standard as it relates to the 31 NRCs backfit regulations.

32 33 This document is organized into three sections: Section G.2, Fuel Cycle Activities, describes 34 the activities, hazards, and data sources for mining, milling, conversion, enrichment, and fuel 35 fabrication for nuclear reactors, as well as activities associated with the storage and disposal of 36 the spent nuclear fuel (SNF) and low-level radioactive waste. Section G.3, Non-Fuel-Cycle 37 Activities, addresses activities involving the use of radioactive (byproduct) material, such as for 38 medical and industrial use (e.g., irradiators, radiography, well-logging, manufacturing, gauges) 39 and for academic or research use. Section G.4, Common Activities, addresses those activities 40 that are common to both groups of non-power reactor activities, such as transportation, security, 41 material control and accountability, and emergency planning and preparedness.

G-1 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 G.2 FUEL CYCLE ACTIVITIES 2

3 Fuel cycle activities are those activities associated with the extraction of uranium (and other 4 materials), production of fuel for use in nuclear reactors, and storage and disposal of SNF and 5 associated radioactive wastes from nuclear reactor operations. The NRC regulations for these 6 activities are 10 CFR Part 40, Domestic Licensing of Source Material; 10 CFR Part 60, 7 Disposal of High-Level Radioactive Wastes in Geologic Repositories; 10 CFR Part 61, 8 Licensing Requirements for Land Disposal of Radioactive Waste; 10 CFR Part 63, Disposal of 9 High-level Radioactive Wastes in a Geologic Repository at Yucca Mountain, Nevada; and 10 parts 70 through 76 of 10 CFR 1. The NRCs environmental protection regulations implement 11 the National Environmental Policy Act (NEPA) and are in 10 CFR Part 51, Environmental 12 Protection Regulations for Domestic Licensing and Related Regulatory Functions.

13 14 NUREG/BR-0280, Revision 1, Regulating Nuclear Fuel, is an overview of the nuclear fuel 15 cycle and includes a high-level discussion of possible hazards in various processes. Figure G-1 16 illustrates general fuel cycle activities.

17 18 19 Figure G-1 The Nuclear Fuel Cycle 20 This section discusses the fuel activities shown in Figure G-1, the most significant hazards for 21 each activity, and useful sources of data 2:

22 23

• source materialuranium recovery (i.e., mining, in situ leaching, heap leaching) 1 10 CFR Part 70, Domestic Licensing of Special Nuclear Material; 10 CFR Part 71, Packaging and Transportation of Radioactive Material; 10 CFR Part 72, Licensing Requirements for the Independent Storage of Spent Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater than Class C Waste; 10 CFR Part 73, Physical Protection of Plants and Materials; 10 CFR Part 74, Material Control and Accounting of Special Nuclear Material; 10 CFR Part 75, Safeguards on Nuclear Material Implementation of Safeguards Agreements between the United States and the International Atomic Energy Agency; and 10 CFR Part 76, Certification of Gaseous Diffusion Plants.

2 Operational/facility decommissioning, while not identified in Figure G-1, is discussed in Section G.2.8 of this appendix.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-2

1

• milling (including reclamation of mill tailings and decommissioning) 2

• conversion 3

• enrichment 4

• fuel fabrication (including fuel refabrication, such as mixed oxide fuel) 5

• SNF storage 6 - spent fuel pool (SFP) 7 - ISFSI or monitored retrievable storage 8

• fuel reprocessing 9

• disposal 10 - low-level waste (LLW) disposal 11 - HLW geologic repository 12 13 Uranium recovery by in situ leach and heap leach, discussed in Section G.2.1, fall under NRC 14 regulatory purview; however; traditional mining is not regulated by the NRC so it is not 15 discussed in this document. The United States currently has no licensed commercial 16 reprocessing facilities. Should a regulatory analysis need to consider a reprocessing facility, 17 staff should rely on NUREG/CR-7232, Review of Spent Fuel Reprocessing and Associated 18 Accident Phenomena. That document includes an overview of typical facility designs and 19 describes historical accidents and the phenomena relevant to accidents at this type of facility.

20 There also are no gaseous diffusion plants operating in the United States. Because the NRC 21 has no knowledge of any plan to use this technology in the future, this appendix does not 22 discuss these facilities or the associated regulations in 10 CFR Part 76.

23 24 A regulatory analysis and an environmental analysis to comply with NEPA are required for all 25 fuel cycle facilities. Additionally, the NRC has backfitting provisions for non-power reactor 26 licenses issued under 10 CFR parts 70, 72, and 76, for uranium enrichment, fuel fabrication, 27 ISFSIs, HLW, and gaseous diffusion plants.

28 29 There are many sources of data available to the analyst who is developing a regulatory analysis 30 for a regulatory action affecting specific fuel cycle activities. The list of general references 31 below provides a starting point for performing the regulatory analysis; specific data sources are 32 included in the discussion for each fuel cycle activity.

33 34

• NUREG/CR-2873, Volume 1, Nuclear-Fuel-Cycle Risk Assessment: Descriptions of 35 Representative Non-Reactor Facilities, provides a general overview of each of the fuel 36 cycle activities and risks associated with each activity.

37 38

• NUREG/CR-2933, Nuclear Fuel Cycle Risk Assessment: Survey and Computer 39 Compilation of Risk-Related Literature, provides a survey and compilation of the 40 risk-related literature for fuel cycle activities, including characterization of the specific 41 risk/safety information.

42 43

• NUREG/CR-3682, Nuclear Fuel Cycle Risk Assessment: Review and Evaluation of 44 Existing Methods, provides a preliminary relative ranking of fuel cycle facilities on the 45 basis of risk and an assessment of the adequacy of the existing (1982 timeframe) risk 46 assessment methods.

47 48 More recent references that are generally applicable to fuel cycle activities include the annual 49 NUREG-0713, Occupational Radiation Exposures at Commercial Nuclear Power Plant G-3 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 Reactors and Other Facilities, the Nuclear Materials Events Database (NMED), and information 2 available to the NRC staff on the Materials Operating Experience (OpE) Gateway.

3 4 NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook, provides 5 guidance on how to calculate the characteristics and consequences of releases of radioactive 6 materials and hazardous chemicals from fuel cycle facilities. The appendices to 7 NUREG/CR-6410 summarize chemical and nuclear information and describe various fuel cycle 8 facilities, details on calculating the characteristics of source terms for releases of hazardous 9 chemicals, and other topics (e.g., high-efficiency particulate air [HEPA] filter performance and 10 uncertainties). NUREG/CR-6410 also provides several sample problems addressing some of 11 the more significant hazards (e.g., uranium hexafluoride [UF6] cylinder rupture and criticality).

12 13 Additionally, a significant amount of risk assessment work has been performed to support the 14 review and licensing of an HLW geologic repository. Section G.2.7 presents an extensive 15 discussion of the aspects associated with the HLW risk analyses, many of which are also 16 applicable to other fuel cycle activities. The analyst should consider this information, as 17 appropriate, for similar activities such as SNF loading, packaging, movement, and 18 transportation.

19 20 G.2.1 Mining and Milling 21 22 The nuclear fuel cycle begins with uranium recovery (i.e., mining), which is the extraction of 23 natural uranium ore from the earth and concentrating (i.e., milling) that ore. These recovery 24 operations produce yellowcake. Uranium recovery operations also generate byproduct material 25 waste.

26 27 NRC has determined that a license is not required for possession, use, or transfer of unrefined 28 and unprocessed ore containing source material (10 CFR 40.13(b)). However, a specific 29 license is required for in situ recovery (ISR) (also referred to as in situ leach (ISL) mining) 30 because uranium ore is processed underground before being pumped to the surface. A license 31 is also required for a uranium mill, which is the next step in processing ore from a conventional 32 mine. The NRC does not directly regulate the active uranium recovery operations in Agreement 33 States. 3 34 35 The majority of uranium extraction operations in the United States use ISR. The final stages of 36 the ISR process produce yellowcake using low temperature vacuum dryers. A source and 37 byproduct materials license is required to recover uranium by ISR extraction techniques under 38 the provisions of 10 CFR Part 20, Standards for Protection against Radiation, and 10 CFR 39 Part 40, Domestic Licensing of Source Material. Uranium milling and disposal of the resulting 40 waste byproduct material by NRC licensees are also regulated under 10 CFR Part 40. Most of 41 the regulations that the NRC has established for this type of byproduct material are found in 10 42 CFR Part 40, Appendix A, Criteria Relating to the Operation of Uranium Mills and the 43 Disposition of Tailings or Wastes Produced by the Extraction or Concentration of Source 3 The NRC assists States expressing interest in establishing programs to assume NRC regulatory authority under the Atomic Energy Act of 1954, as amended (the Act). Section 274 of the Act provides a statutory basis under which the NRC relinquishes to the States portions of its regulatory authority to license and regulate byproduct materials (radioisotopes), source materials (uranium and thorium), and certain quantities of special nuclear material. The mechanism for the transfer of the NRC's authority to a State is an agreement signed by the Governor of the State and the Chairman of the Commission, in accordance with Section 274b of the Act.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-4

1 Material from Ores Processed Primarily for Their Source Material Content. In general, the 2 criteria in 10 CFR Part 40, Appendix A, require uranium recovery facilities to control hazards 3 and address waste and decommissioning concerns.

4 5 Potential accidents during ISR processing do not yield releases much higher than those 6 incurred during normal operation. These include various types of spills resulting from barrier 7 breaches (i.e., failure of thickener tank, pipes, valves, or yellowcake dryer) leading to the 8 release of pregnant lixiviant 4 or yellowcake.

9 10 The primary industrial hazards associated with uranium milling are the occupational hazards 11 found in any metal milling operation that uses chemical extraction, as well as the chemical 12 toxicity of the uranium itself. Because the uranium produced at these facilities is not enriched 13 there is no criticality hazard. Potential hazards at conventional milling facilities include a breach 14 of the yellowcake dryer resulting in a loss of yellowcake powder (release of uranium) and the 15 potential release of chemicals that are used in the milling process. The primary radiological 16 hazard is attributable to the presence of radium in mill tailings, which is a direct radiation hazard 17 and causes radon emissions from the impoundment area. A collapse of the impoundment also 18 could result in the release of hazardous material to the environment.

19 20 An applicant for a license, or for renewal or amendment of an existing license, is required to 21 provide detailed information on the facilities, equipment, and procedures used, as well as an 22 environmental report that discusses the effects of proposed operations on public health and 23 safety and the environment. These licensee submittals provide a foundation for any related 24 analysis.

25 26 The following are some of the key data references that address uranium recovery:

27 28

• NUREG-1569, Standard Review Plan for In Situ Leach Uranium Extraction License 29 Applications, contains guidance for staff review of applications and provides a 30 framework for license applications. The analyst preparing a regulatory analysis should 31 consult the specific sections on the license application and the staff safety evaluation for 32 basic information on hazards and effects.

33 34

• NUREG-2126, Standard Review Plan for Conventional Uranium Mill and Heap Leach 35 Facilities - Draft Report for Comment, provides guidance for staff review of applications 36 for conventional uranium mills and heap leach facilities. In preparing a regulatory 37 analysis, basic information on hazards and effects should be available in specific 38 sections of the license application and the staff safety evaluation.

39 40

• NUREG/CR-6733, A Baseline Risk-Informed, Performance-Based Approach for In Situ 41 Leach Uranium Extraction Licensees, is a foundational risk characterization of general 42 ISL facility operations. NUREG/CR-6733 describes the operations for extracting and 43 processing uranium into yellowcake, explains the process for restoring groundwater 44 quality subsequent to ore extraction, and identifies health and environmental hazards 45 and risks.

46 4 In ISL operations, a suitable injection system introduces a lixiviant into a subterranean uranium ore deposit.

The lixiviant may be an acidic or alkaline medium that solubilizes uranium as it traverses the ore. The pregnant lixiviant (i.e., lixiviant containing soluble uranium) is then withdrawn and treated to recover the uranium, using suitable techniques.

G-5 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1

• NUREG-0706, Final Generic Environmental Impact Statement on Uranium Milling, 2 addresses common environmental issues associated with the construction, operation, 3 and decommissioning of milling facilities and ISR-type facilities, as well as ground water 4 restoration at these facilities. NRC staff use this Generic Environmental Impact 5 Statement as a starting point for a site-specific environmental review of a license 6 application, renewal, or amendment when addressing environmental issues common to 7 milling and the ISR process.

8 9

• NUREG-1910, Generic Environmental Impact Statement for In-Situ Leach Uranium 10 Milling Facilities, analyzes potential environmental impacts associated with the 11 construction, operation, and decommissioning of ISR activities. The analyst can use 12 NUREG-1910 as a starting point for the site-specific environmental review of a license 13 application, renewal, or amendment specific to ISR activities. In particular, Appendix E, 14 Hazardous Chemicals, provides an accident analysis for the more hazardous 15 chemicals associated with ISR operations.

16 17

• NUREG-1620, Revision 1, Standard Review Plan for the Review of a Reclamation Plan 18 for Mill Tailings Sites under Title II of the Uranium Mill Tailings Radiation Control Act, 19 discusses hazards assessment, exposure assessment, corrective action assessment, 20 and compliance monitoring for alternative concentration limits, and addresses 21 reclamation and stabilization cost estimates.

22 23

• Regulatory Impact Analysis of Final Environmental Standards for Uranium Mill Tailings 24 at Active Sites, prepared by the Environmental Protection Agency (EPA) following the 25 passage of the Uranium Mill Tailings Radiation Control Act of 1978, set standards that 26 cover the processing and disposal of byproduct materials at mills that were licensed by 27 the appropriate regulatory authorities at that time. This regulatory impact analysis 28 examines the benefits and costs associated with the disposal of uranium mill tailings and 29 gives some indication of what level of control of the hazards is most cost-effective. This 30 EPA regulatory impact analysis could be useful in preparing future analyses in this area.

31 32 G.2.2 Conversion 33 34 After the uranium ore concentrate is produced at the milling facility, the yellowcake is packaged 35 and shipped to a uranium conversion facility. At the conversion facility, the yellowcake is 36 reacted with fluorine to create UF6, which is suitable for use in enrichment operations. The UF6 37 exits the conversion process as a gas that is then cooled to a liquid and drained into storage 38 and transport cylinders. As the UF6 cools over the course of several days, it transitions from a 39 liquid to a solid. The cylinder containing UF6 in the solid form can then be shipped to an 40 enrichment facility.

41 42 As with mining and milling, the primary risks associated with conversion are chemical, rather 43 than radiological. The process to convert uranium ore concentrate (uranium oxide) powder to 44 UF6 uses a number of flammable, volatile, and soluble chemicals, including fluorine, hydrofluoric 45 acid, and hydrogen. These chemicals, in particular UF6, contribute to significant risks 46 associated with inhalation if a release occurs. UF6 is produced on a large scale for uranium 47 enrichment. It is a white, somewhat volatile, solid at room temperature and pressure. UF6 can 48 form metal fluorides and react, often explosively, with organic material to form fluorinated 49 compounds and hydrogen fluoride (HF). In the presence of water, including moisture in the air, 50 UF6 forms highly corrosive and toxic HF gas and particulate uranyl fluoride (UO2F2), which are NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-6

1 extremely toxic. The UO2F2 and HF, which form quickly during a release to the atmosphere, are 2 readily visible as a white cloud. One of the most significant hazards for a conversion facility 3 occurs during those stages in which the UF6 is in liquid form and is processed. Conversion 4 processes also use hydrogen gas, which is flammable and if released could create an explosive 5 hazard. Nuclear criticality is not a hazard at these facilities because the nuclear material is 6 unenriched (i.e., consists of natural uranium) throughout the process (NRC, 2010a).

7 8 Currently, the United States has only one licensed commercial uranium conversion facility. The 9 NRC issued the license for this facility under 10 CFR Part 40. An applicant for a license, or for 10 renewal or amendment of an existing license, is required to provide information consistent with 11 10 CFR 40.31, Application for Specific Licenses. The applicant provides detailed information 12 on the facilities, equipment, and procedures used, as well as an environmental report that 13 discusses the effects of proposed operations on public health and safety and the environment.

14 These licensee submittals provide a starting point for any related analysis. Regulatory 15 Guide 3.55, Standard Format and Content for the Health and Safety Sections of License 16 Renewal Applications for Uranium Hexafluoride Production, provides high-level guidance for 17 the preparation of the health and safety section of a renewal application. In particular, Part II of 18 the regulatory guide addresses the safety demonstration, including Chapter 14 on accident 19 analysis.

20 21 In addition to the above sources of information, NUREG-1601, Chemical Process Safety at 22 Fuel Cycle Facilities, provides broad guidance on chemical process safety issues relevant to 23 fuel cycle facilities, including some examples for addressing chemical process safety, and sets 24 forth the basic information needed to properly evaluate chemical process safety.

