ML24267A025

TLR-RES/DE/REB-2024-17, Assessment of Technical Information Needs and Considerations for Front-End Activities for Molten Salt Fuel Types

ML24267A025
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Issue date:	09/26/2024
From:	Lahaye N, Lowry P, Napier J, Wendy Reed, Richmond D, Short S, Thomas K, Darrell Todd NRC/RES/DE, Pacific Northwest National Laboratory
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Wendy Reed 301-415-7213
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DE-AC05-76RL01830, 31310019N0001, 31310022F0033 TLR-RES/DE/REB-2024-17
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f Technical Letter Report

[TLR-RES/DE/REB-2024-17]

Assessment of Technical Information Needs and Considerations for Front-End Activities for Molten Salt Fuel Types Date:

September 2024 Prepared in response to Front-End Task 2 in User Need Request NMSS-2022-002, by:

David Richmond Kenneth Thomas Nicole LaHaye Pete Lowry Jon Napier Steve Short Donald Todd Pacific Northwest National Laboratory Richland, Washington 99354 NRC Project Manager:

W. Reed Reactor Engineering Branch Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

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DISCLAIMER This report was prepared as an account of work sponsored by an agency of the U.S. Government.

Neither the U.S. Government nor any agency thereof, nor any employee, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product, or process disclosed in this publication, or represents that its use by such third party complies with applicable law.

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This report does not contain or imply legally binding requirements. Nor does this report establish or modify any regulatory guidance or positions of the U.S. Nuclear Regulatory Commission and is not binding on the Commission.

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f Remote target location PNNL-33899, Rev. 1 Assessment of Technical Information Needs and Considerations for Front-End Activities for Molten Salt Fuel Types September 2024 David Richmond Kenneth Thomas Nicole LaHaye Pete Lowry Jon Napier Steve Short Donald Todd Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Under Contract DE-AC05-76RL01830 Interagency Agreement: 31310019N0001 Task Order Number: 31310022F0033

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DISCLAIMER This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor Battelle Memorial Institute, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof, or Battelle Memorial Institute. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.

PACIFIC NORTHWEST NATIONAL LABORATORY operated by BATTELLE for the UNITED STATES DEPARTMENT OF ENERGY under Contract DE-AC05-76RL01830 Printed in the United States of America Available to DOE and DOE contractors from the Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831-0062 www.osti.gov ph: (865) 576-8401 fox: (865) 576-5728 email: reports@osti.gov Available to the public from the National Technical Information Service 5301 Shawnee Rd., Alexandria, VA 22312 ph: (800) 553-NTIS (6847) or (703) 605-6000 email: info@ntis.gov Online ordering: http://www.ntis.gov

PNNL-33899, Rev. 1 Assessment of Technical Information Needs and Considerations for Front-End Activities for Molten Salt Fuel Types September 2024 David Richmond Kenneth Thomas Nicole LaHaye Pete Lowry Jon Napier Steve Short Donald Todd Prepared for the U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Under Contract DE-AC05-76RL01830 Interagency Agreement: 31310019N0001 Pacific Northwest National Laboratory Richland, Washington 99354

PNNL-33899, Rev. 1 Executive Summary ii Executive Summary Molten salt reactors (MSRs) present distinct technical considerations associated with fuel processing operations, including front-end activities related to the enrichment, mixing, transportation, treatment, and storage of unirradiated fissile materials and carrier salts. The design and operations of new facilities and transportation packages to support new MSR fuel cycles require consideration of technical and safety implications of new fuel compositions, including their corrosivity and need for adequate removal of moisture, sulfur, and oxygen from commercial-grade salts and precursor chemicals.

This report details an assessment of the safety impacts related to front-end operations of MSRs and an assessment of existing regulations and safety guidance. Four different MSR technologies and information from the 1960s about the Atomic Energy Commission programs for the Aircraft Reactor and Molten Salt Reactor Experiment were evaluated. Primary areas of evaluation include fuel cycle facilities, transportation, and reactor fuel mixing operations up to reactor start-up. The available information was assessed to identify potential approaches that could be pursued for front-end operations and technical information needs relating to these operations, which are stated below:

x The possibility of collocating fuel cycle facilities with operating MSRs should be assessed.

While the NRC regulatory framework is flexible and broad enough to allow for collocated facilities, combining the fuel cycle facilities with the operating MSR would be novel.

The assessment in this report shows that current regulations and guidance appear appropriate for most MSR fuel types, and no necessary updates were identified. As noted and then discussed below, several technical considerations were identified:

x potential hazards associated with the stability of stored fuel constituents x

implications of hazards related to beryllium; and x

the need for related transportation packages and package characteristics for shipping molten salt fuels One novel characteristic of fuel processing for MSRs is the ability to collocate the fuel cycle facility at the reactor site. Normally, the fuel cycle facility is not collocated at the reactor site, and the fresh fuel is shipped from the fuel cycle facility to the reactor site, such that two separate licenses are required. However, an applicant for a collocated fuel cycle facility and MSR may consider combining their applications for a fuel cycle facility and an MSR operating license using NRC regulations in 10 CFR 50.31 or 10 CFR 52.8. This arrangement of having a fuel processing facility at the reactor site represents a different operational configuration from the configuration typically found for light-water reactors.

The assessment identified beryllium hazards, stability for stored synthesized molten salt fuel, and gas generation and off-gas systems fuel cycle facilities as novel characteristics of some MSRs. Beryllium is a known industrial hazard (10 CFR 850 2023; OSHA n.d.), but its use in current fuel cycle facilities is limited. The existing beryllium safety guidance (29 CFR 10910.1024 2022; OSHA n.d.) could be adopted, or specific NRC beryllium safety guidance could be developed within the MSR fuel cycle facilities.

The assessment of transportation packages identified the potential for gas generation from fuel constituents, which may cause difficulty for the package containment design. When designing a

PNNL-33899, Rev. 1 Executive Summary iii package, consideration of the potential for over-pressurization after sealing and subsequent loss of containment may need to be considered.