25 26 G.2.3 Enrichment 27 28 As of 2018, only gaseous centrifuge technology is used in the United States for commercial 29 nuclear fuel enrichment. In this process, the UF6 gas is placed in a gas centrifuge cylinder and 30 rotated at a high speed. This rotation creates a strong centrifugal force so that the heavier gas 31 molecules (UF6 containing uranium [U]-238 atoms) move towards the outside of the cylinder.

32 The lighter gas molecules (containing U-235 atoms) collect closer to the center. The stream 33 that is slightly enriched in U-235 is withdrawn and fed into the centrifuge in the next higher 34 stage. The slightly depleted stream (with a lower concentration of U-235) is recycled back into 35 the next lower stage. A gas centrifuge facility contains long lines of many rotating cylinders.

36 These cylinders are connected in both series and parallel formations. Centrifuge machines are 37 interconnected to form trains and cascades. At the final withdrawal point, the UF6 is enriched to 38 the desired amount. Figure G-2 illustrates the operation of a single cylinder within a gas 39 centrifuge process.

40 G-7 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 2 Figure G-2 Schematic of a Single Gas Centrifuge Cylinder 3 Source: NRC Backgrounder - Uranium Enrichment 4

5 The NRC regulates these facilities under 10 CFR Part 70, including Subpart H, Additional 6 Requirements for Certain Licensees Authorized to Possess a Critical Mass of Special Nuclear 7 Material. The licensee or applicant is required to provide detailed information on the facilities, 8 equipment, and procedures used, as well as an environmental report that discusses the effects 9 of proposed operations on the health and safety of the workers and the public. These licensing 10 submittals provide a foundation for any related analysis, which includes the requirement in 11 Subpart H for the licensee to perform an integrated safety analysis (ISA), identify items relied on 12 for safety, and establish management measures.

13 14 Because the enrichment facilities handle and process large amounts of UF6, the same chemical 15 hazards exist at these facilities as at the conversion facility. In addition, the production of 16 enriched uranium could potentially result in an inadvertent nuclear criticality under certain 17 conditions, which would lead to high levels of radiation that can be a significant risk to workers 18 or people in the vicinity of the facility. Hence, the primary chemical hazards for enrichment 19 facilities are releases of UF6, and its subsequent products (HF and UO2F2), from storage 20 containers (both enriched and unenriched). The primary radiological concern is inadvertent 21 criticality during the enrichment process. In the case of an accident, the workers have a greater 22 chance of being impacted than the public. These facilities generally pose a low risk to the 23 public.

24 25 The following are some of the key data references that address enrichment facilities:

26 27

• NUREG-1520, Revision 2, Standard Review Plan for Fuel Cycle Facilities License 28 Applications, contains guidance for staff review of applications provides a framework for NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-8

1 license applications. The analyst preparing a regulatory analysis should consult the 2 specific sections on the license application and the staff safety evaluation for basic 3 information on hazards and effects. In particular, Chapter 3 addresses the performance 4 of an ISA for a facility.

5 6

• NUREG-1513, Integrated Safety Analysis Guidance Document, provides general 7 guidance to licensees and applicants in performing an ISA. The ISA is expected to form 8 the basis of a safety program at these facilities licensed under 10 CFR Part 70, 9 Subpart H. The ISA would involve the following:

10 11 - performance of process hazards analysis 12 - identification of accident initiating events 13 - development of accident sequences and their consequences 14 - the associated identification of items relied on for safety to prevent or mitigate 15 these accident sequences 16 As such, the licensees ISA provides the bases upon which a regulatory analysis can be 17 performed. However, the ISA itself is not part of the license and is not submitted to the 18 NRC. The licensee submits an ISA summary annually to the NRC to reflect changes 19 implemented at the facility in the prior year that resulted in changes to the ISA. In order 20 to appropriately use the ISA information in a regulatory analysis, especially a 21 facility-specific regulatory analysis or backfit analysis, the analyst may need to visit the 22 site to review the specific process ISAs in order to fully understand the potential 23 implications of the staffs proposed actions.

24 25

• NUREG/CR-6410 provides specific guidance on performing a hazard evaluation and 26 developing scenarios, and it contains a number of sample problems, including a liquid 27 spill, an HF release, a UF6 liquid cylinder rupture, and a criticality incident.

28 NUREG/CR-6410 provides significant insight into performing new, or extending existing 29 analyses in analyzing a regulatory action.

30 31 G.2.4 Fuel Fabrication 32 33 Fuel fabrication facilities convert enriched uranium into fuel for nuclear reactors. Fuel 34 fabrication for current commercial nuclear power reactors begins with the receipt of 35 low-enriched uranium as UF6 from an enrichment facility. The solid UF6 is heated to gaseous 36 form, and then the UF6 gas is chemically processed to form uranium dioxide (UO2) powder.

37 This powder is then pressed into pellets, sintered (heated) into ceramic form, loaded into 38 Zircaloy tubes, and constructed into fuel assemblies. Fuel is also fabricated for the U.S. Naval 39 Reactors program and for non-power reactors, which typically are small reactors that do not 40 generate electrical power but are used for research, testing, and training. Non-power reactors 41 can include research reactors and reactors used to produce irradiated target materials 42 (e.g., isotopes for medical use). The fuel design varies with the non-power reactor 43 manufacturer, and there is a wide range of fuel assembly designs. Another potential fuel for 44 commercial nuclear power reactors is mixed oxide fuel, which would comprise both UO2 and 45 plutonium dioxide (PuO2). Figure G-3 depicts a typical light-water reactor fuel fabrication facility 46 process.

47 G-9 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 2 Figure G-3 Typical Light Water Reactor Fuel Fabrication Facility 3 Similar to its approach for enrichment facilities, the NRC regulates fuel fabrication facilities 4 under 10 CFR Part 70, including Subpart H. The licensee or applicant is required to provide 5 detailed information on the facilities, equipment, and procedures used, as well as an 6 environmental report that discusses the effects of proposed operations on the health and safety 7 of the workers and the public. These submittals provide a foundation for any related analysis, 8 which includes the requirement, in Subpart H, for the licensee to perform an ISA, identify items 9 relied on for safety, and establish management measures.

10 11 Fuel fabrication facilities have essentially the same types of hazards as enrichment facilities 12 (i.e., chemical, radiological, and criticality hazards). The primary hazard concerns involve the 13 release of UF6, and its subsequent products (HF and UO2F2) in storage or during processing 14 (i.e., vaporization) and inadvertent criticality during the UO2 fuel fabrication processes. In the 15 case of an accident, the workers have a greater chance of being impacted than the public.

16 These facilities generally pose a low risk to the public.

17 18 The key references for enrichment facilities in Section G.2.3 of this appendix data also address 19 fuel fabrication facilities.

20 21 G.2.5 Spent Nuclear Fuel Storage 22 23 SNF refers to uranium-bearing fuel elements that have been used in nuclear reactors and have 24 been removed from the reactor vessel. After the SNF is removed from the reactor, the spent 25 fuel assemblies still generate significant amounts of radiation and heat. There are two 26 acceptable storage methods for SNF: spent fuel pools (SFPs) and, after a suitable cooling time, 27 dry cask storage at an ISFSI. Most SNF is stored in SFPs at individual reactor sites. The SFP 28 maintains at least 20 feet of water above the stored fuel to provide adequate cooling of the SNF 29 and adequate radiation shielding for plant personnel. The assemblies are moved into the SFPs 30 from the reactor along the bottom of water canals, so that the SNF is always shielded to protect 31 workers. The SNF accident scenarios include inadvertent criticality from misloading SNF in 32 high-density ranks or a loss of SFP water inventory that results in SNF heatup.

33 34 The SNF that has sufficiently cooled can be stored in a variety of dry cask storage systems.

35 There are several dry cask storage system designs, some of which can be used for both 36 storage and transportation. The ISFSI is used when the nuclear power plant licensee 37 approaches the capacity limits of its SFP or following the decommissioning of a reactor site.

38 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-10

1 An NRC-certified dry cask is one for which the NRC staff has performed a technical review of its 2 safety aspects and has found the cask to be adequate to store SNF at a site that the licensee 3 has evaluated and shown meets all of the NRCs requirements in 10 CFR Part 72. The dry cask 4 storage license may be site specific or general. Under a site-specific license, an applicant 5 submits a license application to the NRC and the NRC performs a technical review of all the 6 safety and environmental protection aspects of the proposed ISFSI. The NRC has received and 7 is reviewing applications for consolidated interim storage facilities for specific licenses under 8 10 CFR Part 72 and the ISFSIs, as proposed, are not co-located with a nuclear power reactor.

9 10 A general license authorizes a nuclear power plant licensee, without further NRC review, to 11 store SNF in NRC-certified dry casks at a site that is licensed to operate a power reactor under 12 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, or 13 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants. Licensees 14 are required to perform evaluations of their site to demonstrate that the site is adequate for 15 storing SNF in dry casks. These evaluations must show that the cask Certificate of Compliance 16 conditions and technical specifications can be met, including an analysis of earthquake events 17 and tornado missiles. The licensee also must review its security program, emergency plan, and 18 other programs, and make any necessary changes to incorporate the ISFSI at its reactor site.

19 The dry cask storage license contains the technical requirements and operating conditions 20 (e.g., fuel specifications, cask leak testing, surveillance, and other requirements) for the ISFSI 21 and specifies what the licensee is authorized to store at the site.

22 23 The regulation related to backfitting for ISFSIs, 10 CFR 72.62, uniquely requires the NRC to 24 evaluate changes to both occupational and public radiological exposure, whereas all other 25 backfit regulations require the NRC to evaluate changes to public exposure only. Thus, it is 26 possible that a cost-beneficial backfit could be justified that demonstrates substantial worker 27 benefits from averted exposure but results in small or no public benefit for these facilities.

28 29 The primary hazard associated with dry cask storage is the breach of the cask and subsequent 30 radiological release to the environment during handling and movement from the SFP to the 31 ISFSI area. The breach and subsequent release would be primarily an occupational hazard.

32 Once the spent fuel is placed in the dry cask, the radiological hazard is significantly reduced as 33 the low pressure inside the cask would not provide a significant driving force for a radiological 34 release.

35 36 A number of studies, including the following, have considered the hazards and risks associated 37 with SFPs:

38 39

• NUREG-1353, Regulatory Analysis for the Resolution of Generic Issue 82, 40 Beyond-Design-Basis Accidents in Spent Fuel Pools 41 42

• NUREG-1738, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning 43 Nuclear Power Plants 44 45

• NUREG-2161, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the 46 Spent Fuel Pool for a U.S. Mark I Boiling Water Reactor 47 48 To apply for a site-specific license to store SNF, an applicant submits an application to the NRC 49 for review and approval. The application contains information as described in Regulatory Guide 50 3.50, Revision 2, Standard Format and Content for a Specific License Application for an 51 Independent Spent Fuel Storage Installation or Monitored Retrievable Storage Facility, and the G-11 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 NRC reviews the application in accordance with NUREG-1567, Standard Review Plan for 2 Spent Fuel Dry Storage Facilities. The application must address the safety and operational 3 characteristics of the facility, including the site seismic and environmental conditions, the 4 planned storage system, accident analyses, and the radiological impact of normal operations.

5 The following data references provide information for performing the analysis for a regulatory 6 action related to dry cask storage, applications, and staff reviews:

7 8

• NUREG-1536, Revision 1, Standard Review Plan for Spent Fuel Dry Storage Systems 9 at a General License Facility, provides guidance for staff reviews of applications for a 10 Certificate of Compliance submitted for a general license facility. These applications are 11 prepared based on Regulatory Guide 3.61, Standard Format and Content for a Topical 12 Safety Analysis Report for a Spent Fuel Dry Storage Cask.

13 14

• NUREG-1567 provides guidance for staff reviews of applications submitted for 15 commercial ISFSIs, either co-located with a reactor site or away from a reactor site.

16 These applications are prepared based on Regulatory Guide 3.48, Standard Format 17 and Content for the Safety Analysis Report for an Independent Spent Fuel Storage 18 Installation or Monitored Retrievable Storage Installation (Dry Storage).

19 20

• NUREG-1927, Revision 1, Standard Review Plan for Renewal of Specific Licenses and 21 Certificates of Compliance for Dry Storage of Spent Nuclear Fuel, provides guidance for 22 the safety review of renewal applications for specific licenses of ISFSIs and Certificates 23 of Compliance of dry cask storage systems, including the review of time limited aging 24 analyses and aging management programs.

25 26 G.2.6 Low-Level Waste Disposal 27 28 LLW includes radioactively contaminated protective clothing, tools, filters, rags, medical tubes, 29 and many other items. LLW disposal occurs at commercially operated LLW disposal facilities 30 that must be licensed by either the NRC or an Agreement State. The facilities must be 31 designed, constructed, and operated to meet safety standards. The licensee must extensively 32 characterize the site on which the facility is located and analyze how the facility will perform for 33 thousands of years into the future. The United States currently has four LLW disposal facilities 34 that accept various types of LLW; all are located in and regulated by Agreement States. The 35 Low-Level Radioactive Waste Policy Amendments Act of 1985 encouraged the States to enter 36 into compacts that would allow them to dispose of LLW at a common disposal facility. The NRC 37 provides additional information on the LLW disposal facilities in NUREG-1350, NRC 38 Information Digest.

39 40 The following principal regulations govern LLW:

41 42

• The general provision in 10 CFR 20.2002, Method for Obtaining Approval of Proposed 43 Disposal Procedures, allows for other disposal methods, different from those already 44 defined in the regulations (see 10 CFR 20.2001), if doses are maintained as low as is 45 reasonably achievable and within the dose limits of 10 CFR Part 20.

46 47

• In 10 CFR Part 61, the NRC establishes the licensing procedures, criteria, and terms 48 and conditions for land disposal of radioactive waste.

49 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-12

1 The NRC provides an additional resource for further understanding the history of LLW 2 operations and the associated regulations in NUREG/BR-0121, Regulating the Disposal of 3 Low-Level Radioactive Waste: A Guide to the Nuclear Regulatory Commissions 4 10 CFR Part 61. NUREG-1853, History and Framework of Commercial Low-Level 5 Radioactive Waste Management in the United States: ACNW [Advisory Committee on Nuclear 6 Waste] White Paper, also provides background information on the disposal of commercial LLW.

7 8 The primary hazard from LLW disposal is the radiological release to the environment from the 9 disposal site (e.g., resulting from the failure of the disposal liner or barrier).

10 11 In addition to the information in the license application and associated staff safety evaluation, 12 the following data references are useful in performing a regulatory analysis associated with LLW 13 disposal:

14 15

• NUREG-1573, A Performance Assessment Methodology for Low-Level Radioactive 16 Waste Disposal Facilities: Recommendations of NRCs Performance Assessment 17 Working Group, provides information and recommendations on performance 18 assessment methodology as it relates to the objective of radiological protection of the 19 general public in accordance with 10 CFR 61.41, Protection of the General Population 20 from Releases of Radioactivity.

21 22

• NUREG-1200, Revision 3, Standard Review Plan for the Review of a License 23 Application for a Low-Level Radioactive Waste Disposal Facility, contains the staff 24 guidance on reviewing applications to construct and operate LLW disposal facilities.

25 26

• NUREG-1300, Environmental Standard Review Plan for the Review of a License 27 Application for a Low-Level Radioactive Waste Disposal Facility: Environmental Report, 28 provides NRC staff with guidance on performing the environmental review of an 29 application to construct and operate a LLW disposal facility.

30 31

• NUREG-1199, Revision 2, Standard Format and Content of a License Application for a 32 Low-Level Radioactive Waste Disposal Facility, is the companion guidance for 33 prospective applicants for a license to dispose of LLW pursuant to 10 CFR Part 61.

34 35

• NUREG-1241, Licensing of Alternative Methods of Disposal of Low-Level Radioactive 36 Waste, provides NRC staff with guidance on specific methods of disposal under 37 consideration as alternatives to shallow land burial.

38 39 In February 2016, the U.S. Department of Energy (DOE) issued a final Environmental Impact 40 Statement (DOE, 2016) to consider the potential environmental impacts associated with 41 constructing and operating a new facility or facilities, or using an existing facility, for the disposal 42 of greater-than-Class C (GTCC) 5 and GTCC-like waste anticipated to be generated through 43 2083. In addition, on October 23, 2018, the DOE issued an environmental assessment for the 44 disposal of GTCC and GTCC-like waste at the Waste Control Specialists land disposal facility in 45 Andrews County, Texas (DOE, 2018).

5 The NRC divides LLW into three classes: Class A, Class B, and Class C. The gradation of these classes is based on the radiological hazard as determined by the quantity and type of radionuclides permitted in each class, as further delineated by concentrations of certain radionuclides. Therefore, Class A waste is the least hazardous and Class C waste is the most hazardous. In addition, some waste streams have radionuclide concentrations exceeding the limits for Class C waste and as such are referred to as greater-than-Class C (GTCC) waste.

G-13 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 G.2.7 Geological Repository Risks 2

3 This section provides the analyst with (1) the risks associated with the development of a 4 repository, (2) a quantitative perspective on the overall risks for repository development, and 5 (3) the risks associated with specific activities during repository development (e.g., handling, 6 transportation, and disposal). The development of a repository involves the handling, 7 transportation, storage, and disposal of SNF primarily in sealed canisters. Thus, the risks for 8 the development of a repository tend to be significantly less than for activities associated with 9 other nuclear facilities (e.g., nuclear power reactors) that rely on active safety systems as 10 compared to the passive safety nature of a sealed canister. However, the handling of SNF 11 relies on active safety systems, and the transportation of SNF results in the potential for 12 exposure to larger segments of the population than would occur near a specific site. Although 13 the development time for a repository is long (on the order of 100 years or more), the risks 14 associated with specific activities of repository development (e.g., emplacement of waste in a 15 repository) are expected to occur over a much shorter time period.

16 17 Estimating the risks related to the development of a repository includes consideration of three 18 major activities:

19 20 (1) transporting the SNF from individual sites to a repository 21 22 (2) constructing a repository, operating the surface facilities, and emplacing the wastes 23 (repository pre-closure period) 24 25 (3) performance of a repository after it has been permanently closed (repository 26 post-closure period) 27 28 The analyses for repository risks in this section are based on an evaluation of various designs 29 and site conditions (e.g., population densities, transportation routes) that are considered to be 30 applicable to a range of potential sites in the United States. This approach provides a generic 31 analysis that would be useful for application at a variety of potential sites and a comparison with 32 other published studies on the risks from the transportation and disposal of radioactive wastes.

33 Assumptions about the basis for the risk estimates are provided to add the appropriate context 34 for understanding the risks in this section. Further evaluations may be necessary for site 35 conditions and designs that vary significantly from those used for estimating the risks presented 36 in this section.

37 38 G.2.7.1 Transportation of Spent Nuclear Fuel to a Repository 39 40 SNF is stored at numerous facilities around the country. A transportation campaign would be 41 conducted to ship SNF from these sites to a repository; this campaign is expected to include 42 both rail and truck transport, and result in radiation exposures from the loading of transportation 43 casks and exposures along the transportation routes from both routine (i.e., incident-free) 44 transportation and potential accidents.

45 46 Radiological Impacts from Loading Spent Nuclear Fuel 47 48 Impacts to workers during loading activities occur during the loading of SNF into canisters, 49 during the loading of canisters into rail casks, and, at some sites, during the loading of SNF into 50 truck casks. DOE/EIS-0250F-S1, Final Supplemental Environmental Impact Statement for a 51 Geological Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-14

1 Waste at Yucca Mountain, Nye County, Nevada (DOE, 2008), states that a transportation, 2 aging, and disposal (TAD) canister is similar to a dry storage canister in appearance, capacity, 3 and the operational procedures that would be in use for loading. Therefore, based on historical 4 information for the loading of SNF into TAD canisters at commercial generator sites, the DOE 5 estimates that the average radiation occupational dose for loading commercial SNF fuel into 6 canisters is 0.400 person-rem per canister (DOE, 2008, page G-2). The DOE also estimates 7 that the average radiation doses for workers during (1) loading of uncanistered SNF into truck 8 casks and loading the casks onto trailers is 0.432 person-rem per cask, and (2) transferring 9 canisters from storage, loading canisters into rail casks, and loading the casks onto railcars is 10 0.663 person-rem per cask (DOE, 2008, Table G-2). The DOE used a specific number of cask 11 shipments (i.e., 6,499 TAD canisters in rail casks, 307 dual-purpose canisters [DPCs] in rail 12 casks, and uncanistered SNF in 2,650 truck casks) to accommodate 63,000 metric tons of 13 heavy metal (MTHM) of SNF from commercial reactors. Table G-1 summarizes the data used 14 to estimate radiation doses to workers for loading.

15 16 Table G-1 Data Used to Estimate Radiation Doses for Loading Number of Canisters or Type Operation Radiation Dose Casks for Operation Load commercial SNF into 0.400 person-rem Rail Cask 6,499 canistersa canister per canister Transfer canister from storage, Transfer 0.663 person-rem load into rail cask, load rail 6,806 casksb,c Rail Cask per cask cask onto rail car Load uncanistered SNF into Truck 0.432 person-rem truck cask, load truck cask 2,650 casksb Cask per cask onto truck trailer 17 a Includes only TAD canisters.

18 b Includes commercial SNF only.

19 c The estimate is based on 6,499 casks of commercial SNF containing TAD canisters and 307 casks of commercial 20 SNF containing DPC.

21 Source: DOE, 2008, Table G-2 22 23 Table G-2 provides the collective worker dose for the loading campaign, which is obtained by 24 multiplying the number of canisters and casks by the appropriate cask or canister collective 25 dose rate for the workers. The DOE also estimates that a maximally exposed worker 26 conducting loading activities at generator sites would receive an exposure rate of 500 millirem 27 (mrem) per year based on the DOEs administrative control limit of 500 mrem per 28 year (DOE, 2017).

29 30 Radiation doses to members of the public near generator sites could occur because of the 31 venting of radioactive gases during the handling of SNF in SFPs and the potential release of 32 surface contamination from dry transfer casks. The DOE estimates that the population dose to 33 members of the public within 16 kilometers (10 miles) of the generator sites would be 34 2.9 person-rem over the duration of loading operations. The probability of a latent cancer 35 fatality based on the estimated dose would be 0.0017, or about 1 chance in 600 that one 36 member of the exposed population would develop a latent cancer fatality. The DOE also 37 estimated that a maximally exposed individual located 800 meters (0.5 mile) from the generator 38 site would receive an exposure of 7.7x10-6 rem (DOE, 2008, page 6-12). Table G-2 presents 39 the estimated exposures for workers and the public for the loading activities over a 50-year 40 period and a shipment of 63,000 MTHM of SNF.

41 G-15 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 Table G-2 Estimated Radiological Impacts from Loading Operations Collective Dose Latent Cancer Maximally Exposed Exposure Group (person-rem) Fatalities Individual (mrem/yr)

Involved Workersa 8,300 5.0 500 Public 2.9 0.0017 0.0077 2 a Radiation exposures from loading operations would not occur among noninvolved workers because these workers 3 would not be exposed to radiation from the operations.

4 5 Impacts from Transportation of Spent Nuclear Fuel to the Repository 6

7 Routine (i.e., accident-free) transportation is estimated to result in a low dose to the public. In 8 particular, in Figure 6-3 of NUREG-2125, Spent Fuel Transportation Risk Assessment Final 9 Report, the NRC estimates the dose to the maximally exposed individual during transportation 10 to be less than 0.001 mrem per shipment by either rail or truck. In NUREG-2125, the NRC 11 estimated the risks of SNF transportation for a variety of shipping routes (both rail and truck),

12 primarily from locations in the eastern United States to locations in the western United States, 13 that could result in potential doses to workers and the public. The NRC estimated the average 14 collective public dose from routine truck transport to be 0.14 rem per shipment (NRC, 2014a, 15 Figure 6-1). The collective public dose includes doses to the population along the route (6 16 percent of collective dose), doses to occupants of vehicles sharing the route (38 percent of 17 collective dose), and doses to the public at stops (56 percent of the collective dose) (NRC, 18 2014a, page 133). The NRC estimated the average collective dose for workers for routine truck 19 shipments to be 0.10 rem per shipment, which includes doses to the vehicle crew and other 20 workers (e.g., escorts, inspectors, truck stop workers) (NRC, 2014a, Figure 6-1).

21 22 NUREG-2125 also provides radiation dose estimates for two types of rail packagesa cask 23 using lead gamma shielding and a cask using steel gamma shielding. The results reported in 24 Table G-3 of this appendix are for the lead package because of the larger doses 25 (i.e., approximately 30 percent larger). The NRC estimated the maximally exposed individual 26 during routine rail transport to receive less than 0.001 mrem per shipment, similar to the 27 estimate for a truck shipment (NRC, 2014a, Figure 6-3); however, the average collective doses 28 were less for rail shipments than for truck shipments. The estimated average collective public 29 dose during rail shipment using a lead cask is estimated to be 0.026 rem per shipment, and the 30 average collective worker dose is estimated to be 0.041 rem per shipment, for a total collective 31 dose of 0.67 rem per shipment (NRC, 2014a, Figure 6-2).

32 33 Table G-3 presents the collective dose for a transportation campaign, with the number of 34 shipments consistent with the loading campaign discussed below (i.e., 6,806 rail shipments and 35 2,650 truck shipments).

36 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-16

1 Table G-1 Estimated Worker and Public Impacts from Shipping Commercial SNF to 2 the Repository over a 50-Year Period Total Average Annual Average Exposure Group Collective Dose Collective Dose (person-rem) (person-rem/yr)

Workers (Rail) 279 5.6 Workers (Truck) 265 5.3 Total Workers 544 10.9 Public (Rail) 177 3.5 Public (Truck) 371 7.4 Total Public 548 10.9 3

4 Table S-4 in 10 CFR Part 51 provides a range of doses for exposed individuals per reactor year.

5 The values in Table S-4 would result in larger doses if they were used to generate the collective 6 doses presented in Table G-3 of this appendix. The values in Table G-3 are based on recent 7 analyses and represent realistic doses compared to the more conservative values used in 8 Table S-4.

9 10 Loading and Handling Canisters and Transportation Accidents 11 12 The SNF presents a significant hazard from direct radiation for individuals that are in close 13 proximity to SNF and from internal exposure if radioactive material is released into the 14 environment. Thus, the loading and transportation of SNF is performed under requirements that 15 are intended to minimize the occurrence of accidents and the consequences resulting from an 16 accident (e.g., single-failure-proof cranes for limiting the probability of drop events, HEPA filters 17 for buildings where handling of SNF assemblies occur, storage and transportation casks 18 designed to withstand a variety of potential accidents). Compliance with the NRCs 19 requirements in 10 CFR Part 71 and 10 CFR Part 72 ensures that the risks associated with 20 loading and transportation accidents are low.

21 22 The NRC has considered a comprehensive set of initiating events for estimating the risks of 23 handling, transfer, and storage of dry cask storage systems in NUREG-1864, A Pilot 24 Probabilistic Risk Assessment of a Dry Cask Storage System at a Nuclear Power Plant.

25 NUREG-1864 estimates the probability and consequences of a loading accident based on 26 considerations such as stresses on the fuel cladding and canister as a function of drop height, 27 breach of the canister sealed lid, and reliability of secondary containment to isolate accidental 28 releases during handling (e.g., containment within a handling facility). The largest identified risk 29 occurs for a 19-foot drop during loading and handling of a canister that results in failure of the 30 cladding and a 0.28 probability that the canister is breached (NRC, 2007b, Table 19, item 20).

31 In the event of a breach, the average individual dose for those living within 16 kilometers 32 (10 miles) of the facility where SNF is being loaded is 600 mrem 6 with a peak individual dose of 33 185 rem for individuals within 1.2 and 1.6 kilometers (0.75 and 1 mile) of the facility (NRC, 34 2007b, Table 18 and Table E.3). However, the probability of such a release occurring is low.

6 NUREG-1864 reports consequences in terms of latent cancer fatality. A factor of 5 x 10-4 latent cancer fatality per rem of exposure is used to derive dose exposures from the latent cancer fatalities.

G-17 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 NUREG-1864 provides estimated low likelihoods for a 19-foot drop that breaches the canister 2 (i.e., 5.6 x 10-5 per lift x 0.28). Combined with a low probability for the subsequent failure of the 3 secondary containment building to isolate the released waste after it escapes the canister 4 (i.e., 1.5x10-4 per demand), the overall probability of release is 2.4x10-9 per lift. 7 The total risk 5 from loading all of the SNF is the risk for a single lift multiplied by the total number of canisters 6 that would be loaded over the transportation campaign (i.e., approximately 10,000 canisters).

7 Thus, the total risk from a 19-foot drop during the transportation campaign for those individuals 8 residing within 16 kilometers (10 miles) of the facility where SNF is loaded is calculated as 9 follows:

10 11 2.4x10-9 (probability of release per lift) 12 x 600 mrem (average individual dose given release occurs) 13 x 10,000 (number of lifts in transportation campaign) 14 = 0.014 mrem (total risk from transportation campaign) 15 16 These probability weighted doses are small, partly because the canister drop is assumed to 17 occur inside a building that would provide secondary containment to minimize potential releases 18 in accordance with 10 CFR Part 50. If a canister drop occurs outside of a 10 CFR Part 50 19 building (i.e., at the ISFSI pad) the probability of a release occurring would increase to 1.6x10-5 20 or higher because a secondary containment building is not available and a single-failure-proof 21 crane may not be used (NRC, 2007b, Table 19, note 13). The Electric Power Research 22 Institutes Report 1002877, Probabilistic Risk Assessment (PRA) of Bolted Storage Casks:

23 Updated Quantification and Analysis Report, also contains information about loading, 24 transportation, and storage of dry casks.

25 26 NUREG-2125 also provides the evaluated risks of transportation accidents. Transportation 27 packages for radioactive materials such as SNF are designed to maintain their integrity in 28 severe accidents. Almost all spent fuel casks are shipped without incident. However, even this 29 routine, incident-free transportation causes radiation exposures because all loaded spent fuel 30 casks emit some radiation (Table G-3 presents exposures for routine [accident free]

31 transportation) whether or not radioactive material is released from the cask. This is because 32 transportation cask shielding attenuates but does not eliminate the radiation that spent fuel 33 emits. In NUREG-2125, the NRC concluded that the collective dose risk from accidents 34 involving a release of radioactive material and a loss of lead shielding are negligible compared 35 to the risk from a no-release, no-loss-of-shielding accident because of the low likelihood of 36 accidents that affect the performance of the transportation cask (NRC, 2014a, page xxxvii).

37 Given the robust design of the waste package, NUREG-2125 concludes that (1) only about 1 in 38 1,000 trips would result in an accident, (2) if an accident occurs, only about 1 in 2,000 accidents 39 is more severe than the regulatory accident conditions, and (3) if an accident is more severe 40 than the regulatory accident conditions, only about 3 in 1,000,000 will result in either a loss of 41 gamma shielding or a release of radioactive material (NRC, 2014a, page xxxiv). For example, 42 [I]f there were an accident during a spent fuel shipment, there is only about a one-in-a-billion 43 chance that the accident would result in a release of radioactive material (NRC, 2014a, Figure 44 PS-8 and page 139).

45 46 NUREG-2125, the NRC evaluated a variety of potential rail and truck accident scenarios over 47 16 truck routes and 16 rail routes and estimated that the average collective dose risk 7 If a release of radioactive material were to occur inside the secondary containment following a transfer cask drop, three distinct functions are designed to occur to (1) detect radioactive material, (2) isolate the secondary containment, and (3) operate the standby gas treatment system.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-18

1 (i.e., probability weighted dose) is less than 0.003 person-rem (NRC, 2014a, Tables 6-4 and 2 6-5). The average collective dose for accidents with no release and no shielding loss for the 3 rail-lead cask is 0.0022 person-rem and for the truck cask is 0.0024 person-rem. The most 4 significant parameters contributing to the risk are the accident frequency and the length of time 5 the intact transportation package sits at the accident location. The significant parameter for the 6 radiological effect of the accident is not the amount or rate of radiation released but the time of 7 exposure to the immobilized cask at the accident site. The total average collective dose risk for 8 a transportation campaign (i.e., 6,806 rail shipments and 2,650 truck shipments) is 9 21.3 person-rem, which is approximately 25 times less than the collective public dose estimated 10 for routine (incident-free) transportation shown in Table G-3 of this appendix.

11 12 G.2.7.2 Repository Construction, Operation, and Closure 13 14 The development of a repository is generally characterized by three periods that are associated 15 with separate regulatory decisions: (1) construction (decision to grant or deny a construction 16 authorization), (2) operations (decision to allow the receipt of radioactive waste at the repository 17 and the start of emplacement operations), and (3) closure (decision to permanently close the 18 facility). These three periods are typically referred to as the pre-closure period. After 19 permanent closure, the post-closure period begins, and the repository relies on passive safety 20 barriers. Although the DOE is required to provide continual oversight of the repository after 21 permanent closure, this DOE oversight is not relied on for compliance with the post-closure 22 dose limits.

23 24 Although the decisions occur at a distinct time, the periods of construction, operations, and 25 closure will overlap in time to varying degrees. For example, the licensee can apply for a 26 license to begin operations (i.e., receive waste) once the repository is ready to begin initial 27 operations. However, all emplacement drifts, 8 hereafter referred to simply as drifts, and some 28 surface facilities would not necessarily be completed, and thus additional surface facilities and 29 repository drifts would continue to be developed during the initial operations.

30 31 The development of a repository has the potential to expose individuals to radiation from both 32 human-made radionuclides in the waste and natural radionuclides associated with the 33 underground facility (common in hard rock mines with natural radioactive materials). Exposures 34 from the release of radionuclides into the air can affect all workers at the site and members of 35 the public residing offsite, whereas an external dose is limited to those individuals engaged in 36 activities near radioactive materials (e.g., handling operations for waste packages). In 37 considering the risks from the development of a geological repository, it is useful to estimate the 38 risks for each of the phases of development because of the potentially long time 39 (e.g., 100 years) over which these activities will occur and the different types of activities 40 associated with a specific phase. It is also useful to represent the risks for different exposure 41 groups because the potential for significant variations in risks among exposure groups.