PNNL-33899, Rev. 1 Acronyms and Abbreviations iii Acronyms and Abbreviations ACU Abilene Christian University AEC Atomic Energy Commission, the predecessor of the U.S. Nuclear Regulatory Commission and U.S. Department of Energy AgNO3 silver nitrate ANR Advanced Nuclear Reactor ANSI American National Standards Institute ARE Aircraft Reactor Experiment BeF2 beryllium fluoride BWR boiling water reactor CCl4 carbon tetrachloride CFR Code of Federal Regulations 37Cl chlorine-37 isotope Cl2 molecular chlorine CoC Certificates of Compliance COCl2 phosgene (carbonyl chloride)

DOE U.S. Department of Energy DOT U.S. Department of Transportation FLiBe LiF-BeF2 salt mixture GDC general design criteria H2 molecular hydrogen HAC hypothetical accident condition HALEU high-assay low-enriched uranium HCl hydrochloric acid (hydrogen chloride)

He helium HEU highly enriched uranium HF hydrogen fluoride HNO3 nitric acid IROFS items relied on for safety IMSR Integral Molten Salt Reactor INL Idaho National Laboratory ISA integrated safety analysis kg kilogram KP-FHR Fluoride Salt-Cooled High-Temperature Reactor LEU low-enriched uranium Li lithium 6Li lithium-6 isotope

PNNL-33899, Rev. 1 Acronyms and Abbreviations iv 7Li lithium-7 isotope LiF lithium fluoride LiOH lithium hydroxide, also LiOHaq LSA-1 low specific activity LWR light-water reactor MC&A Material Control and Accounting MCFR Molten Chloride Fast Reactor mol%

mole percentage MSR molten salt reactor MSRE Molten Salt Reactor Experiment MSRR Molten Salt Research Reactor MWe megawatts-electric MWt megawatts-thermal mT millitesla (a unit of magnetic flux density)

NaF sodium fluoride Na2UF6 sodium fluoride and uranium tetrafluoride salt mixture NaZrF5 sodium fluoride and zirconium fluoride salt mixture NCT normal conditions of transport NH4F ammonium fluoride NO nitric oxide NOF nitrosyl fluoride NRC U.S. Nuclear Regulatory Commission NUREG Nuclear Regulatory Report NUREG/CR Nuclear Regulatory/Contractor Report ORNL Oak Ridge National Laboratory PNNL Pacific Northwest National Laboratory PWR pressurized water reactor 239Pu plutonium-239 isotope RRIP DOEs Research Reactor Infrastructure Program SiC silicon carbide SNM special nuclear material SSC system, structure, and component TEUSA Terrestrial Energy USA TRIGA Training, Research, Isotopes, General Atomics Nuclear Research Reactor TRISO tristructural isotropic fuel particle U

uranium 233U uranium-233 isotope

PNNL-33899, Rev. 1 Acronyms and Abbreviations v

235U uranium-235 isotope UF4 uranium tetrafluoride UF6 uranium hexafluoride UO2 uranium dioxide (uranium oxide) wt%

weight percent Y-12 Y-12 National Security Complex ZrF4 zirconium fluoride

PNNL-33899, Rev. 1 Contents vii Contents Executive Summary...................................................................................................................... ii Acronyms and Abbreviations........................................................................................................ iii Contents....................................................................................................................................... vii 1.0 Introduction....................................................................................................................... 1 1.1 Report Nomenclature............................................................................................ 1 1.2 Report Objectives.................................................................................................. 2 1.3 Report Organization.............................................................................................. 2 2.0 Background....................................................................................................................... 4 2.1 Prior Operating Experience in Front-End MSR Operations................................... 4 2.1.1 Aircraft Reactor Experiment.................................................................... 4 2.1.2 MSRE..................................................................................................... 4 2.2 Current MSR Designs........................................................................................... 5 2.3 Integral Molten Salt Reactor under Development (Terrestrial Energy USA)......... 8 2.4 Fluoride Salt-Cooled High-Temperature Reactor (Kairos Power)......................... 8 2.5 Molten Chloride Fast Reactor (TerraPower/Southern Nuclear)............................. 9 2.6 Molten Salt Research Reactor (ACU).................................................................... 9 3.0 Regulatory Frameworks for Molten Salt Reactor Front-End Operations......................... 10 3.1 10 CFR Part 40, Domestic Licensing of Source Material.................................... 10 3.2 10 CFR Part 70, Domestic Licensing of Special Nuclear Material...................... 10 3.3 10 CFR Part 71, Packaging and Transportation of Radioactive Material............ 11 3.3.1 General Information.............................................................................. 14 3.3.2 Structural.............................................................................................. 14 3.3.3 Thermal................................................................................................. 15 3.3.4 Containment......................................................................................... 15 3.3.5 Shielding............................................................................................... 16 3.3.6 Criticality............................................................................................... 16 3.3.7 Materials............................................................................................... 16 3.3.8 Operating Procedures........................................................................... 17 3.3.9 Acceptance Test and Maintenance Program........................................ 17 3.3.10 Quality Assurance................................................................................. 18 3.4 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities and Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants....................................................................................................... 18 4.0 Potential Approaches to Front-End Fuel Cycle Operations............................................ 19 4.1 Recovery and Conversion................................................................................... 19 4.2 SNM Enrichment................................................................................................. 19 4.3 Special Nuclear Material Fabrication and Away-From-MSR Salt Mixing............. 22

PNNL-33899, Rev. 1 Contents viii 4.4 Nonradioactive Carrier Salts............................................................................... 22 5.0 Fresh MSR Fuel Feed Extraction and Concentration Operations................................... 25 5.1 Production Facilities............................................................................................ 25 5.1.1 Baseline LEU/UO2 Domestic Experience............................................. 25 5.1.2 HALEU Domestic Experience............................................................... 25 5.1.3 Possible Advanced Fuel Pathways....................................................... 26 5.1.4 Fabrication/Fuel Synthesis................................................................... 27 5.2 Storage................................................................................................................ 27 5.2.1 Short-Term Storage.............................................................................. 28 5.2.2 Long-Term Storage............................................................................... 28 5.3 Summary of Safety Hazards and Constraints..................................................... 29 6.0 Summary......................................................................................................................... 31 6.1 Recovery............................................................................................................. 31 6.2 Fuel Cycle Facilities............................................................................................ 31 6.3 Transportation..................................................................................................... 32 6.4 At-Reactor Fuel Processing................................................................................. 32 7.0 References...................................................................................................................... 33

PNNL-33899, Rev. 1 Figures and Tables ix Figures Figure 1 Onsite Fuel and Salt Mixing and Processing......................................................... 3 Figure 2 Offsite Fuel and Salt Mixing and Processing......................................................... 3 Figure 3 Candidate Approaches for MoltenSalt Reactor Fuel Front-End Operations........ 21 Figure 4 Typical Feedstock Flows to Support Advanced Reactor Fuel Fabrication Needs Using UF6 Feed........................................................................................ 26 Tables Table 1 Characteristics and Updated Status of MSR Technologies Under Commercial Development..................................................................................... 6 Table 2 Applicable Regulatory Frameworks for Front-End Activities per MSR Fuel or Material Types................................................................................................. 20 Table 3 Advantages and Challenges of Separate License for Front-End Fuel Cycle Operations from a Reactor License........................................................... 23 Table 4 Safety Hazards and Constraints Associated with Fresh MSR Fuel Feed Extraction and Concentration Operations............................................................ 29

PNNL-33899, Rev. 1 Introduction 1

1.0 Introduction For this report. the U.S. Nuclear Regulatory Commission (NRC) asked Pacific Northwest National Laboratory (PNNL) to assess hazards and safety impacts related to front-end activities for the potential fuel cycles of molten salt reactors (MSRs). PNNLs assessment relies only on publicly available information and includes summaries of four different MSR technologies1 (see Section 9.2 for detailed information about the technologies), the regulatory framework that may apply to the technologies, and information that can be considered when engaging with applicants. For the purposes of this report, the front-end includes fuel cycle activities relating to uranium (U) recovery, production (synthesis),2 transportation, and processing (e.g., mixing) of fresh (unirradiated) fissile fuel materials. This includes primary activities up to the insertion of fuel into a reactor.

1.1 Report Nomenclature On January 14, 2019, the President signed the Nuclear Energy Innovation and Modernization Act into law. Section 103 of the law requires the NRC to develop a regulatory framework to support developing and commercializing advanced nuclear reactors (ANRs). The law aims to provide a program to develop the expertise and regulatory processes necessary to allow innovation and the commercialization of advanced nuclear reactors (NEIMA 2019).

This report uses the following relevant definitions from the law applicable to the task:

1. ANR means a nuclear fission or fusion reactor, including a prototype plant, with significant improvements compared to commercial nuclear reactors under construction as of the date of the enactment of the law. An ANR may have greater fuel use (burnup level), increased proliferation resistance, or increased thermal efficiency.

2. Regulatory framework means the framework for reviewing requests for certifications, permits, approvals, and licenses for nuclear reactors.

3. Technology-inclusive regulatory framework means a framework developed using evaluation methods that are flexible and practicable for application to a variety of reactor technologies, including, where appropriate, the use of risk-informed and performance-based techniques and other tools.

For this document, vendor refers to the MSR technology developer or designer, and the vendor and applicant can be two different organizations, where one is a technology supplier, and the other is the MSR owner or operator that could apply to the NRC for a permit or license.

This report focuses on one type of ANR, namely MSRs, which are nuclear-fission non-light water reactors. NRC generally uses the term advanced non-light water reactor to describe non-light water reactors.

1 (1) The Integral Molten Salt Reactor from Terrestrial Energy USA, (2) the Fluoride Salt-Cooled High-Temperature Reactor from Kairos Power, (3) the Molten Chloride Fast Reactor from TerraPower and Southern Nuclear, and (4) the Molten Salt Research Reactor from Abilene Christian University.

2 Production means converting yellowcake into uranium hexafluoride.

PNNL-33899, Rev. 1 Introduction 2

1.2 Report Objectives The NRC requested PNNL to assess hazards and safety impacts related to front-end operations of advanced non-light water reactor that use molten salt as a coolant or within a liquid fuel.

PNNL researchers assessed the front-end operations using information from 1960s Atomic Energy Commission (AEC) programs for the Aircraft Reactor Experiment (ARE) (Cottrell 1954, 1955a, 1955b) and the Molten Salt Reactor Experiment (MSRE) (Peretz 1996), the current state of knowledge of the U.S. Department of Energy (DOE), and commercial vendor plans and fuel compositions. This report provides the physical and chemical characteristics of associated MSR fuels, as available from public documents and presentations.

The front-end operations fall into the following four broad categories:

x recovery of materials x

fuel cycle facilities x

transportation of fresh fuel x

at-reactor fuel mixing operations Because of the uncertainty related to front-end operations of fuel cycles for MSRs, this report considers potential approaches that the vendors and applicants could pursue for front-end operations and the relevance of the different regulatory frameworks (Title 10 of the Code of Regulations Part 50 [10 CFR Part 50], 10 CFR Part 70, and 10 CFR Part 71) in the various steps of those approaches. The primary NRC related regulatory documents for these technologies are within Title 10 of the Code of Regulations (CFR): 10 CFR Part 50, 10 CFR Part 70, and 10 CFR Part 71.

Different licensing strategies may be used based on an applicants descriptions of its front-end operations and the requests in the application. For an LWR fuel cycle, all fuel fabrication is done at an away-from-reactor facility, and the only activity after transporting the fuel to the reactor site is loading fuel into the reactor following brief onsite storage. Because of the unique nature of an MSR, some fuel cycle processes could be undertaken at the reactor site, as shown in Figure 5, as opposed to at a separate facility, depicted in Figure 6.

1.3 Report Organization In this report, Section 9.0 provides the fuel cycle background for the MSRs and details about the historical U.S. MSR experiments. Section 10.0 describes the regulatory frameworks for front-end operations of MSR fuels. Section 11.0 provides potential front-end fuel cycle approaches.

Section 12.0 describes the fresh MSR fuel feed extraction and concentration operations.

Section 13.0 provides a summary of the assessment. Section 14.0 provides the references.

PNNL-33899, Rev. 1 Introduction 3

Figure 1 Onsite Fuel and Salt Mixing and Processing Figure 2 Offsite Fuel and Salt Mixing and Processing

PNNL-33899, Rev. 1

Background

4

2.0 Background

2.1 Prior Operating Experience in Front-End MSR Operations The front-end fuel cycle operation of MSRs represents a promising pathway for ANRs, potentially providing environmental and safety benefits. The MSRs being considered today are an evolution of research experiments from the mid-20th century: the ARE and the MSRE.