42 43 The DOE developed an environmental impact statement (EIS) (DOE, 2008) for a geological 44 repository at Yucca Mountain. Regardless of the site, the development of a repository will require 45 a number of similar activities (e.g., handling and emplacement of waste packages, potential 46 repackaging of waste, excavation of drifts). Therefore, the EIS is used to assist in quantifying the 47 potential risks from the development of a repository. The risks are generally applicable to any 8 In mining terminology, a drift is a horizontal underground passage. As used in this appendix, an emplacement drift represents the horizontal excavations for the emplacement of radioactive waste packages.

G-19 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 repository site that has similar activities. In order to estimate the potential risks, the DOE made 2 numerous assumptions. These assumptions include the length of time for development to occur 3 and the amount of waste that is expected to be emplaced at the repository. The EIS specifies an 4 operating period of 105 years, beginning with construction and ending with the permanent closure 5 of the repository, and a statutory limit of 70,000 MTHM, of which 63,000 MTHM is commercial 6 SNF.

7 8 The EIS provides dose estimates for involved workers (i.e., workers performing physical work 9 such as constructing, operating, monitoring, and closing the repository), noninvolved workers 10 (i.e., managerial, technical, supervisory, and administrative personnel onsite), a maximally 11 exposed offsite individual residing near the facility, and a potential population of 12 117,000 individuals residing within 84 kilometers (52 miles) of the facility. The EIS also 13 estimates the radiation exposure from both SNF and the radioactivity from natural sources.

14 15 Exposure to Natural Sources of Radiation during the Preclosure Period 16 17 The site of an underground facility may contain certain naturally occurring radionuclides in the 18 walls of the host rock that result in an external dose to individuals that spend a significant 19 amount of time in the underground facility. The underground host rock may also release radon 20 gas that will cause internal exposure to individuals in the underground facility. In addition, the 21 ventilation of the radon gas to the surface environment results in exposure to individuals on the 22 surface (both to onsite workers and those offsite). Radiation exposures from these natural 23 sources of radiation will occur as the repository drifts begin to be excavated and before the 24 presence of any radioactive waste at the site. These radiation exposures will continue as long 25 as the underground facility is ventilated or workers are present in the underground facility.

26 Naturally occurring radon would account for more than 99.8 percent of the radiological impacts 27 to the offsite public (DOE, 2008, page A-4).

28 29 The release of radon from the underground facility will begin at the start of construction and 30 continue to increase until all the drifts are completed, after which the release should remain 31 constant until the drifts are permanently closed and the ventilation has ended. The DOE has 32 estimated that annual radon releases would steadily increase linearly over the 27-year period 33 needed to complete all repository drifts. They would then remain at a constant annual release 34 through ventilation of the underground facility of 4,700 curies (Ci) until the beginning of the 35 closure activities. At the start of a repository project only construction activities would take 36 place, and thus the only exposure would be from natural radionuclides such as radon.

37 Table G-4 estimates potential radon releases over the different time periods of repository 38 development.

39 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-20

1 Table G-2 Atmospheric Release of Radon through Ventilation of the Underground 2 Facility during Repository Development Annual Radon Release (Ci/yr)

Time Period Activities/Phase Annual (years) Initial Ending Average Construction Only 0-5 0 870 425 6 - 27 870 4,700a 2,775 Construction and Emplacement 28 - 55 4,700 4,700 4,700 Monitoring 56 - 95 4,700 4,700 4,700 Closure 96 - 105 4,700 0 2,350 3 a The radon release peaks at 4,700 Ci per year in year 27 and remains at that rate until the initiating of closure 4 activities in year 96.

5 Source: DOE, 2008, Figure D-1 6

7 Radon that is released through ventilation has the potential to result in exposures to 8 underground and surface workers as well as members of the public residing offsite. The DOE 9 identified workers as either involved workers or noninvolved workers, and estimated the 10 potential radiation exposures for the surface and subsurface locations within the geological 11 repository operations area. Radon is assumed to disperse sufficiently over the geological 12 repository operations area such that the maximum exposure for surface workers (both involved 13 and noninvolved) would be similar. Public exposures are dependent on the location of the 14 members of the public. A repository will have an area where public access is restricted, and this 15 would vary from site to site. The DOE determined that the location for the maximally exposed 16 offsite member of the public, based on air dispersion of atmospheric releases from the Yucca 17 Mountain repository, is just outside the restricted area (approximately 18 kilometers [11 miles]

18 from the subsurface facilities) in the south-southeast direction from the repository. In 19 determining exposure for the maximally exposed individual, the DOE assumed a hypothetical 20 individual resided continuously for 70 years at this location and determined a population dose 21 based on an estimated population distribution for the year 2067 that resulted in a population of 22 approximately 117,000 individuals residing within 84 kilometers (52 miles) of the facility 23 (DOE, 2008, page D-12). Although the maximally exposed individual is assumed to reside 24 continuously outside the controlled area in the prevailing downwind direction from the repository 25 at a location where the estimated dose is a maximum, the estimated population dose provides a 26 collective dose for all individuals assumed to reside within 84 miles of the facility where the dose 27 for a given individual within the population of 117,000 individuals will vary based on with the 28 individuals distance from the repository as well as the location relative to the direction of the 29 wind. Table G-5 provides unit dose conversion factors for evaluating radon exposures for 30 workers and the public present on the land surface.

31 G-21 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 Table G-5 Internal Dose Conversion Factors for the Maximally Exposed Individuals 2 and the Offsite Population Present on the Land Surface Based on a Release 3 of 1 Ci of Radon Maximally Exposed Individual Dose Conversion Factors Offsite Population Involved Noninvolved Noninvolved Collective Dose Offsite Conversion Factor Surface Subsurface Surface Individual (person-rem per Ci)

Worker Worker Worker (mrem per Ci)

(mrem per Ci) (mrem per Ci) (mrem per Ci) 0.0010 0.0011 0.00097 0.0016 0.033 4 Source: DOE, 2008, Table D-5 5

6 Subsurface workers have the potential to receive larger external exposures because the 7 proximity to radionuclides in drift walls. In addition, exposure to a higher concentration of radon 8 in the underground facility results in higher internal exposure from radon in the air. The DOE 9 assumed that the subsurface involved worker is exposed to an average radon concentration of 10 5.8 picocuries per liter of air, resulting in an internal dose of 70 mrem per year (DOE, 2008, 11 page D-19). The subsurface involved worker would also receive an external dose of 50 mrem 12 per year from radionuclides in the drift walls as a result of spending 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> underground 13 during the work year (DOE, 2008, page 4-63). The subsurface noninvolved worker would not 14 receive a significant external dose because of the limited amount of time spent in close 15 proximity to the drift walls. The radon dose is estimated differently to account for the 16 noninvolved worker who is subject to radon levels in the south portal rather than other areas of 17 the subsurface containing higher concentrations of radon. As presented in Table G-6, the DOE 18 estimated a larger unit exposure rate for the noninvolved subsurface worker during the 19 construction-only period (i.e., initial 5 years of development) versus all the other times when the 20 radon dose is comparable to the radon dose estimated for the surface workers (DOE, 2008, 21 Table D-5). The higher rate during the construction-only period is because all ventilation 22 exhaust exits through the south portal of the underground facility, whereas ventilation after the 23 construction-only period is exhausted through six ventilation shafts that replace the use of the 24 south portal for the air exhaust (DOE, 2008, pages D-12 and 13).

25 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-22

1 Table G-6 Internal and External Dose Conversion Factors for the Maximally Exposed 2 Worker in the Subsurface from Natural Sources of Radiation throughout 3 Repository Development Construction-Only All Other Periods Subsurface Worker Period (years 0-5) (years 6-105)

Involved Worker 50 mrem/year 50 mrem/year (External Dose)a Involved Worker 70 mrem/year 70 mrem/year (Internal Dose)b Noninvolved Worker 0.066 mrem/Ci radon 0.0011 mrem/Ci radon (Internal Dose)c 4 a Source: DOE, 2008, page D-19 5 b Source: DOE, 2008, page 4-63 6 c Source: DOE, 2008, Table D-5 7

8 Table G-7 provides the maximum individual (i.e., workers and public) dose from natural sources 9 of radiation over the three major periods of repository development: (1) construction and 10 emplacement (the initial 55 years), (2) monitoring (40 years following the end of emplacement),

11 and (3) closure (the last 10 years of the 105-year development period). The maximum annual 12 dose to workers and the public present on the surface is estimated by multiplying the maximum 13 radon release in Table G-4 for a given time period by the appropriate dose conversion factor in 14 Table G-5. For example, the offsite individual has a maximum dose of 7.5 mrem per year 15 during the monitoring period based on an annual release of 4,700 Ci of radon and a conversion 16 factor of 0.0016 mrem per curie of radon released. The maximum annual exposure for the 17 subsurface involved worker is based on the summation of the internal exposure (70 mrem per 18 year) and the external exposure (50 mrem per year) in Table G-5. Maximum annual individual 19 exposure for the noninvolved subsurface worker is estimated using the same approach for the 20 surface workers, except Table G-6 provides the dose conversion factors. The offsite maximally 21 exposed individual has a larger dose than the noninvolved surface worker. The DOE 22 considered the maximally exposed offsite individual to be a hypothetical member of the public 23 residing continuously at the site boundary. Alternatively, the onsite worker is present for 24 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per year (i.e., the offsite resident at the site boundary is exposed to radon release 25 for a longer time during the year than the site worker).

26 G-23 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 2 Table G-7 Maximum Annual Individual Dose for Workers and the Public from 3 Naturally Occurring Radionuclides during the Development of a Repository Annual Individual Dose for Workers and the Public (mrem/yr)

Exposure Construction Construction and Monitoring Closure (Years Group Only Emplacement (Years 56-95)96-105)

(Years 0-5) (Years 6-55)

Involved Worker 120 120 120 120 (Subsurface)

Involved Worker 0.43 3.9 4.7 2.8 (Surface)

Noninvolved Worker 29 8.5 10 5.2 (Subsurface)

Noninvolved 0.41 3.7 4.6 2.3 Worker (Surface)

Offsite Individual 0.68 6.2 7.5 3.8 4 Note: Rounded to two significant figures.

5 6 Collective population dose over the entire repository development period is also estimated for 7 the worker groups (e.g., involved subsurface workers and noninvolved surface workers) and the 8 public residing within 84 kilometers (52 miles) of the repository. The DOE estimated that 9 approximately 86,000 worker-years would be required to develop the repository, of which 10 approximately 66,000 worker-years are for construction and emplacement, 10,000 worker-years 11 are for the monitoring period, and 10,000 worker-years are for the closure period. Table G-8 12 breaks down the workforce during the 105 years of development (DOE, 2008, Figure D-2).

13 Table G-9 presents the collective population dose for the worker groups and the public from 14 natural sources of radiation. The collective population dose for the worker groups is estimated 15 by multiplying the annual individual dose for workers in Table G-7 by the appropriate number of 16 worker years in Table G-8 for the worker groups. The collective population dose for the offsite 17 public is estimated by multiplying the annual average radon release for each time period 18 (Table G-4) by the internal dose conversion for the offsite population dose (Table G-5) and the 19 appropriate number of years for each time period to determine the person-rem for each of the 20 time periods.

21 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-24

1 Table G-8 Population Size for Workers (Worker-Years) during Repository 2 Development Construction and Emplacement Monitoring Closure Worker Group years 0-5 years 6-55 years 56-95 years96-105 Surface Workers Construction 5,000 5,000 NA NA Involved NA 30,000 5,000 5,000 Noninvolved 2,000 15,000 1,000 2,000 Subsurface Workers Construction 300 4,000 NA NA Involved NA 4,000 3,000 3,000 Noninvolved NA 200 400 800 Overall Totals 7,300 58,200 9,400 10,800 Overall Annual 1,500 1,200 230 1,100 Averagea 3 a Overall annual average values are rounded to two significant figures.

4 Source: DOE, 2008, Figure D-2 5

6 Table G-3 Collective Population Dose for Workers and the Public in Person-Rem from 7 Natural Sources of Radiation during the Development of a Repository Construction Construction Exposure and Monitoring Closure Only Group Emplacement (Years 56-95) (Years96-105)

(Years 0-5)

(Years 6-55)

Surface 19 2 NA NA Construction Onlya (ends year 32)

Surface NA 120 24 14 (Involved Worker)

Surface 1 56 5 5 (Noninvolved Worker)

Subsurface 480 36 NA NA Construction Onlya (ends year 32)

Sub-surface NA 480 360 360 (Involved Worker)

Subsurface NA 2 4 4 (Noninvolved Worker)

Worker Totals 39 1,200 390 380 Offsite Population (117,000 within 2 6,400 6,200 780 84 kilometers) 8 a Considered an involved worker for the purpose of estimating dose.

9 b Fractional values rounded to nearest whole number and no more than two significant figures used.

10 G-25 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 Exposure to Radioactive Waste during Routine Operations 9 2

3 After radioactive waste is received at the repository, potential exposures can occur from 4 the release of radionuclides into the atmosphere resulting in an internal dose and an external 5 dose for individuals conducting activities in close proximity to waste packages. As discussed 6 above for radon, radionuclides released into the atmosphere can impact all workers at the site 7 as well as offsite individuals. External dose, however, would be limited to the involved workers.

8 The DOE considered two major sources for atmospheric releases from SNF: (1) releases that 9 may occur when waste is being handled at a wet handling facility (WHF) and (2) releases from 10 surface contamination on the waste canisters sitting on the surface aging pad or the waste 11 packages that are emplaced underground but before permanent closure of the facility. The 12 DOE estimated potential exposures for a variety of onsite locations associated with the 13 repository facilities to account for dispersion affecting the variation in concentrations of 14 atmospheric release of radioactive materials that a potential onsite worker might experience 15 over the repository site (BSC, 2008a).

16 17 Individual Doses from Spent Nuclear Fuel 18 19 The DOE estimated potential exposures to atmospheric releases of radioactive material for 20 30 distinct surface locations of the repository. Exposure to atmospheric releases of SNF are 21 estimated to be very low for workers, with the exception of involved workers at the WHF where 22 uncanistered SNF would be handled. In particular, the DOE estimated that the largest internal 23 annual dose to workers from the combined atmospheric releases (i.e., releases from the WHF, 24 the aging pad, and the sub-surface) would be 15.3 mrem per year, whereas potential exposures 25 at the other 29 locations would vary from 0.0248 to 0.288 mrem per year (BSC, 2008d, Table 5).

26 The releases from the WHF are the largest contributor to worker exposures. The potential dose 27 for the WHF workers from the WHF releases is 13.7 mrem per year (BSC, 2008a, Table 12).

28 29 As for the exposures from natural radioactive sources released from the repository to the 30 atmosphere, the DOE considered the maximally exposed offsite individual to be a hypothetical 31 member of the public residing continuously at the site boundary approximately 19 kilometers 32 (12 miles) from the surface facilities of the repository (DOE, 2008). The DOE estimated that the 33 maximally exposed individual would receive a maximum dose of 0.018 mrem per year as a 34 result of SNF releases to the atmosphere from individual doses of (1) 0.003 mrem per year from 35 the WHF release (assuming an annual throughput of 300 MTHM that represents 10 percent of 36 the overall throughput of 3,000 MTHM that is assumed to be handled at the WHF),

37 (2) 0.012 mrem per year from aging pad releases during each year of operations, and 38 (3) 0.003 mrem per year from subsurface facility releases that would occur each year until final 39 closure of the facility (DOE, 2008, Table D-5). Table G-10 summarizes the information on 40 internal dose from atmospheric releases.

41 9 These estimates deal with commercial spent fuel and do not include risk from defense waste (typically referred to as HLW) that DOE would have also disposed at the Yucca Mountain geologic repository.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-26

1 Table G-4 Annual Individual Internal Dose for Atmospheric Releases of SNF during 2 Operations Annual Dose (mrem/yr)a Exposure Group Emplacement Monitoringb Closureb (Years 6-55) (Years 56-95) (Years96-105)

WHF Workers 15 NA NA All Workers 0.025 to 0.29 0.011 to 0.040c 0.011 to 0.040c (Except WHF Workers)

Offsite Individual 0.018 0.003 0.003 (Maximally Exposed) 3 a Values rounded to two significant figures.

4 b Subsurface releases only during these periods.

5 c Values based on BSC, 2008a, Table 14.