The ARE, conducted in the 1950s, was the first demonstration of an MSR system aimed at powering long-range aircraft. Although the Air Force and the AEC did not develop the aircraft, the ARE pioneered the fundamental mechanics of MSRs, focusing on using a circulating molten fluoride salt fuel.

The MSRE, a subsequent development in the 1960s, further refined the MSR concept. It was the first reactor to use uranium-233 (233U) fuel generated from thorium-232 (232Th) while demonstrating the feasibility of liquid fuel in a graphite-moderated core.

2.1.1 Aircraft Reactor Experiment Through the Aircraft Reactors Branch, the AEC supported the Aircraft Nuclear Propulsion Project at Oak Ridge National Laboratory (ORNL) to develop a molten salt aircraft propulsion reactor, which was the ARE reactor. During the ARE, ORNL constructed and operated a thermal reactor in the early 1950s (Cottrell 1955a, 1955b). The ARE operators added fuel by combining sodium fluoride (NaF) and uranium tetrafluoride (UF4) using 93.4%-enriched 235U into a heated carrier mixture of NaF and zirconium fluoride (ZrF4). Some literature refers to the concentrate as sodium fluoride and UF4 salt mixture (Na2UF6) and the carrier as sodium fluoride and zirconium fluoride salt mixture (NaZrF5), which are the chemical formulas of the mixtures after combining the components.

A review of the project reports developed by ORNL (Cottrell 1954) and the AEC provided no further insight into any specific details about the development or the production of UF4 beyond that the fuel was manufactured by Y-12 National Security Complex (Y-12) Production Division to permit control of the 253 pounds (lb) (approximately 115 kilograms [kg]) of highly enriched uranium (Cottrell 1954). The Y-12 Production Division provided 15 batches of approximately 30 lb (13.6 kg) of the concentrate to the ARE facility (Cottrell 1954). The two laboratories maintained good alignment of the analytical results of the specific amounts of 235U despite using different analytical methods and some early faulty sampling of poorly milled UF4. The Y-12 laboratory used spectrographic analyses to determine the quantity and type of metallic impurities in the concentrate, whereas ORNL used chemical analyses to quantify the metallic impurities (Cottrell 1954).

2.1.2 MSRE The Y-12 facility produced the MSRE reactor fuel and coolant salts using existing AEC fissile material stock and commercial materials. Lithium (Li) separation operations at Y-12 enriched Li to over 99.99% 7Li by further enriching the tailings. Fuel salts were produced by bubbling hydrogen and hydrogen fluoride through the precursor materials to remove moisture, sulfur, and free oxygen. Then, residual hydrogen was removed using a helium (He) sparge. The fuel salt was nominally 65 mole percent (mol%) lithium fluoride (LiF), 29.1 mol% beryllium fluoride (BeF2), 5 mol% ZrF4, and 0.9 mol% UF4, while the coolant and flush salt were 66 mol% LiF and

PNNL-33899, Rev. 1

Background

5 34 mol% BeF2. Initial operations were conducted with 33 weight percent (wt%) 235U; however, in 1968, the enriched U was replaced with 233U; the fissile material was again replaced in 1969 with PuF3 (94 wt% plutonium-239 isotope [239Pu]) (McFarlane et al. 2019).

While the ARE and the MSRE had sufficient fuel for the limited time the AEC and ORNL operated the reactors, there were no extended operations for either experiment to drive a need for indefinite front-end fuel cycle operations specific to these MSR designs.

In the construction permit application (ML22227A203) at Abilene Christian University (ACU),

the applicant references the MSRE fuel composition. The ACU fuel salt composition of 67:28:1 LiF-BeF2-UF4 is similar to the MSRE fuel salt composition of 65:29:5:1 LiF-BeF2-ZrF4-UF4 (NRC 2022a).

2.2 Current MSR Designs While multiple commercial vendors, international organizations, and countries are developing MSR technologies, this report focuses on the technologies listed below whose vendors have engaged with the NRC:

1. Integral Molten Salt Reactor (IMSR) from Terrestrial Energy USA (TEUSA)

2. Fluoride Salt-Cooled High-Temperature Reactor (KP-FHR) from Kairos Power

3. Molten Chloride Fast Reactor (MCFR) from TerraPower and Southern Company

4. Molten Salt Research Reactor (MSRR) from ACU Table 5, reproduced from Torres et al. (2022), summarizes the available information for these MSR technologies. The sections immediately following Table 5 provide specific information concerning the front-end fuel cycle considerations.

PNNL-33899, Rev. 1

Background

6 Table 1 Characteristics and Updated Status of MSR Technologies Under Commercial Development (Torres et al. 2022)

Technology IMSR(a)

KP-FHR(b)

MCFR(c)

MSRR(d)

Commercial Vendor TEUSA Kairos Power TerraPower and Southern Company ACU Base Model Output (MWe, MWt) 195 MWe; 442 MWt 140 MWe (full-scale)

Kairos 1 Hermes Unit 35 MWt non-power reactor facility.

Kairos 2 Hermes - Two-unit 35-MWt non-power reactor facility with a shared steam powered conversion system.

500-1200 MWe (grid-scale);30-300 MWe (mid-scale) 1 MWt Neutron Spectrum Thermal Thermal Fast Thermal Moderator Graphite Graphite Multiple materials under consideration Graphite Heat Transfer Medium (Coolant)

Molten fluoride salt Molten fluoride salt Molten chloride salt Molten fluoride salt Outlet Temperature

~700qC (1292qF) 650qC (1,202qF)

Unknown 590°C Thermal Efficiency 44%

45% (net)

Unknown Not applicable Fuel Enrichment (235U)

<5%

19.75%

~12%

19.9%

Fuel Form and Composition Fluoride salt TRISO (U oxycarbide)

Chloride salt UF4 in a carrier salt mixture

(~67%) LiF-(~28%) BeF2-(~5%) UF4 Li enriched to >99.99% in 7Li Salt Coolant Composition Proprietary fluoride salt Primary salt: FLiBe (enriched with 7Li)

Secondary salt: NaNO3/KNO3 mixture called solar Proprietary chloride salt Secondary salt: FLiBe Initial Reactor Demonstration/Siting Unknown Low-power demonstration reactors at the East Tennessee Technology Park in Oak Ridge. Hermes Reactor expected to be operational in 2026 The small-scale Molten Chloride Reactor Experiment will be located at the Idaho National Laboratory site and will inform the design of a small to mid-scale MCFR demonstration reactor ACU MSRR in Abilene, Texas

PNNL-33899, Rev. 1

Background

7 Technology IMSR(a)

KP-FHR(b)

MCFR(c)

MSRR(d)

DOE Support U.S. Industry Opportunities for Advanced Nuclear Technology Development ARDP Risk-Reduction Awardee ARDP Risk-Reduction Awardee RRIP Awardee DOE Laboratory Partner Argonne National Laboratory Oak Ridge National Laboratory Idaho National Laboratory Undecided NRC Status Pre-application engagement.

Regulatory engagement plan submitted in October 2021.

Kairos-NRC issued Construction Permit CPTR-6 December 14, 2023.

Kairos 2-Construction Permit Application for two units submitted to the NRC on July 14, 2023. (Docket Numbers 50-611 and 612)

Pre-application engagement (limited).

Construction Permit application submitted to the NRC on August 12, 2022.

(a) Hill 2020 (b) Kairos 2020, 2024; NRC 2023a, 2024; Haugh 2021 (c) Latkowski 2021, TerraPower 2020 (d) Schubert 2020, NRC 2022b IMSR = Integral Molten Salt Reactor; KP-FHR = Fluoride Salt-Cooled High-Temperature Reactor; MCFR = Molten Chloride Fast Reactor; MSRR = Molten Salt Research Reactor; TEUSA = Terrestrial Energy USA; MWe = megawatt electric; MWt= megawatt thermal; TRISO = tristructural isotropic fuel particle; FLiBe = LiF-BeF2 salt mixture; ARDP = DOE Advanced Reactor Demonstration Program; RRIP = DOE Research Reactor Infrastructure Program

PNNL-33899, Rev. 1

Background

8 2.3 Integral Molten Salt Reactor under Development (Terrestrial Energy USA)

The IMSR is intended to operate with 2-3 wt% enrichment of 235U fuel as a molten fluoride salt circulating through the reactor at start-up, with the fuel addition enrichment increasing up to 4.95 wt% 235U (Hill 2020).

The fuel cycle of the IMSR involves a 7-year operation cycle, during which makeup fuel is added to increase the total salt volume by about 50%. After 7 years, two-thirds of the final salt volume is piped as start-up fuel for the next core-unit, and one-third is sent to nearby storage in an adjacent vault. This cycle is repeated for 8 cycles or 56 years. The makeup fuel has a low starting 235U enrichment, <2%, and allows the makeup fuel additions to remain standard assay

(<5% enrichment) in low enough volume to avoid the need for any fuel salt to be removed over the 7-year design life of each core-unit. This fuel cycle offers many benefits, including a significant improvement of uranium use, lower plutonium content in the used fuel, significantly reduced buildup of used fuel within the facility, and the fact that such used fuel then represents possible starting material for additional IMSR onsite or even offsite (Choe et al. 2018).

TEUSA and Orano recently completed an evaluation that determined that existing nuclear fuel packaging is suitable for the intended IMSR fuel due to its standard low enrichment. TEUSA is deploying a multiple-sourcing strategy for fuel procurement. Orano was identified as one of its partners for fuel enrichment, chemical conversion, and transportation (Terrestrial Energy 2022b). TEUSA has signed a similar agreement with Westinghouse and the United Kingdoms National Nuclear Laboratory for nuclear fuel development and supply (Terrestrial Energy 2021).

Based on the written materials, it is anticipated that the fuel suppliers will mix the fuel stocks and salts off-site and transport the fuel salts to the site for fuel loading operations.

TEUSA partnered with Argonne National Laboratory to develop a fuel salt testing program through the DOE Gateway for Accelerated Innovation in Nuclear program. The fuel salt is expected to be measured to determine thermophysical properties and to estimate fission product solubility (Terrestrial Energy 2022a).

2.4 Fluoride Salt-Cooled High-Temperature Reactor (Kairos Power)

Kairos has issued a topical report detailing its fuel performance methodology (Kairos 2020).

Kairos plans to fuel its reactor using TRISO particles. The TRISO particles consist of a U oxycarbide fuel kernel surrounded by a porous carbon buffer layer, an inner pyrolytic carbon layer, a silicon carbide (SiC) layer, and an outer pyrolytic carbon layer. The TRISO fuel form provides an additional barrier to releasing fission products (DOE 2019). The reactor core is designed to have fuel pebbles with TRISO particles embedded in a graphite matrix material covered by a fuel-free outer shell.

In 2022 Kairos submitted a topical report to support the licensing of the KP-FHR, Fuel Qualification Methodology for the Kairos Power Fluoride Salt-Cooled High-Temperature Reactor (KP-FHR) (Kairos 2022). In Section 2.2.3 of the report, Kairos states that the fuel pebbles may contain enriched 235U up to 19.55% by weight.

PNNL-33899, Rev. 1

Background

9 2.5 Molten Chloride Fast Reactor (TerraPower/Southern Nuclear)

No specific information regarding TerraPower/Southern Nuclear plans for fuel enrichment, fabrication, and transportation for the MCFR has been identified. For the MCFR, there is no public information on any final specification of the fissile material to be used. Based on public information, one of the fissile material possibilities is UCl3 (McFarlane et al. 2019).

2.6 Molten Salt Research Reactor (ACU)

The MSRR is a loop-type, 1-MWt molten fluoride salt reactor with UF4 dissolved in the salt operating at a maximum temperature of 650°C. The nominal fuel salt composition is ~67 mol%

LiF, ~28 mol%, BeF2-, ~5 mol% UF4. The final composition is to be developed under the Nuclear Energy eXperimental Testing Laboratory Quality Assurance Program with the involvement of DOE and reported in the operating license application. The uranium is anticipated to be approximately 19.75% enriched in 235U, and the Li enriched to be greater than 99.99% in 7Li. A hexagonal lattice of graphite blocks with flow channels moderates the reactor vessels fuel salt in the core. Criticality is prevented in other regions of the primary salt loop and also in the fuel handling system by geometry and a lack of neutron moderation (NRC 2022a).