6 7 Certain involved workers performing activities sufficiently close to SNF (e.g., handling activities) 8 can receive a measurable external dose. The DOE assumed an overall annual throughput of 9 500 casks (3,000 MTHM) per year with design-basis source terms to estimate the external dose 10 to workers. However, the WHF only processes 10 percent of the overall throughput or 11 300 MTHM. These annual individual external doses for surface workers ranged from 200 to 12 1,300 mrem per year for operators and 200 to 800 mrem per year for health physics technicians 13 (BSC, 2008b, Table 1.0). The average external dose to these involved surface workers was 14 480 mrem per year for operators and 358 mrem per year for health physics technicians 15 (BSC, 2008b, Table 1). Individual external exposures for the subsurface involved workers 16 ranged from 100 to 209 mrem per year during emplacement operations, 120 to 204 mrem per 17 year during the monitoring period, and 8.74 to 39.4 mrem per year during closure operations 18 (BSC, 2008c, Tables 16-18). Waste packages are emplaced in the subsurface using a 19 transport and emplacement vehicle that is remotely controlled and monitored; thus, the 20 subsurface workers receive lower external doses than the surface workers. Table G-11 21 provides the maximum and collective doses for specific worker activities.

22 G-27 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 Table G-11 External Dose for Involved Workers for Specific Activities Maximum Individual Collective Dose Activity a,b Dose (person-rem)

(rem per year)

Receipt Facility 1.3 840 Initial Handling Facility 0.8 110 Wet Handling Facility 0.4 300 Canister Receipt and Closure Facilities 0.29 630 Aging Facility 0.30 200 Low-Level Waste Facility 0.7 310 Cask Receipt Security Station 0.4 230 Total Surface Worker 1.3 2,620 External Dose Subsurface (Operations) 0.21 510 Subsurface (Monitoring) 0.2 510 Subsurface (Closure) 0.039 80 Total Subsurface Worker 0.21 1,100 External Dose 2 a Annual doses based on processing 500 casks per year or about 3,000 MTHM of commercial SNF throughput per 3 year.

4 b Collective doses based on processing a total waste throughput of 70,000 MTHM.

5 Source: DOE, 2008, Tables D-9 and D-10 6

7 Table G-12 provides the estimated individual doses for site workers and an offsite maximally 8 exposed individual over the operational period of repository development (i.e., initial 9 construction until permanent closure of the underground facility) based on the values contained 10 in Tables G-10 and G-11. The highest exposure is estimated for the involved worker and 11 specifically for the workers at the receipt facility (RF) based on an external exposure from the 12 incoming waste containers. By comparison, the annual individual dose for all worker categories 13 from the atmospheric release of SNF is less than 1 mrem per year, except for the involved 14 workers at the WHF who are estimated to have an individual internal dose of 15 mrem per year.

15 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-28

1 Table G-12 Largest Annual Individual Dose for Workers and the Maximally Exposed 2 Offsite Individual from SNF during the Operational Period Largest Annual Individual Exposures (mrem/yr)

Exposure Exposure Construction Construction Group Source and Monitoring Closure Only Emplacement (Years 56-95) (Years96-105)

(Years 0-5)

(Years 6-55)

Internal Involved NAa <1 <1 <1 Dose Worker (Sub-surface) External NAa 210 200 390 Dose Internal Involved NAa 15 (WHF) <1 <1 Dose Worker (Surface) External NAa 1,300 (RF) <1 <1 Dose Noninvolved Internal Worker NAa <1 <1 <1 Dose (Sub-surface)

Noninvolved Internal Worker NAa <1 <1 <1 Dose (Surface)

Offsite Internal NAa <1 <1 <1 Individual Dose 3 a No applicable exposure because no SNF is at the site during this period.

4 5

G-29 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 Collective Dose from Spent Nuclear Fuel 2

3 The estimation of collective dose for worker groups should account for the variation in 4 exposures from atmospheric releases that occur as a result of the different locations of the 5 workers on the surface of the geological repository operations area, different periods of time for 6 these releases to occur (e.g., only subsurface releases occur during the monitoring and the 7 closure periods), and differing external exposures for workers because of their different work 8 activities. Internal dose from atmospheric releases is the largest for the WHF workers.

9 Therefore, the estimation of worker collective dose is simplified by estimating the WHF worker 10 dose as one component of the overall collective internal dose. The second component of the 11 overall collective internal dose represents all of the other workers and conservatively uses the 12 largest annual internal dose estimated for the other geological repository operations area 13 locations (e.g., the largest value for internal doses for all other workers in Table G-10). The 14 following equation shows how to calculate the collective worker dose for these two components 15 for internal exposure of workers from atmospheric releases of SNF:

16 17 CWD = AWD x WY 18 19 where CWD: collective worker dose 20 AWD: annual worker dose (from Table G-10) 21 WY: number of worker-years 22 The number of worker-years is based on Table G-8 for all workers other than WHF workers.

23 The number of WHF worker-years is estimated based on the assumption that 10 percent of the 24 SNF received at the repository is processed by the WHF, and the activity requires 36 workers to 25 process 300 MTHM in a year (BSC, 2008b, .0). Therefore, based on the statutory limit of 26 63,000 MTHM of commercial fuel, the number of WHF worker-years is 756.

27 28 The DOE estimated the collective doses from atmospheric releases from SNF operations at the 29 repository for 117,000 individuals residing within 84 kilometers (52 miles) of the facility to be 30 (1) 0.88 person-rem from WHF releases as a result of processing 6,300 MTHM, 31 (2) 5.5 person-rem from the aging pad releases based on an annual collective dose of 32 0.11 person-rem per year for each of the 50 years during the emplacement period, and 33 (3) 3.3 person-rem from subsurface facility releases based on an annual collective dose of 34 0.033 person-rem per year for each of the 100 years the subsurface facility will be ventilated 35 (DOE, 2008, Table D-5).

36 37 The DOE estimated the collective external doses for workers for the activities presented in 38 Table G-11. Table G-13 provides the collective dose for workers and the offsite public for both 39 atmospheric release and direct radiation associated with SNF during the operational period of a 40 repository.

41 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-30

1 Table G-13 Collective Internal Dose for Workers and the Public from SNF during 2 Repository Development Collective Dose for the Workers and the Public (person-rem)

Exposure Exposure Construction Construction Monitoring Closure Group Source and Only (Years 56- (Years Totals Emplacement (Years 0-5) 95)96-105)

(Years 6-55)

Surface Internal NA 1 NA NA 1 Construction Dose Internal Surface NA 20a <1 <1 20 Dose (Involved External Worker) NA 2,620 NA NA 2,620 Dose Surface Internal (Noninvolved NA 4 <1 <1 4 Dose Worker)

Subsurface Internal NA 1 NA NA 1 Construction Dose Internal Subsurface NA 1 <1 <1 1 Dose (Involved Worker) External NA 510 510 80 1,100 Dose Subsurface Internal (Noninvolved NA <1 <1 <1 <1 Dose Worker)

Internal NA 27 <1 <1 27 Total Worker Dose Doses External NA 3,183 510 80 3,773 Dose Offsite Population Internal (117,000 0 8 1 <1 9 Dose within 84 kilometers) 3 a The WHF workers contribute 60 percent of the total collective dose and represent less than 3 percent of the total 4 worker-years.

5 6 Exposure to Radioactive Waste from Operational Accidents 7

8 The DOE analyzed 10 postulated accident scenarios associated with SNF that could occur 9 during the operational period of the repository. The evaluations considered accidents caused 10 by seismic events as well as internal events such as the breach of a sealed TAD canister in the 11 WHF. The evaluations conservatively assumed the wind would blow in the direction of the 12 largest population sector (i.e., the south-southeast sector) with a population of 104,000 within 13 80 kilometers (50 miles) of the facility. The DOE described this assumption as the unfavorable 14 95th percentile meteorological conditions for the offsite public exposure assessment. Table G14 15 estimates the conditional consequences for these accident scenarios.

G-31 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 Table G-14 Conditional Doses for Accident Scenarios Involving SNF Releases during 2 Repository Handling and Emplacement Activities Individual Offsite Maximum Annual Maximum Collective Noninvolved Accident Scenario Frequency Offsite Dose Dose Worker Dose (mrem) (person-rem) (mrem)

Seismic event causing LLW facility collapse and failure of 2x10-4 35 310 3,500 HEPA filters and ductwork in other facilities Breach of uncanistered SNF 2x10-3 1 2.7x10-5 83 in sealed truck transport cask Breach of uncanistered SNF in unsealed truck cask in 1x10-5 0.94 26 52 WHF pool Breach of sealed DPC in air 2x10-6 9.1 250 55 Breach of SNF in unsealed 4x10-6 8.4 230 740 DPC in WHF pool Breach of SNF in sealed TAD 4x10-5 5.3 140 430 in WHF pool Breach of SNF in unsealed 1x10-5 4.9 130 290 TAD in WHF pool Drop of uncanistered SNF in 6x10-3 0.47 13 27 WHF pool Breach of uncanistered SNF

< 2x10-6 0.23 6.4 14 in WHF pool Breach of sealed truck 4x10-4 4.4 42 1,300 transport cask due to fire 3 Source: DOE, 2008, Table 4-25 4

5 Table G-14 accounts for the risk that the accident scenarios could occur during repository 6 development. The accident risk for the operational period is estimated by multiplying 50 years 7 (the period during which handling and emplacement of SNF occurs) by the annual frequency for 8 the specific scenario and the doses presented in Table G-14. Table G-15 presents the annual 9 frequency for each of the accident scenarios and the estimated risk over the 50 year period 10 when handling and waste emplacement activities are expected to occur.

11 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-32

1 Table G-15 Risk Associated with Accident Scenarios Involving SNF Releases during 2 Repository Handling and Emplacement Activities Offsite Individual Maximum Collective Annual Maximum Noninvolved Accident Scenario Risk Frequency Offsite Risk Worker Risk (person-(mrem/year) (mrem/year) rem/year)

Seismic event causing LLW facility collapse and failure of 2x10-4 0.35 3.1 35 HEPA filters and ductwork in other facilities Breach of uncanistered SNF 2x10-3 0.1 2.7x10-6 8.3 in sealed truck transport cask Breach of uncanistered SNF in unsealed truck cask in 1x10-5 4.7x10-4 0.01 0.03 WHF pool Breach of sealed DPC in air 2x10-6 9x10-4 0.02 0.006 Breach of SNF in unsealed 4x10-6 0.002 0.05 0.1 DPC in WHF pool Breach of SNF in sealed TAD 4x10-5 0.001 0.03 0.09 in WHF pool Breach of SNF in unsealed 1x10-5 0.002 0.06 0.1 TAD in WHF pool Drop of uncanistered SNF in 6x10-3 0.1 3.9 8.1 WHF pool Breach of uncanistered SNF

<2x10-6 <2x10-5 <6x10-4 <1x10-3 in WHF pool Breach of sealed truck 4x10-4 0.09 0.8 26 transport cask from fire 3 Source: DOE, 2008, Table 4-25 4

5 The risks for handling and emplacement activities presented in Table G-15 are low because of 6 the limited conditional doses for the accident scenarios (Table G-14) and the low annual 7 accident frequencies (e.g., all accident frequencies are 6x10-3 per year or less). Thus, the 8 maximum offsite individual risks are all less than 1 mrem per year, and the offsite collective risks 9 are under 1 person-rem per year except for two accident scenarios in which the collective risks 10 are 3.1 and 3.9 person-rem per year. The maximum noninvolved worker risk is larger than the 11 offsite individual dose estimate because the worker is present at the site and thus closer to any 12 accidental release. However, the largest risk for the noninvolved worker is 35 mrem per year, 13 and most of the accident scenarios result in a maximum noninvolved worker risk less than 14 1 mrem per year.

15 16 G.2.7.3 Repository Risks after Permanent Closure 17 18 When a repository is permanently closed, all openings into the underground facility are sealed 19 to eliminate any type of easy access to the waste. This begins the post-closure period of a 20 repository, and the NRC expects that the multiple barriers of the repository will continue to 21 perform their passive functions for maintaining safety (i.e., the repository site and design are 22 intended to preserve safety without maintenance). Although it is expected that a release in the G-33 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 distant future will occur, it is also expected that the repository will continue to function safely 2 (i.e., within the post-closure dose limits). Repository programs around the world have regulatory 3 dose limits for annual individual exposures of 15 mrem for long time periods 4 (e.g., 10,000 years). Collective doses are rarely estimated for post-closure safety because such 5 an estimate is of limited value given the large uncertainties in projecting populations far into the 6 future.

7 8 G.2.7.4 Summary of Risks of a Geological Repository 9

10 Table G-16 presents the collective dose for all the activities that would occur at each stage of 11 the pre-closure period. In particular, exposures during loading come from Table G-2, exposures 12 during transport come from Table G-3, exposures from natural occurring radionuclides during 13 repository development come from Table G-9, and exposures from the presence of SNF during 14 repository development come from Table G-13.

15 16 The transportation campaign results in the largest overall collective dose for workers and the 17 public from exposure to SNF. For workers, the collective dose is primarily from the potential 18 external dose during the loading of uncanistered fuel, whereas, public exposure is principally 19 from the external dose along the transportation routes. The potential exposures are well within 20 regulatory limits. Worker exposures to SNF during repository development are approximately 21 double the worker exposure to natural radioactivity (e.g., radon release). Worker exposure to 22 SNF during repository development is dominated by the external dose for those workers 23 performing activities in close proximity to the SNF (see Table G-13). Public exposure to SNF 24 during repository development is very limited (i.e., 0.09 person-rem/year); however, public 25 exposure to radon released to the atmosphere during repository development is significantly 26 larger (i.e., an annual collective dose of 127 person-rem, which is approximately 1,000 times 27 greater than the dose from the SNF).

28 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-34

1 Table G-16 Collective Dose for Workers and the Public from the Transportation 2 Campaign and Repository Development Exposure Workers Public Time Period and Activities Source (person-rem) (person-rem)

TRANSPORTATION CAMPAIGN Years 6-55 (Loading) SNF 8,300 3 Years 6-55 (Transportation) SNF 544 548 Transportation Campaign Totals NA 8,844 551 Transportation Annual Average NA 177 11 (based on 50 years)

REPOSITORY DEVELOPMENT Years 0-5 (Construction only) Natural 39 2 Natural 1,200 6,400 Years 6-55 (Construction and emplacement)

SNF 3,183 8 Natural 390 6,200 Years 56-95 (Monitoring)

SNF 510 1 Natural 380 780 Years96-105 (Closure)

SNF 80 <1 Natural 2,009 13,382 Repository Totals SNF 3,773 9 Repository Annual Average Natural 19 127 (based on 105 years) SNF 38 0.09 3

4 G.2.8 Decommissioning 5

6 The NRCs nuclear regulatory activities include decommissioning nuclear facilities, which means 7 reducing residual radioactivity to a level that permits either of the following actions:

8 9

• Release the property for unrestricted use, and terminate the license.

10

• Release the property under restricted conditions, and terminate the license.

11 12 The NRCs decommissioning regulations are found in 10 CFR Part 20, Subpart E, Radiological 13 Criteria for License Termination, which applies to facilities licensed under 10 CFR Part 30, 14 Rules of General Applicability to Domestic Licensing of Byproduct Material, 10 CFR Part 40, 15 10 CFR Part 50, 10 CFR Part 52, 10 CFR Part 60, 10 CFR Part 61, 10 CFR Part 63, 16 10 CFR Part 70, and 10 CFR Part 72, and also provides the main requirements for license 17 termination.

18 G-35 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 The NRC and Agreement States regulate the decommissioning of nuclear facilities, with the 2 ultimate goal of license termination. The following facilities would be expected to undergo 3 decommissioning:

4 5

• complex materials sites (byproduct sites) 6

• power reactors 7

• research and test reactors 8

• uranium recovery sites 9

• fuel cycle facilities 10 11 Approximately 100 materials licenses are terminated each year. Most of these license 12 terminations are routine, and the affected sites require little, if any, remediation to meet the 13 NRCs criteria for unrestricted use. A site released under restricted conditions has residual 14 radiological contamination present above NRC levels for unrestricted release.

15 16 As with LLW disposal sites, the primary hazard at decommissioning sites would likely be a 17 radiological release. This could occur from the failure of institutional or physical barriers.

18 19 The primary decommissioning guidance document used by licensees and the NRC is 20 NUREG-1757, Consolidated Decommissioning Guidance. NUREG-1757 is a three-volume 21 series that consolidates the current policies and guidance of the NRCs decommissioning 22 program. Volume 1 of NUREG-1757 provides guidance on the decommissioning process for 23 materials licensees, with applicability in some areas to reactor licensees. Volume 2 contains 24 guidance on characterization, surveys, and the determination of radiological criteria for all 25 licensees subject to the license termination regulations of 10 CFR Part 20, Subpart E.

26 Volume 3 addresses financial assurance, recordkeeping, and timeliness. The staff revises 27 NUREG-1757 periodically to reflect updates to the NRCs decommissioning policy, with the 28 latest revision issued in February 2012.

29 30 The NRC also provides decommissioning guidance related to non-power reactors in 31 NUREG-1537, Part 1, Guidelines for Preparing and Reviewing Applications for the Licensing of 32 Non-Power Reactors: Format and Content. In addition, Section 5 of NUREG-1620 provides 33 decommissioning guidance for uranium recovery (i.e., in situ leaching) facilities subject to 34 10 CFR Part 40, Appendix A.