The MSRR is fueled with a uranium-bearing, fluoride-based salt containing minimal oxidative impurities. The mass of the fuel salt will be approximately 1,600 kg (3,527 lb) and is a thoroughly mixed combination of less than 500 kg (1,102 lb) of UF4 and approximately 1,100 kg (2,425 lb) of LiF-BeF2 (ACU 2023).

The MSRR project is supported by a cooperative fuel agreement with DOEs RRIP. The fuel will be owned by the U.S. government and loaned to ACU by the DOE. ACU intends for fuel to be premixed and ready to be loaded when it comes onsite. No specific information regarding ACUs plans for fuel transportation for the MSRR has been identified. In addition to receiving the premixed fuel, UF4 is added to the base fuel salt when the reactor is subcritical. The additional UF4 is added during start-up to compensate for the depletion of fissile material during operation using high-assay low-enriched uranium (HALEU). The construction permit application does not describe the fuel salt addition system, and ACU will provide the isotopic composition and system description in the operating license application.

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 10 3.0 Regulatory Frameworks for Molten Salt Reactor Front-End Operations The regulatory frameworks pertinent to front-end activities for potential MSR fuel cycles were evaluated by reviewing Title 10 of the CFR and supporting NRC NUREGs. It is noted that additional regulatory frameworks from other competent authorities, such as the U.S. Department of Transportation (DOT) or the Environmental Protection Agency, are outside the scope of this report and were not evaluated (e.g., regulations in Title 49 of the CFR). The review focused on identifying information needs related to potential regulatory approaches driven by differences in MSR front-end design and operations (e.g., mixing at an off-reactor-site vs. mixing at an on-reactor-site), which are further discussed in Section 12.0.

Fuel fabrication facilities convert enriched uranium into fuel for nuclear reactors (NRC 2020a).

TRISO fuel for solid-fueled MSRs would be fabricated using similar processes outlined on the NRC Fuel Fabrication page (NRC 2020a). However, for liquid-fueled MSRs, fuel fabrication will look very different from the solid-fueled MSR fuel fabrication process. Fuel fabrication would consist of mixing the UF4 or UCl3 with a carrier salt, and this could be carried out at the reactor site.

3.1 10 CFR Part 40, Domestic Licensing of Source Material The regulations in 10 CFR Part 40 (10 CFR 40) are promulgated by the NRC under the Atomic Energy Act of 1954, as amended (AEA 1954), Title II of the Energy Reorganization Act of 1974 (88 Stat. 1242) (42 U.S.C 23 § 2011) and Titles I and II of the Uranium Mill Tailings Radiation Control Act of 1978, as amended (Uranium Mill Tailings Radiation Act 1978). The regulations in 10 CFR Part 40 establish procedures and criteria for issuing licenses to receive title to and receive, possess, use, transfer, or deliver source and byproduct materials and establish and provide for the terms and conditions upon which the NRC will issue such licenses. Uranium recovery operations supporting MSRs are not expected to deviate from those supporting the fuel cycle for current LWR technologies; therefore, it is thought that the requirements in 10 CFR Part 40 will broadly apply.

3.2 10 CFR Part 70, Domestic Licensing of Special Nuclear Material 10 CFR Part 70 applies to MSRs by regulating the possession and use of special nuclear material (SNM), assuring criticality safety, enforcing stringent safety and security measures, and requiring comprehensive emergency planning and environmental protection. Compliance with these regulations is essential for the safe and secure operation of MSRs. MSRs use SNM.

10 CFR Part 70 requires stringent controls over the accounting and handling of these materials.

This includes measures for preventing theft or diversion, maintaining accurate records, and regular reporting to the NRC.

The regulations mandate comprehensive safety and security measures for facilities handling SNM. For MSRs, this means implementing robust safety systems to manage the unique risks associated with liquid fuels, such as high-temperature operation and potential release of radioactive materials. Part 70 includes provisions to prevent accidental criticality events. MSRs must be designed and operated to make sure the configuration of fuel and other materials does not lead to an unintended chain reaction. This involves detailed criticality safety evaluations and the implementation of safety controls.

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 11 Operations at an MSR would require a 10 CFR Part 70 license to possess and use SNM. The regulations in 10 CFR Part 70 establish procedures and criteria for the issuance of licenses to receive title to own, acquire, deliver, receive, possess, use, and transfer SNM associated with fuel cycle facilities, including uranium conversion, enrichment, and fuel fabrication and establish the terms and conditions upon which the NRC will issue such licenses.

The safety regulations in 10 CFR Part 70 are primarily performance-based. Therefore, the requirements appear adequate for licensing and regulation of fuel facilities related to uranium enrichment and fuel production for MSR materials. The regulations adequately address the HALEU feed material needs for MSR technologies in development.

Subpart H of 10 CFR Part 70 identifies risk-informed performance requirements and requires applicants and existing licensees to conduct an integrated safety analysis (ISA). Subpart H of 10 CFR Part 70 also requires adherence to baseline design criteria, as specified in 10 CFR 70.64, Requirements for New Facilities or New Processes at Existing Facilities, for new facilities that have not already been designed, built, licensed, and operated. An ISA identifies potential accident sequences in the facilitys operations, designates items relied on for safety (IROFS) to either prevent such accidents or mitigate their consequences to an acceptable level, and describes management measures to provide reasonable assurance of the availability and reliability of IROFS. Section 3.0 of NUREG-1520 provides guidance for the review of the ISA program, particularly for controlling (1) radiological hazards, (2) chemical hazards that could increase radiological risk, (3) facility hazards that could increase radiological risk, (4) potential accident sequences, (5) consequences and the likelihood of each accident sequence, and (6) IROFS, including the assumptions and conditions under which they support compliance with the performance requirements of 10 CFR 70.61.

3.3 10 CFR Part 71, Packaging and Transportation of Radioactive Material The regulations in 10 CFR Part 71 (10 CFR 71) establish requirements for the packaging, preparation for shipment and transportation of radioactive material, and procedures and standards for NRC approval of packaging and shipping procedures for such licensed material.