35 36 Other studies and guidance that contain additional information on performing regulatory 37 analyses for specific decommissioning activities include the following documents.

38 39

• NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of 40 Nuclear Facilities, issued in 1988; and Supplement 1, Generic Environmental Impact 41 Statement on Decommissioning of Nuclear Facilities: Regarding the Decommissioning 42 of Nuclear Power Reactors, issued in 2002, address decommissioning generically.

43 Supplement 1 incorporated technological advances in decommissioning operations, 44 experience gained by licensees, and changes made to NRC regulations since the initial 45 publication of NUREG-0586 in 1988.

46 47

• NUREG-1738 addresses the principal safety concern for the decommissioning of the 48 current fleet of operating reactors, which is the storage of SNF in the SFP or at an ISFSI.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-36

1 G.3 NON-FUEL CYCLE ACTIVITIES 2

3 Non-fuel cycle activities involve the use of byproduct material. The primary NRC regulations for 4 these activities are those promulgated under 10 CFR Part 30 through 10 CFR Part 39 10.

5 Non-fuel cycle activities do not have backfitting regulations. Non-fuel cycle activities are diverse 6 in scope and use, with more than 18,500 specific materials licenses and approximately 100,000 7 general licenses in the following general areas:

8 9

• medical use (e.g., radiation therapy and nuclear medicine) 10

• irradiators 11

• radiography 12

• well logging 13

• manufacturing 14

• fixed and portable gauges 15

• measuring systems 16

• academics (e.g., for education and research) 17 18 NUREG-1350 provides background information on non-power reactor activities, including the 19 number of active licensees and general locations. The technical report series, NUREG-1556, 20 Consolidated Guidance about Materials Licenses, contains comprehensive reference 21 information about the various aspects of materials licensing and materials program 22 implementation for the non-fuel cycle activities. The Materials OpE Gateway is available on the 23 internal NRC Web site at http://drupal.nrc.gov/nmss/ope and consolidates various information 24 sources for ease in accessing and analyzing operating experience in the regulated materials 25 program.

26 27 The NRC regulates a wide variety of activities with diverse characteristics under 10 CFR Part 30 28 through 10 CFR Part 39. The materials may be solids, liquids, or gases and may be sealed or 29 unsealed sources. Quantities in use may range from microcuries to megacuries, and access to 30 the byproduct materials may be unlimited (e.g., for consumer products) to tightly controlled 31 (e.g., for large irradiators). All these factors affect risk and associated impact and 32 implementation costs.

33 34 This section summaries the data reference materials related to byproduct materials that may be 35 useful in preparing a regulatory analysis.

36 37 NUREG-1556 is an extensive, 21-volume document that provides program-specific guidance to 38 assist in preparing and reviewing applications for licenses for the use of byproduct material.

39 The program-specific guidance is intended for use by applicants, licensees, NRC staff, and 40 Agreement States.

41 10 These regulations are 10 CFR Part 31, General Domestic Licenses for Byproduct Material, 10 CFR Part 32, Specific Domestic Licenses to Manufacture or Transfer Certain Items Containing Byproduct Material, 10 CFR Part 33, Specific Domestic Licenses of Broad Scope for Byproduct Material, 10 CFR Part 34 ,

Licenses for Industrial Radiography and Radiation Safety Requirements for Industrial Radiographic Operations, 10 CFR Part 35, Medical Use of Byproduct Material, 10 CFR Part 36, Licenses and Radiation Safety Requirements for Irradiators, 10 CFR Part 37, Physical Protection of Category 1 and Category 2 Quantities of Radioactive Material, and 10 CFR Part 39, Licenses and Radiation Safety Requirements for Well Logging.

There is no 10 CFR Part 38.

G-37 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 NUREG/CR-6642, Risk Analysis and Evaluation of Regulatory Options for Nuclear Byproduct 2 Material Systems, includes activities regulated under 10 CFR Part 30 through 3 10 CFR Part 39 11. The three-volume report uses insights obtained from risk analyses to identify 4 regulatory options for the oversight of the various nuclear byproduct material licenses. The 5 methodology includes the following:

6 7

• organizing all nuclear byproduct material licenses into 40 systems (i.e., groups of 8 activities) 9 10

• describing each system in terms of tasks, hazards, barriers (i.e., physical and 11 administrative barriers that limit doses to workers and the public), and receptors 12 13

• performing a radiation risk assessment for each system, including normal and 14 accident doses to workers and a maximally exposed member of the public 15 16

• considering the social and economic benefits, associated costs, and the risks, for 17 each system 18 19 NUREG/CR-6642 includes the initial information that is useful to perform a regulatory analysis 20 for byproduct material. The primary risk associated with the studied activities is overexposure to 21 workers and the public from the failure of one or more protective barrier(s) or lost or misplaced 22 sources. The analyst should consider the assumptions used in the analysis to assess whether 23 they impact the regulatory analysis. In particular, the analysis for medical systems does not 24 consider patient doses because the individual receives a benefit from the use of the system; 25 therefore, the NRC considers such doses a special category separate from those to workers 26 and the public. In addition, NUREG/CR-6642 only addresses the maximally exposed member 27 of the public and does not calculate population doses.

28 29 The analyst can find occupational exposure data in NUREG-0713, which summarizes the 30 annual occupational exposure data that are maintained in the NRC Radiation Exposure 31 Information and Reporting System (REIRS). The current version is Volume 39, the 50th annual 32 report, which was issued in March 2019. The annual reports compile information from five of 33 the seven NRC licensee categories subject to 10 CFR 20.2206, Reports of Individual 34 Monitoring:

35 36 (1) commercial nuclear power reactors and test reactor facilities 37 (2) industrial radiography 38 (3) fuel processing, including uranium enrichment facilities; fabricating; or reprocessing of 39 special nuclear material above specified amounts 40 (4) processing or manufacturing for distributing byproduct material above specified amounts 41 (5) ISFSIs 42 11 In 10 CFR Part 37, the NRC addresses the level of security for nuclear byproduct material system activities, which is not covered in the report. There is no 10 CFR Part 38.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-38

1 Two licensee categories, facilities for land disposal of LLW and geologic repositories for HLW, 2 do not report as there are no NRC-licensed LLW disposal facilities and there are no geologic 3 repositories.

4 5 NMED is another source of information on non-fuel cycle activity events, accessible through the 6 NRC internal Web site Material OpE Gateway. The NMED Annual Report presents information 7 on trending and analysis of events reported to the NRC that involve radioactive materials.

8 NMED contains information from materials, fuel cycle, and non-power reactor licensees on 9 events such as personnel radiation overexposures, medical misadministration, losses of 10 radioactive material, and potential criticality events. The current version is the Nuclear Material 11 Event Database Annual Report for Fiscal Year 2017.

12 13 NUREG-1717, Systematic Radiological Assessment of Exemptions for Source and Byproduct 14 Materials, systematically assesses potential individual and collective (population) radiation 15 doses associated with the current exemptions from licensing for the majority of 10 CFR Part 30, 16 10 CFR Part 40, and 10 CFR Part 70 licenses. The report estimates doses for the normal life 17 cycle of a particular product or material, covering distribution and transport, intended or 18 expected routine use, and disposal. In addition, it estimates assessments potential doses from 19 accidents and misuse. Finally, it assesses potential radiological impacts associated with 20 selected products containing byproduct material that currently may be used under a general or 21 specific license and may be candidates for exemption from licensing requirements.

22 23 The regulatory analysis for the 2018 10 CFR Part 35 final rule, Medical Use of Byproduct 24 MaterialMedical Event Definitions, Training, and Experience, and Clarifying Amendments is a 25 recent example of an analysis for this group of non-power reactor and non-fuel cycle activities.

26 G-39 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 G.4 COMMON ACTIVITIES 2

3 Activities common to both groups of non-power reactor activities include the following:

4 5

• transportation (10 CFR Part 71) 6 7

• security (10 CFR Part 73) 8 9

• material control and accountability (10 CFR Part 74) 10 11

• emergency planning and preparedness (10 CFR 40.31, Application for Specific 12 Licenses, 10 CFR 70.22, Contents of Applications, and 10 CFR 76.91, Emergency 13 Planning) 14 15 The following sections provide additional data references that should be used in concert with the 16 specific non-power reactor activities being evaluated.

17 18 G.4.1 Transportation 19 20 About 3 million packages of radioactive materials are shipped each year in the United States, 21 either by highway, rail, air, or water. Regulating the safety of these shipments is the joint 22 responsibility of the NRC and the U.S. Department of Transportation, as established by a 23 memorandum of understanding (44 FR 38690) between the two agencies.

24 25 To apply for a Certificate of Compliance for a package design, a vendor submits an application 26 to the NRC for review and approval in accordance with 10 CFR Part 71. The application would 27 address the safety and operational characteristics of the package, including design analysis for 28 structural, thermal, radiation shielding, nuclear criticality, and material containment. In addition, 29 the application would contain operational guidance, such as any testing and maintenance 30 requirements, operating procedures, and conditions for package use. For the NRC to certify a 31 transportation package design, actual tests or computer analyses must demonstrate that, after 32 the tests for normal conditions of transport and hypothetical accident conditions, the package 33 will meet the appropriate containment, dose rates, and criticality safety criteria in 10 CFR Part 34 71. The tests for hypothetical accident conditions are performed in sequence to determine their 35 cumulative effects on the package. If the package design meets NRC requirements, the NRC 36 issues a Radioactive Material Package Certificate of Compliance to the vendor.

37 38 NRC licensees are authorized to ship radioactive material in an approved package under the 39 general license provisions of 10 CFR Part 71; Agreement State licensees ship radioactive 40 materials under Department of Transportation regulations. Before any shipment can occur, the 41 shipper must review the package Certificate of Compliance to determine whether any testing or 42 maintenance is required. The shipper may be required to check or change package seals and 43 other components or perform leak testing. In addition, the shipper must take radiation 44 measurements at specific locations on and around the package to make sure that the levels are 45 below the required limits. The shipper must also meet the U.S. Department of Transportation 46 requirements for shipment of the nuclear material, including with regard to route selection, 47 vehicle condition and placarding, driver training, package marking, labeling, and other shipping 48 documentation.

49 50 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-40

1 Certain specific requirements apply to shippers of SNF, including the following:

2 3

• A licensee must use NRC-approved highway routes for the transport of SNF.

4 5

• The licensee must make sure that SNF is protected against radiological sabotage.

6 Shippers that transport or deliver SNF to a carrier for transport must meet specific 7 requirements that include the following:

8 9 o notifying the NRC of the shipment 10 11 o having procedures for addressing emergencies 12 13 o having a communications center 14 15 o having a written log of shipment events 16 17 o making arrangements with local law enforcement agencies for shipments while 18 en route 19 20 o using armed escorts in heavily populated areas 21 22 Additional regulations governing nuclear materials transportation include the following:

23 24

• 10 CFR Part 37, Subpart D, Physical Protection in Transit 25 26

• 10 CFR Part 73 27 28 The primary radiological hazards associated with transportation are the loss of containment of the 29 hazardous material being transported, failure of the shielding to perform its function, or, for certain 30 materials, inadvertent criticality.

31 32 The following data reference materials may be useful in preparing regulatory analyses for 33 regulatory actions affecting non-power reactor activities:

34 35

• NUREG-1609, Standard Review Plan for Transportation Packages for Radioactive 36 Material 37 38

• NUREG-1617, Standard Review Plan for Transportation Packages for Spent Nuclear 39 Fuel 40 41

• NUREG/CR-6407, Classification of Transportation Packaging and Dry Spent Fuel 42 Storage System Components According to Importance to Safety 43 44

• NUREG-2125, Spent Fuel Transportation Risk Assessment 45 46

• NUREG/BR-0292, Revision 2, Safety of Spent Fuel Transportation 47 48

• NUREG-0561, Revision 2, Physical Protection of Shipments of Irradiated Reactor Fuel 49 G-41 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1

• NUREG-2155, Revision 1, Implementation Guidance for 10 CFR Part 37, Physical 2 Protection of Category 1 and Category 2 Quantities of Radioactive Material (Subpart D) 3 4

• NUREG-0170, Final Environmental Statement on the Transportation of Radioactive 5 Material by Air and Other Modes 6

7

• NUREG/CR-4829, Volumes 1 and 2, Shipping Container Response to Severe Highway 8 and Railroad Accident Conditions 9

10

• NUREG/CR-6672, Volume 1, Reexamination of Spent Fuel Shipment Risk Estimates 11 12

• Regulatory Guide 7.9, Revision 2, Standard Format and Content of Part 71 Applications 13 for Approval of Packages for Radioactive Materials 14 15 G.4.2 Security 16 17 The NRC and Agreement States regulate the use of radioactive material in order to protect 18 people and the environment. Material licensees have the primary responsibility to maintain the 19 security and accountability of the radioactive material in their possession. The events of 20 September 11, 2001, put new emphasis on security to prevent the malicious use of radioactive 21 material. The NRC works with its Federal and State partners and the international community, 22 to provide appropriate safety and security requirements for radioactive materials without 23 discouraging their beneficial use.

24 25 In 10 CFR Part 73, the NRC prescribes requirements for the establishment and maintenance of 26 plants that use special nuclear material and of a physical protection system that has capabilities 27 for the protection of special nuclear material at fixed sites and in transit. Design-basis threats 28 referenced in this regulation are used to design safeguards systems to protect against acts of 29 radiological sabotage and to prevent the theft or diversion of special nuclear material. The 30 provisions of 10 CFR Part 73 apply to the following:

31 32

• power reactor licensees 33

• research and test reactors 34

• decommissioning reactors 35

• fuel cycle licensees 36 37 Radioactive byproduct material provides critical capabilities in the oil and gas, electrical power, 38 construction, and food industries. It is used to treat millions of patients each year in diagnostic 39 and therapeutic medical procedures and is used in technology research and development. 10 40 CFR Part 37 contains security requirements for the following types of facilities:

41 42

• industrial licensees 43

• academic and research licensees 44

• medical licensees 45 46 Material Security 47 48 On March 19, 2013, the NRC published the a final rule, Physical Protection of Byproduct 49 Material. This regulation established security requirements in 10 CFR Part 37 for the use and 50 transport of Category 1 and Category 2 quantities of radioactive materials, as well as for NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-42

1 shipments of small amounts of irradiated reactor fuel. Category 1 and Category 2 quantities of 2 radioactive materials are thresholds established by the International Atomic Energy Agency in 3 its Code of Conduct on the Safety and Security of Radioactive Sources. The objective of this 4 rule is to provide reasonable assurance of preventing the theft or diversion of Category 1 and 5 Category 2 quantities of radioactive materials. The final rule incorporates lessons learned by 6 the NRC and the Agreement States in implementing the post-September 11, 2001, security 7 measures, as well as stakeholder input on the proposed rule. The supporting regulatory 8 analysis determined that the rule was cost justified because the regulatory initiatives potentially 9 will reduce unnecessary radiation exposure to patients. Additionally, the rule updated, clarified, 10 and strengthened the existing regulatory requirements and thereby promotes public health and 11 safety. The analysis estimated that cost savings would be realized by the removal of attestation 12 requirements for certain board-certified individuals, by the modification of medical event 13 reporting criteria to ensure only significant events need to be reported, and by other 14 modifications to the regulations.

15 16 The Part 37 rule was effective on May 20, 2013, and NRC licensees were to comply by 17 March 19, 2014. The Agreement States issued compatible 10 CFR Part 37 requirements by 18 March 19, 2016. Upon the effective date of the NRC and Agreement State requirements, the 19 NRC rescinded the NRC security orders issued under the NRCs common defense and security 20 authority (with one exception described below). Following the effective dates of their respective 21 requirements, the NRC and the Agreement States also rescinded NRC- or Agreement 22 State-issued orders, removed license conditions, or took any other necessary action with 23 respect to the increased controls and fingerprinting security requirements issued to licensees 24 under the NRCs and the Agreement States public health and safety authority.

25 26 However, the NRC did not rescind EA-09-293, Issuance of Orders Imposing Trustworthiness 27 and Reliability Requirements for Unescorted Access to Certain Radioactive Material upon the 28 effective date of 10 CFR Part 37. The NRC issued this order to licensees at their request to 29 voluntarily commit to requirements to assess the trustworthiness and reliability of employees to 30 enable them to provide services to licensees with unescorted access. The NRC subsequently 31 rescinded this order on March 27, 2017.

32 33 Cybersecurity 34 35 High-profile cyber attacks, such as the December 2015 attack on Ukraines power grid, 36 underscore the importance of continuing to evaluate the need for a cyber security regulatory 37 framework for all classes of NRC licensees. The NRC has gained in-depth experience with 38 cybersecurity as a result of the development, implementation, and inspections performed under 39 10 CFR 73.54, Protection of Digital Computer and Communication Systems and Networks.