This includes any material containing radionuclides for which both the activity concentration and the total activity in the consignment exceed the values specified in the table in 49 CFR 173.436, or values derived according to the instructions in 49 CFR 173.433 (see 49 CFR 173.403)

(NRC 2011). The regulations in 10 CFR 71 establish both:

x requirements for packaging, preparation for shipment, and transportation of licensed material x

procedures and standards for NRC approval of packaging and shipping procedures for fissile material and other licensed material above a Type A quantity (see 10 CFR 71.0)

It is important to note that the packaging and transport of licensed material also are subject to other NRC regulatory frameworks (e.g., 10 CFR Parts 20, 21, 30, 40, 70, and 73) and to the regulations of other agencies (e.g., the DOT) that have jurisdiction over means of transport.

The regulations in 10 CFR Part 71 (10 CFR 71) are largely performance-based, and no gaps were identified concerning activities related to transporting fresh (unirradiated) radioactive MSR fuel materials. Transportation of unirradiated fuel elements with coated particle fuel has been previously conducted to support the Fort Saint Vrain High-Temperature Gas-Cooled Reactor.

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 12 The shipments of these elements based on fissile (up to 93.5 wt% 235U) and fertile particle fuel were approved by the AEC and the DOT. The NRC has also recently approved a transportation package for UF6 enriched up to 20 wt% 235U. However, the NRC has not yet approved package designs for MSR salt fuel materials, which will likely present the most technical challenges. The following discussion provides a review of the regulations and regulatory review guidance that may require specific consideration because of the novel risks associated with the chemical and radiological attributes of MSR fuel salt materials. The discussion assumes transport of the MSR fuel salts in a Type B (unirradiated, fissile material) package containing < 20 wt% 235U.

10 CFR 71.31, Contents of the application, paragraph (c), discusses the requirement to identify established codes and standards proposed for use in package design, fabrication, testing, maintenance, and use, or otherwise justify the basis and rationale used to formulate the package quality assurance program. Although consensus codes and standards are available to evaluate material performance for packaging components, applicants must independently identify credible chemical reactions and relevant material environments where the materials must be tested to demonstrate adequate performance for all transport conditions. Weld performance for the confinement components for the MSR fuel salts may require increased scrutiny due to the corrosive nature of the salts.

x 10 CFR 71.35, Package evaluation, paragraph (c), discusses that the application identifies any proposed special controls and precautions for transport, loading, unloading, and handling and any proposed special controls in case of an accident or delay. The controls and precautions for the operations will likely differ from the transport of traditional LWR fuels to protect personnel from increased risks due to the chemical reactivity of MSR fuel salts due to inadvertent exposure to air or moisture.

x 10 CFR 71.43, General standards for all packages, paragraph (c), discusses that each package must include a containment system closed by a positive fastening device that cannot be opened by a pressure that may arise within the package. The potential for corrosive gases and increased pressures from chemical reactions between MSR fuel salts and the packaging or anticipated environments may need to be considered. Paragraph (d) also states that a package must be made of materials and construction that assure there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the package contents, including possible reactions from in-leakage of water, to the maximum credible extent. Account must be taken of the behavior of materials under irradiation.

Demonstrating compliance with this paragraph may necessitate a comprehensive assessment of potential chemical reactions from the MSR fuel salt contents. The performance of the packaging components is also an essential consideration for compliance with 10 CFR 71.43(f) and 10 CFR 71.71.

x 10 CFR 71.55, General requirements for fissile material packages, provides requirements to ensure that the package is designed so the contents remain subcritical under normal conditions of transport (10 CFR 71.55(b), (d)) and hypothetical accident conditions (10 CFR 71.55(b), (d)). The enrichment and configuration of the package's MSR fuel salts will differ relative to LWR fuels, and compliance with these requirements may need increased scrutiny. It is also noted that 10 CFR 71.55(g) allows for an exception to 10 CFR 71.55(b) for packages containing UF6 enriched up to 5 wt% 235U. Because front-end activities will necessitate UF6 enriched beyond this limit, packages will either need an exemption for 10 CFR 71.55(b) or will need to be designed and fabricated to remain subcritical for the most reactive credible configuration for the UF6 contents if water were to leak into the package containment.

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 13 x

10 CFR 71.71, Normal conditions of transport, and 10 CFR 71.73, Hypothetical Accident Conditions, provide requirements for evaluating a package per prescribed conditions and tests. These analyses may require increased consideration for ensuring adequate packaging performance for potential reconfiguration of the contents.

x 10 CFR 71.85, Preliminary determinations, paragraph (b), states the need for additional containment system testing before a package with high maximum normal operating pressures is used. MSR fuel salts should be confined inside the package containment boundary in separate containers. The performance of these containers will be an important consideration for defining the applicability of this requirement.

The NRC published guidance for the NRC staff review of applications and acceptable methods and guidance for meeting the requirements for transportation of licensed material. Those documents include the following:

x NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material (Smith et al. 2019) x NUREG/CR-6407, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety (McConnell et al. 1996)

NUREG-2216 addresses the transport of low-enriched UF6 (i.e., < 20 wt% 235U), but the discussion is focused on historical practices involving the shipment of UF6 enriched to <5 wt%,

consistent with 10 CFR 71.55(g). The NRC has endorsed using American National Standards Institute (ANSI) N14.1, Nuclear MaterialsUranium HexafluoridePackagings for Transport, which specifies the design and fabrication of the UF6 cylinder (NRC 2020b). ANSI N14.1 and USEC-651, The UF6 Manual: Good Handling Practices for Uranium Hexafluoride, provide information regarding UF6 overpacks. It is noted that the DOT requires UF6 packaging be designed, fabricated, inspected, tested and marked in accordance with(i) American National Standard N14.1 in effect at the time the packaging was manufactured (see 49 CFR 173.420(a)(2)(i)).