40 The NRCs oversight of cyber security implementation at operating reactors has positioned the 41 agency to develop, as needed, cybersecurity regulations or other measures for various types of 42 NRC licensees.

43 44 Fuel Cycle Facilities 45 46 Fuel cycle facility licensees comprise a broad spectrum of facility types and processes. The 47 special nuclear material and hazardous chemicals at fuel cycle facilities present safety and 48 security concerns that could lead to potential consequences of concern, such as diversion, theft, 49 sabotage, and radiological or chemical release as a result of a cyber attack. Currently, fuel 50 cycle facility licensees are under interim compensatory measures orders to address certain 51 security threats, including a cyber attack. The two Category I fuel cycle facility licensees under G-43 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 NRC regulatory jurisdiction are also required to protect against the design-basis threat as 2 described in 10 CFR 73.1, Purpose and Scope, which includes a cyber attack as an element 3 of the design-basis threat.

4 5 On March 24, 2015, the Commission issued staff requirements memorandum 6 SRM-SECY-14-0147, Staff RequirementsSECY-14-0147Cyber Security for Fuel Cycle 7 Facilities, which directed the staff to initiate an expedited cybersecurity rulemaking for fuel cycle 8 facility licensees. The staff engaged with external stakeholders and developed the regulatory 9 basis document (NRC, 2016). On October 4, 2017, the staff provided the Commission the 10 proposed rule to establish cyber security requirements for certain nuclear fuel cycle facility 11 applicants and licensees (NRC, 2017c).

12 13 Non-Power Reactors 14 15 Non-power reactor designs vary significantly both in terms of maximum licensed power levels 16 and in the quantity, enrichment, and form of nuclear materials maintained at the facility. In 17 2012, the NRC formed a working group that included representation from the National 18 Organization of Test Research and Training Reactors. The working group had the goals to 19 (1) gather information on the cybersecurity protection currently in place at non-power reactor 20 facilities through licensee self-assessments, (2) conduct surveys to validate information in the 21 licensee self-assessments, and (3) analyze the self-assessments and survey information within 22 the framework of the risk posed to public health and safety.

23 24 Following receipt of the licensee self-assessments in 2013 and 2014, the working group 25 conducted site visits at four representative non-power reactor facilities to determine what 26 measures are in place to protect critical digital assets from cyber attacks, and whether the NRC 27 needs to take any action to require licensees to strengthen their programs. Based on the 28 observations and assessments from the site visit, the working group concluded that non-power 29 reactor licensees have implemented an adequate level of cyber security at their facilities. The 30 working group developed and in January 2016 published a guidance document, Cyber 31 Security: Effective Practices for the Establishment and Maintenance of Adequate Cyber 32 Security at Non-Power (Research and Test) Reactor Facilities, which provides information for 33 non-power reactor licensees on how to use instrument and control technologies and modern 34 computer/networking technologies in a manner that provides adequate cybersecurity protection 35 and mitigates the risks from cyber-based threats.

36 37 Independent Spent Fuel Storage Installations 38 39 Spent fuel that has already been cooled in the SPF is typically placed in a storage cask 40 surrounded by inert gas and stored at an ISFSI. Licensees subject to 10 CFR 72.212, 41 Conditions of General License Issued under §72.210 (i.e., licenses limited to storage of spent 42 fuel in casks), must also comply with specific portions of 10 CFR 73.55, Requirements for 43 Physical Protection of Licensed Activities in Nuclear Power Reactors against Radiological 44 Sabotage, and the additional security measures orders. However, they are not subject to 45 10 CFR 73.54, which specifically applies only to operating reactors and combined operating 46 license holders.

47 48 In 2012, the staff formed a working group and studied cybersecurity protections at three ISFSIs 49 to determine whether the potential cyberthreats to ISFSI systems warrant additional cyber 50 protections. At the time, the staff determined that the licensees cybersecurity efforts NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-44

1 adequately protect the ISFSIs from a cyber attack. The NRC performs further evaluations of 2 ISFSI system cybersecurity as part of the re-evaluation of ISFSI physical security protections.

3 4 Byproduct Materials Licensees 5

6 The staff performed an evaluation and concluded that byproduct materials licensees that 7 possess Category 1 or Category 2 quantities of radioactive material do not rely solely on digital 8 assets to ensure safety or physical protection. Rather, these licensees generally use a 9 combination of measures, such as doors, locks, barriers, human resources, and operational 10 processes, to ensure security, which reflects a defense-in-depth approach to physical protection 11 and safety. As a result, the staff concluded that a compromise of any of the digital assets would 12 not result in a direct dispersal of Category 1 or Category 2 quantities of radioactive material, or 13 the exposure of individuals to radiation, without a concurrent and targeted breach of the physical 14 protection measures in force for these licensees.

15 16 Therefore, the staff determined that the current cybersecurity threat and potential consequences 17 do not warrant regulatory action. The NRC published a Federal Register notice to discontinue 18 the rulemaking activity that would have developed cyber security requirements for byproduct 19 materials licensees possessing Category 1 or Category 2 quantities of radioactive materials 20 (NRC, 2018a).

21 22 G.4.3 Material Control and Accountability 23 24 The NRC provides the principal requirements for special nuclear material licensing in 25 10 CFR Part 70 and 10 CFR Part 74. In 10 CFR 70.22(b), the NRC specifies that a license 26 application must contain a full description of the applicants program for the control and 27 accounting of such special nuclear material to show how compliance with the graded material 28 control and accounting requirements of 10 CFR Part 74, Subparts B-E, will be accomplished.

29 In 1987, the NRC revised the material control and accountability requirements for NRC 30 licensees authorized to possess and use a formula quantity (i.e., 5 formula kilograms or more) 31 of strategic special nuclear material (NRC, 1987a). Those revisions, issued as 10 CFR Part 74, 32 Subpart E, Formula Quantities of Strategic Special Nuclear Material, require timely monitoring 33 of in-process inventory and discrete items to detect anomalies potentially indicative of material 34 losses. Timely detection and enhanced loss localization capabilities are beneficial to resolve 35 alarms and recover material in the event of an actual loss.

36 37 The following are useful data references for performing the regulatory analyses in this area:

38 39

• NUREG-1280, Revision 2, Acceptable Standard Format and Content for Material 40 Control and Accounting Plan Required for Strategic Special Nuclear Material, was first 41 published in 1987 to present criteria that could be used by applicants, licensees, and 42 NRC license reviewers in the initial preparation and subsequent review of fundamental 43 nuclear material control plans submitted in response to the Reform Amendment. The 44 report addressed general performance objectives, system capabilities, process 45 monitoring, item monitoring, alarm resolution, quality assurance, and accounting.

46 47

• NUREG-2159, Acceptable Standard Format and Content for the Material Control and 48 Accounting Plan Required for Special Nuclear Material of Moderate Strategic 49 Significance, describes the standard format and content suggested by the NRC for use 50 in preparing material control and accountability plans for facilities authorized to hold 51 special nuclear material of moderate strategic significance.

G-45 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 2

• NUREG-1065, Revision 3, Acceptable Standard Format and Content for the Material 3 Control and Accounting Plan Required for Special Nuclear Material of Low Strategic 4 Significance, contains information that the licensee or applicant should provide in its 5 fundamental nuclear material control plan to implement the requirements of 6 10 CFR 74.31, Nuclear Material Control and Accounting for Special Nuclear Material of 7 Low Strategic Significance.

8 9

• NUREG/CR-5734, Recommendations to the NRC on Acceptable Standard Format and 10 Content for the Fundamental Nuclear Material Control (FNMC) Plan Required for 11 Low-Enriched Uranium Enrichment Facilities, recommends information that the licensee 12 or applicant should provide in the FNMC Plan to implement the requirements of 13 10 CFR 74.33, Nuclear Material Control and Accounting for Uranium Enrichment 14 Facilities Authorized to Produce Special Nuclear Material of Low Strategic Significance.

15 This document also describes methods that should be acceptable for compliance with 16 the general performance objectives.

17 18

• Regulatory Guide 5.29, Revision 2, Special Nuclear Material Control and Accounting 19 Systems for Nuclear Power Plants, describes acceptable methods and procedures for 20 the implementation and maintenance of a special nuclear material control and 21 accounting system for non-fuel cycle facilities, including nuclear power reactors, 22 research and test reactors, and ISFSIs.

23 24

• In 10 CFR Part 75, the NRC implements the requirements established by the safeguards 25 agreements between the United States and the International Atomic Energy Agency.

26 This regulation contains requirements to ensure that the United States meets its nuclear 27 non-proliferation obligations under the safeguards agreements. These obligations 28 include providing information to the International Atomic Energy Agency on the physical 29 location of applicant, licensee, or certificate holder activities; information on sources and 30 special nuclear materials; and access to the physical location of applicant, licensee, or 31 certificate holder activities.

32 33 G.4.4 Emergency Planning and Preparedness 34 35 The objective of the emergency planning program is to ensure that fuel facility licensees, 36 non-power reactor licensees, and some materials licensees are capable of implementing 37 adequate measures to protect public health and safety in the event of a radiological emergency.

38 As a condition of their licenses, these licensees must develop and maintain emergency plans 39 that meet comprehensive NRC emergency planning requirements. These licensees are 40 responsible for preventing accidents. Should an accident occur, local public safety authorities, 41 such as fire and police departments, will act to protect the public.

42 43 After a large, toxic release of UF6 at the Sequoyah Fuels Corporation conversion facility in 1986, 44 the NRC decided emergency plans for fuel facilities should also account for hazardous chemical 45 releases. At uranium conversion, enrichment and fuel fabrication facilities, the most significant 46 accidents would be a UF6 release, fire, or criticality (e.g., an unintended, self-sustaining nuclear 47 chain reaction). There is likely to be little or no warning time before these accidents start.

48 However, most can be controlled within roughly half an hour.

49 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-46

1 Regulatory Guide 3.67, Revision 1, Standard Format and Content for Emergency Plans for 2 Fuel Cycle and Materials Facilities, contains detailed guidance on emergency planning. In 3 general, the scope and depth of fuel cycle facility plans are more variable than and not as 4 extensive as those of power reactors, reflecting the diverse nature of these facilities and the 5 hazards and risks associated with their operation. For example, fuel cycle facility emergency 6 plans have the following:

7 8

• no designated emergency planning zones 9

• no extraordinary provisions to alert and notify the general public 10

• only two levels of emergency classifications 11 - Alertrequiring no offsite response 12 - Site Area Emergencycould require offsite response 13 14 The Federal Emergency Management Agency has no oversight over State and local 15 governments with regard to fuel cycle facilities. This reduced scope and depth are justified 16 because the EPA protective action guidelines will not be exceeded beyond the site boundary.

17 18 Regulatory Guide 3.67 may be useful in preparing regulatory analyses for regulatory actions 19 affecting emergency preparedness for certain fuel cycle, non-power reactor, and other 20 radioactive material licensees.

G-47 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 G.5 REFERENCES 2

3 42 U.S.C. 2011 et seq., Atomic Energy Act of 1954, as amended. Available at 4 https://www.nrc.gov/docs/ML1327/ML13274A489.pdf#page=23.

5 6 42 U.S.C. 2021b, Low-Level Radioactive Waste Policy Amendments Act of 1985. Available at 7 https://www.govinfo.gov/content/pkg/STATUTE-99/pdf/STATUTE-99-Pg1842.pdf.

8 9 42 U.S.C. 7901 et seq., Uranium Mill Tailings Radiation Control Act of 1978.

10 11 Bechtel SAIC Company (BSC), GROA Airborne Release Dose Calculation 12 000-PSA-MGR0-01200-000-00C, March 26, 2008a. Agencywide Documents Access and 13 Management System (ADAMS) Accession No. ML090770783.

14 15 BSC, Repository ALARA Goal Compliance, 000-30R-MGR0-04000-000, February 5, 2008b.

16 ADAMS Accession No. ML092790253.

17 18 BSC, Subsurface Worker Dose Assessment, 800-00C-SS00-00600-000-00B, 19 March 26, 2008c. ADAMS Accession No. ML092790265.

20 21 BSC, GROA Worker Dose Calculation, 000-PSA-MGR0-01400-000-00C, March 27, 2008d.

22 ADAMS Accession No. ML090770784.

23 24 Code of Federal Regulations (CFR), Title 10, Energy, Part 20, Standards for Protection against 25 Radiation. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part020/.

26 27 CFR, Title 10, Energy, Part 30, Rules of General Applicability to Domestic Licensing of 28 Byproduct Material. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part030/.

29 30 CFR, Title 10, Energy, Part 31, General Domestic Licenses for Byproduct Material. Available 31 at https://www.nrc.gov/reading-rm/doc-collections/cfr/part031/.

32 33 CFR, Title 10, Energy, Part 32, Specific Domestic Licenses to Manufacture or Transfer Certain 34 Items Containing Byproduct Material. Available at https://www.nrc.gov/reading-rm/doc-35 collections/cfr/part032/.

36 37 CFR, Title 10, Energy, Part 33, Specific Domestic Licenses of Broad Scope for Byproduct 38 Material. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part033/.

39 40 CFR, Title 10, Energy, Part 34, Licenses for Industrial Radiography and Radiation Safety 41 Requirements for Industrial Radiographic Operations. Available at 42 https://www.nrc.gov/reading-rm/doc-collections/cfr/part034/.

43 44 CFR, Title 10, Energy, Part 35, Medical Use of Byproduct Material. Available at 45 https://www.nrc.gov/reading-rm/doc-collections/cfr/part035/.

46 47 CFR, Title 10, Energy, Part 36, Licenses and Radiation Safety Requirements for Irradiators.

48 Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part036/.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-48

1 CFR, Title 10, Energy, Part 37, Physical Protection of Category 1 and Category 2 Quantities of 2 Radioactive Material. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part037/.

3 4 CFR, Title 10, Energy, Part 39, Licenses and Radiation Safety Requirements for Well Logging.

5 Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part039/.

6 7 CFR, Title 10, Energy, Part 40, Domestic Licensing of Source Material. Available at 8 https://www.nrc.gov/reading-rm/doc-collections/cfr/part040/.

9 10 CFR, Title 10, Energy, Part 50, Domestic Licensing of Production and Utilization Facilities.

11 Available at http://www.nrc.gov/reading-rm/doc-collections/cfr/part050/.

12 13 CFR, Title 10, Energy, Part 51, Environmental Protection Regulations for Domestic Licensing 14 and Related Regulatory Functions. Available at https://www.nrc.gov/reading-rm/doc-15 collections/cfr/part051/.

16 17 CFR, Title 10, Energy, Part 52, Licenses, Certifications, and Approvals for Nuclear Power 18 Plants. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part052/.

19 20 CFR, Title 10, Energy, Part 60, Disposal of High-Level Radioactive Wastes in Geologic 21 Repositories. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part060/.

22 23 CFR, Title 10, Energy, Part 61, Licensing Requirements for Land Disposal of Radioactive 24 Waste. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part061/.

25 26 CFR, Title 10, Energy, Part 63, Disposal of High-Level Radioactive Wastes in a Geologic 27 Repository at Yucca Mountain, Nevada. Available at https://www.nrc.gov/reading-rm/doc-28 collections/cfr/part063/.

29 30 CFR, Title 10, Energy, Part 70, Domestic Licensing of Special Nuclear Material. Available at 31 https://www.nrc.gov/reading-rm/doc-collections/cfr/part070/.

32 33 CFR, Title 10, Energy, Part 71, Packaging and Transportation of Radioactive Material.

34 https://www.nrc.gov/reading-rm/doc-collections/cfr/part071/.

35 36 CFR, Title 10, Energy, Part 72, Licensing Requirements for the Independent Storage of Spent 37 Nuclear Fuel, High-Level Radioactive Waste, and Reactor-Related Greater Than Class C 38 Waste. Available at https://www.nrc.gov/reading-rm/doc-collections/cfr/part072/.

39 40 CFR, Title 10, Energy, Part 73, Physical Protection of Plants and Materials.

41 https://www.nrc.gov/reading-rm/doc-collections/cfr/part073/.

42 43 CFR, Title 10, Energy, Part 74, Material Control and Accounting of Special Nuclear Material.

44 https://www.nrc.gov/reading-rm/doc-collections/cfr/part074/.

45 46 CFR, Title 10, Energy, Part 75, Safeguards on Nuclear MaterialImplementation of 47 Safeguards Agreements Between the United States and the International Atomic Energy 48 Agency. https://www.nrc.gov/reading-rm/doc-collections/cfr/part075/.

G-49 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 CFR, Title 10, Energy, Part 76, Certification of Gaseous Diffusion Plants. Available at 2 https://www.nrc.gov/reading-rm/doc-collections/cfr/part076/.

3 4 Electrical Power Research Institute, Probabilistic Risk Assessment (PRA) of Bolted Storage 5 Casks: Quantification and Analysis Report, EPRI Report 1002877, December 2003.

6 7 International Atomic Energy Agency, Code of Conduct on the Safety and Security of 8 Radioactive Sources, January 2004. Available at https://www-9 pub.iaea.org/MTCD/publications/PDF/Code-2004_web.pdf.

10 11 U.S. Department of Energy (DOE), Final Supplemental Environmental Impact Statement for a 12 Geological Repository for the Disposal of Spent Nuclear Fuel and High-Level Radioactive 13 Waste at Yucca Mountain, Nye County, Nevada, DOE/EIS-0250F-S1, June 2008. Available at 14 https://www.energy.gov/nepa/downloads/eis-0250-s1-final-supplemental-environmental-impact-15 statement.

16 17 DOE, Final Environmental Impact Statement for the Disposal of Greater-Than-Class C (GTCC) 18 Low-Level Radioactive Waste and GTCC-Like Waste, DOE/EIS-0375, January 2016. Available 19 at https://www.energy.gov/nepa/downloads/eis-0375-final-environmental-impact-statement.

20 21 DOE, DOE Standard: Radiological Control, DOE-STD-1098-2017, January 2017. Available at 22 https://www.standards.doe.gov/standards-documents/1000/1098-AStd-2017/@@images/file.

23 24 DOE, Environmental Assessment for the Disposal of Greater-Than-Class C (GTCC) Low-Level 25 Radioactive Waste and GTCC-Like Waste at Waste Control Specialists, Andrews County, 26 Texas, DOE/EA-2082, October 2018. Available at 27 https://www.energy.gov/sites/prod/files/2018/11/f57/final-ea-2082-disposal-of-gtcc-llw-2018-28 10.pdf.

29 30 U.S. Department of Transportation and NRC, Transportation of Radioactive Material; 31 Memorandum of Understanding, Federal Register, July 1979 (44 FR 38690).

32 33 U.S. Environmental Protection Agency, Regulatory Impact Analysis of Final Environmental 34 Standards for Uranium Mill Tailings at Active Sites, EPA 520/1-83-010, September 1983.

35 36 U.S. Nuclear Regulatory Commission (NRC), Uranium Conversion, available at 37 https://www.nrc.gov/materials/fuel-cycle-fac/ur-conversion.html.

38 39 NRC, Backgrounder - Uranium Enrichment. ADAMS Accession No. ML031330159.

40 41 NRC, Final Environmental Statement on the Transportation of Radioactive Material by Air and 42 Other Modes, NUREG-0170, December 1977. ADAMS Accession No. ML022590355 43 (package).

44 45 NRC, Final Generic Environmental Impact Statement on Uranium Milling, NUREG-0706, 46 September 1980. ADAMS Accession No. ML032751661 (package).

47 NRC, Nuclear-Fuel-Cycle Risk Assessment: Descriptions of Representative Non-Reactor 48 Facilities, NUREG/CR-2873, Volume 1, September 1982.

49 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-50

1 NRC, Nuclear Fuel Cycle Risk Assessment: Survey and Computer Compilation of 2 Risk-Related Literature, NUREG/CR-2933, October 1982.

3 4 NRC, Nuclear Fuel Cycle Risk Assessment: Review and Evaluation of Existing Methods, 5 NUREG/CR-3682, May 1984.

6 7 NRC, Standard Format and Content for the Health and Safety Sections of License Renewal 8 Applications for Uranium Hexafluoride Production, Regulatory Guide 3.55, April 1985. ADAMS 9 Accession No. ML003739469.

10 11 NRC, Licensing of Alternative Methods of Disposal of Low-Level Radioactive Waste, 12 NUREG-1241, December 1986. ADAMS Accession No. ML053010322.

13 14 NRC, Material Control and Accounting Requirements for Facilities Licensed To Possess and 15 Use Formula Quantities of Strategic Special Nuclear Material, Federal Register, March 1987a 16 (52 FR 10033).

17 18 NRC, Environmental Standard Review Plan for the Review of a License Application for a 19 Low-Level Radioactive Waste Disposal Facility: Environmental Report, NUREG-1300, 20 April 1987b. ADAMS Accession No. ML053010347.

21 22 NRC, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, 23 NUREG-0586, 1988.

24 25 NRC, Shipping Container Response to Severe Highway and Railroad Accident Conditions, 26 NUREG/CR-4829, October 1988. Available at https://www.nrc.gov/reading-rm/doc-27 collections/nuregs/contract/cr4829/.

28 29 NRC, Standard Format and Content for a Topical Safety Analysis Report for a Spent Fuel Dry 30 Storage Cask, Regulatory Guide 3.61, February 1989. ADAMS Accession No. ML003739511.

31 32 NRC, Regulatory Analysis for the Resolution of Generic Issue 82, Beyond Design Basis 33 Accidents in Spent Fuel Pools, NUREG-1353, April 1989. ADAMS Accession 34 No. ML082330232.

35 36 NRC, Regulating the Disposal of Low-Level Radioactive Waste: A Guide to the Nuclear 37 Regulatory Commissions 10 CFR Part 61, NUREG/BR-0121, August 1989. ADAMS 38 Accession No. ML120720225.

39 40 NRC, Standard Format and Content for the Safety Analysis Report for an Independent Spent 41 Fuel Storage Installation or Monitored Retrievable Storage Installation (Dry Storage).

42 Regulatory Guide 3.48, Revision 1, August 1989. ADAMS Accession No. ML003739163.

43 44 NRC, Backfitting Guidelines, NUREG-1409, July 1990. ADAMS Accession 45 No. ML032230247.

46 47 NRC, Standard Format and Content of a License Application for a Low-Level Radioactive 48 Waste Disposal Facility, NUREG-1199, Revision 2, January 1991. ADAMS Accession 49 No. ML022550605.

50 G-51 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 NRC, Recommendations to the NRC on Acceptable Standard Format and Content for the 2 Fundamental Nuclear Material Control (FNMC) Plan Required for Low-Enriched Uranium 3 Enrichment Facilities, NUREG/CR-5734, November 1991. ADAMS Accession 4 No. ML15120A354.

5 6 NRC, Standard Review Plan for the Review of a License Application for a Low-Level 7 Radioactive Waste Disposal Facility, NUREG-1200, Revision 3, April 1994. ADAMS Accession 8 No. ML061370484.

9 10 NRC, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power 11 Reactors, Part 1, Format and Content, NUREG-1537, Part 1, February 1996a. Available at 12 https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1537/part1/.

13 14 NRC, Classification of Transportation Packaging and Dry Spent Fuel Storage System 15 Components According to Importance to Safety, NUREG/CR-6407, February 1996b. ADAMS 16 Accession No. ML15127A114.

17 18 NRC, Chemical Process Safety at Fuel Cycle Facilities, NUREG-1601, August 1997 19 (non-public). ADAMS Accession No. ML093070119.

20 21 NRC, Nuclear Fuel Cycle Facility Accident Analysis Handbook, NUREG/CR-6410, 22 March 1998. ADAMS Accession No. ML072000468.

23 24 NRC, Standard Review Plan for Transportation Packages for Radioactive Material, 25 NUREG-1609, March 1999. Available at https://www.nrc.gov/reading-rm/doc-26 collections/nuregs/staff/sr1609/final/.

27 28 NRC, Risk Analysis and Evaluation of Regulatory Options for Nuclear Byproduct Material 29 Systems, NUREG/CR-6642, February 2000. ADAMS Accession Nos. ML003693052 and 30 ML003693028.

31 32 NRC, Standard Review Plan for Spent Fuel Dry Storage Facilities, NUREG-1567, March 2000a.

33 ADAMS Accession No. ML003686776.

34 35 NRC, Reexamination of Spent Fuel Shipment Risk Estimates, NUREG/CR-6672, Volume 1, 36 March 2000b. ADAMS Accession No. ML003698324.

37 38 NRC, Standard Review Plan for Transportation Packages for Spent Nuclear Fuel, 39 NUREG-1617, March 2000c. ADAMS Accession No. ML003696262.

40 41 NRC, A Performance Assessment Methodology for Low-Level Radioactive Waste Disposal 42 Facilities: Recommendations of NRCs Performance Assessment Working Group, 43 NUREG-1573, October 2000. ADAMS Accession No. ML053250352.

44 45 NRC, Technical Study of Spent Fuel Pool Accident Risk at Decommissioning Nuclear Power 46 Plants, NUREG-1738, February 2001. ADAMS Accession No. ML010430066.

47 48 NRC, Integrated Safety Analysis Guidance Document, NUREG-1513, May 2001. ADAMS 49 Accession No. ML011440260.

50 NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-52

1 NRC, Systematic Radiological Assessment of Exemptions for Source and Byproduct Materials, 2 NUREG-1717, June 2001. ADAMS Accession No. ML011980433.

3 4 NRC, A Baseline Risk-Informed, Performance-Based Approach for In Situ Leach Uranium 5 Extraction Licensees, NUREG/CR-6733, September 2001a. ADAMS Accession 6 No. ML012840152.

7 8 NRC, Regulating Nuclear Fuel, NUREG/BR-0280, Revision 1, September 2001b. ADAMS 9 Accession No. ML15104A215.

10 11 NRC, Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities, 12 NUREG-0586, Supplement 1, Regarding the Decommissioning of Nuclear Power Reactors, 13 November 2002. Available at https://www.nrc.gov/reading-rm/doc-14 collections/nuregs/staff/sr0586/.

15 16 NRC, Standard Review Plan for In Situ Leach Uranium Extraction License Applications, 17 NUREG-1569, June 2003a. ADAMS Accession No. ML031550302 (package).

18 19 NRC, Standard Review Plan for the Review of a Reclamation Plan for Mill Tailings Sites under 20 Title II of the Uranium Mill Tailings Radiation Control Act of 1978, NUREG-1620, Revision 1, 21 June 2003b. ADAMS Accession No. ML032250190.

22 23 NRC, Consolidated Decommissioning Guidance, NUREG-1757. Available at 24 https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1757/.

25 26 NRC, Standard Format and Content of Part 71 Applications for Approval of Packages for 27 Radioactive Material, Regulatory Guide 7.9, Revision 2, March 2005. ADAMS Accession 28 No. ML050540321.

29 30 NRC, History and Framework of Commercial Low-Level Radioactive Waste Management in the 31 United States: ACNW White Paper, NUREG-1853, January 2007a. ADAMS Accession 32 No. ML070600684.

33 34 NRC, A Pilot Probabilistic Risk Assessment of a Dry Cask Storage System at a Nuclear Power 35 Plant, NUREG-1864, March 2007b. ADAMS Accession No. ML071340012.

36 37 NRC, Risk-Informed Decisionmaking for Nuclear Material and Waste Applications, Revision 1, 38 February 2008. ADAMS Accession No. ML080720238.

39 40 NRC, Generic Environmental Impact Statement for In-Situ Leach Uranium Milling Facilities, 41 NUREG-1910, May 2009. Available at https://www.nrc.gov/reading-rm/doc-42 collections/nuregs/staff/sr1910/.

43 44 NRC, Issuance of Orders Imposing Trustworthiness and Reliability Requirements for 45 Unescorted Access to Certain Radioactive Material, EA-09-293, December 16, 2009. ADAMS 46 Accession No. ML101670484.

47 48 NRC, Module 7.0: Health and Safety Fundamentals and Hazards within the Nuclear Fuel 49 Cycle, F201S - Fuel Cycle Processes, March 2010a. ADAMS Accession No. ML12045A011.

50 G-53 NUREG/BR-0058, Rev. 5, App. G, Rev. 0

1 NRC, Standard Review Plan for Spent Fuel Dry Storage Systems at a General License 2 Facility, NUREG-1536, Revision 1, July 2010b. ADAMS Accession No. ML101040620.

3 4 NRC, Standard Format and Content for Emergency Plans for Fuel Cycle and Materials 5 Facilities, Regulatory Guide 3.67, Revision 1, April 2011. ADAMS Accession 6 No. ML103360487.

7 8 NRC, Physical Protection of Byproduct Material, Federal Register, March 2013 (78 FR 16922).

9 Available at https://www.govinfo.gov/content/pkg/FR-2013-03-19/pdf/2013-05895.pdf.

10 11 NRC, Physical Protection of Shipments of Irradiated Reactor Fuel, NUREG-0561, Revision 2, 12 April 2013. ADAMS Accession No. ML13120A230.

13 14 NRC, Special Nuclear Material Control and Accounting Systems for Nuclear Power Plants, 15 Regulatory Guide 5.29, Revision 2, June 2013. ADAMS Accession No. ML13051A421.

16 17 NRC, Acceptable Standard Format and Content for Material Control and Accounting Plan 18 Required for Strategic Special Nuclear Material, NUREG-1280, Revision 2, September 2013a.

19 ADAMS Accession No. ML13253A308.

20 21 NRC, Acceptable Standard Format and Content for the Material Control and Accounting Plan 22 Required for Strategic Special Nuclear Material of Moderate Strategic Significance, 23 NUREG-2159, September 2013b. ADAMS Accession No. ML13253A310.

24 25 NRC, Acceptable Standard Format and Content for the Material Control and Accounting Plan 26 Required for Special Nuclear Material of Low Strategic Significance, NUREG-1065, Revision 3, 27 September 2013c. ADAMS Accession No. ML13253A305.

28 29 NRC, Management of Facility-Specific Backfitting and Information Collection, Management 30 Directive 8.4, October 9, 2013. ADAMS Accession No. ML12059A460.

31 32 NRC, Spent Fuel Transportation Risk Assessment, NUREG-2125, January 2014a. ADAMS 33 Accession No. ML14031A323.

34 35 NRC, Consequence Study of a Beyond-Design-Basis Earthquake Affecting the Spent Fuel Pool 36 for a U.S. Mark I Boiling Water Reactor, NUREG-2161, September 2014b. ADAMS Accession 37 No. ML14255A365.

38 39 NRC, Standard Format and Content for a Specific License Application for an Independent 40 Spent Fuel Storage Installation or Monitored Retrievable Storage Facility, Regulatory 41 Guide 3.50, Revision 2, September 2014. ADAMS Accession No. ML14043A080.

42 43 NRC, Standard Review Plan for Conventional Uranium Mill and Heap Leach Facilities - Draft 44 Report for Comment, NUREG-2126, November 2014. ADAMS Accession No. ML14325A634.

45 46 NRC, Implementation Guidance for 10 CFR Part 37, Physical Protection of Category 1 and 47 Category 2 Quantities of Radioactive Material, NUREG-2155, Revision 1, January 2015.

48 ADAMS Accession No. ML15016A172.

49 50 NRC, Staff RequirementsSECY-14-0147Cyber Security for Fuel Cycle Facilities, 51 SRM-SECY-14-0147, March 24, 2015. ADAMS Accession No. ML15083A175.

NUREG/BR-0058, Rev. 5, App. G, Rev. 0 G-54

1 2 NRC, Standard Review Plan for Fuel Cycle Facilities License Applications, NUREG-1520, 3 Revision 2, June 2015. ADAMS Accession No. ML15176A258.

4 5 NRC, Cyber Security: Effective Practices for the Establishment and Maintenance of Adequate 6 Cyber Security at Non-Power (Research and Test) Reactor Facilities, January 2016. ADAMS 7 Accession No. ML15253A060.

8 9 NRC, Rulemaking for Cyber Security at Fuel Cycle Facilities, Regulatory Analysis Document, 10 March 2016. ADAMS Accession No. ML15355A466.

11 12 NRC, Standard Review Plan for Renewal of Specific Licenses and Certificates of Compliance 13 for Dry Storage of Spent Nuclear Fuel, NUREG-1927, Revision 1, June 2016a. ADAMS 14 Accession No. ML16179A148.

15 16 NRC, Consolidated Guidance about Materials Licenses, NUREG-1556, June 2016b. Available 17 at https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1556/.

18 19 NRC, Review of Spent Fuel Reprocessing and Associated Accident Phenomena, 20 NUREG/CR-7232, February 2017a. ADAMS Accession No. ML17045A577.

21 22 NRC, Safety of Spent Fuel Transportation, NUREG/BR-0292, Revision 2, February 2017b.

23 ADAMS Accession No. ML17038A460.

24 25 NRC, Proposed RuleCyber Security at Fuel Cycle Facilities, SECY-17-0099, 26 October 4, 2017c. ADAMS Accession No. ML17018A219.

27 28 NRC, Cyber Security for Byproduct Materials Licensees, Federal Register, May 15, 2018a 29 (83 FR 22413).

30 31 NRC, Medical Use of Byproduct Material-Medical Event Definitions, Training and Experience, 32 and Clarifying Amendments, Federal Register, July 16, 2018b (83 FR 33046). Available at 33 https://www.govinfo.gov/content/pkg/FR-2018-07-16/pdf/2018-14852.pdf.

34 35 NRC, 2018-2019 NRC Information Digest, NUREG-1350, Volume 30, August 2018c.

36 Available at https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1350/.

37 38 NRC, Occupational Radiation Exposures at Commercial Nuclear Power Plant Reactors and 39 Other Facilities, NUREG-0713, Volume 39, Nuclear Material Event Database Annual Report for 40 Fiscal Year 2017, March 2019.

41 G-55 NUREG/BR-0058, Rev. 5, App. G, Rev. 0