ANSI N14.1 describes standard UF6 cylinders for transport and their characteristic limits. The most common UF6 cylinder for transport for enrichments up to 5 wt% 235U is the 30B standard design, with a nominal diameter of 30 inches and a maximum fill mass of 5,020 lb. By contrast, the standard cylinders for transport of enrichments beyond 5 wt% 235U provide for much lower volumes and masses. For instance, the 8A standard design allows for enrichments up to 12.5 wt% 235U with a maximum fill mass of 255 lb, whereas the 5A and 5B standard designs allow for enrichments up to 100 wt% 235U but a maximum fill mass of 54.9 lb. Therefore, ANSI N14.1 standard designs reach much smaller volumes for shipments of HALEU UF6, than for enrichments of under 5 wt% 235U. The reevaluation of ANSI N14.1 appears to be an important consideration to assess the possibility of additional designs for safe and cost effective HALEU UF6 shipments for MSR fuel cycles.

The technical review disciplines in NUREG-2216 will have the same regulatory acceptance requirements independent of the unirradiated fissile material form. NUREG-2216 was reviewed to identify items requiring increased consideration when reviewing fresh MSR fuel salts relative to standard LWR fresh fuel assemblies. It is important to note that input parameters and assumptions for package evaluation should broadly consider the quality control of the MSR fuel types for fabrication, as the rate of deficiencies should be incorporated in the physical and chemical specifications of the contents.

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 14 3.3.1 General Information The general information section of safety analyses typically requires specific details. While the general information section may include novel content and configurations unrelated to currently licensed transportation casks, it remains distinct from most LWR fuel packages. Unlike LWR fuel packages, which share common design principles and licensing strategies, the general information section must be carefully evaluated based on reactor technology, licensing approach, and operational plans.

x physicochemical form and material properties:

describe the physicochemical form of the material, including density, moisture content, and any moderating constituents address materials prone to chemical, galvanic, or other reactions, including the potential generation of combustible gases x

fissile material details:

identify the maximum quantity of fissile material, specifying fissile nuclides, concentrations, enrichments, and masses x

location and configuration:

discuss the location and configuration of contents, including the need for secondary containers, wrapping, or shoring x

geometry control and safety measures:

provide information on spacers or other features used for geometry control or confinement of fissile material highlight safety impacts related to these spacers (e.g., maintaining subcriticality) x neutron absorbers and moderators:

clearly identify and quantify neutron absorber materials or moderators During safety reviews, independent verification of supporting data is essential. Making sure appropriate test methods and quality control are implemented during testing is crucial. For additional guidance, refer to NUREG/CR-5502, which outlines technical content for different intended functions of packaging drawings.

3.3.2 Structural The structural evaluation will remain largely the same as conducted for radioactive material packages for loose or odd-shaped contents (e.g., powders). Additional consideration may be provided to ensure that the content configurations are adequately bounded during transit or drop scenarios, which may be accomplished using a separate single container or multiple sub-containers within the packages primary containment vessel. These approaches have been previously implemented for other types of radioactive materials such as in the TN Versa-Pac package and generally would require ensuring the adequate structural performance of the separate containers and any necessary internal structures. The extent of potential reconfiguration or shifting of MSR fuel salts and TRISO pebble compacts within the package is an important consideration, notably for criticality safety during normal conditions of transport (NCT) and hypothetical accident condition (HAC). The evaluation of the structural models and analyses may need coordination with the materials review to ensure the assumed material

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 15 properties are adequate for the temperatures and loads of these conditions. Any potential breach of separate containers for MSR fuel salts during HAC drop scenarios would impact the reactivity of the salts when exposed to potential environmental conditions. Further, any gas generation during transport inside these containers during NCT and HAC would likely need to be evaluated for sealed canisters without pressure relief devices.

3.3.3 Thermal The thermal evaluation for MSR fuel salt and TRISO pebble contents will be like evaluations of LWR fuel packages. For fresh fuel, there is no decay heat generation. The evaluation of temperatures related to any seals or material performance limits will be the same as in an LWR package. The thermal performance of MSR fuel salts during an HAC thermal test should be carefully considered because, depending on the package configuration and material properties, it may be possible to approach the melting temperature of different salts. There is no specific prohibition against a change in physical form, but it could affect the assumptions and input parameters for other safety reviews, particularly containment, materials, and criticality.

It is noted that the review of thermal material properties for the fuel contents may necessitate increased rigor, particularly for limited datasets and nonstandard characterization methods.

These thermal material properties are generally needed to evaluate the package performance during NCT and HAC tests and may include thermal conductivity, specific heat, density, and coefficient of thermal expansion.

3.3.4 Containment The containment evaluation will depend on the package type and the performance of the MSR fuel forms or its secondary containers. Type A (fissile) packages must demonstrate no loss or dispersal of the radioactive material contents during NCT tests (see 10 CFR 71.43(f)) and should demonstrate that the fuel form will be maintained in a known geometry. Type B packages must satisfy specific release rates (see 10 CFR 71.51), and compliance would likely be based on a similar experience for other radioactive material packages. Although a Type B package may not be required for fresh fuel, it may be needed if fuel is derived from downblending or recycling. The NRC has endorsed the use of ANSI N14.5 for the design and testing of leak-tight packages that do not require additional release rate calculations and analyses. For packages that are not leak-tight per ANSI N14.5, the containment evaluation would be expected to verify the acceptability of design calculations for volumetric and reference air leakage rates. ANSI N14.5 defines an acceptable method for calculating these values and appears valid for MSR fuel forms. The fuel material release risks are higher for MSR salts than TRISO pebbles. TRISO particles are robust because of their multiple barrier coatings and outer SiC pressure boundary layer, which protects the outer pyrolytic carbon layer. The pebbles graphite-based matrix material further protects the particles, providing significant mechanical strength against external forces.

On the other hand, MSR fuel salts would be dispersible. NUREG/CR-6487 provides methods for determining acceptance standard leakage rates for Type B packages with solid SNM dispersible contents, which may assist in the containment evaluation for these package designs. An important consideration in these leakage rate analyses will be the pressure conditions for the backfill gas to be used inside the primary containment boundary.

Transport of MSR fuel types is not expected to generate combustible gases during NCT and HAC test scenarios; however, the potential generation of corrosive species may need to be considered for impacts on the assumed leakage rates and pressure limits for the primary

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 16 containment boundary. Sealed secondary containers for MSR fuel salts may mitigate the effects if adequately justified. The basis for the maximum normal operating pressure for evaluating NCT and HAC tests may need increased consideration in the review.

3.3.5 Shielding The shielding evaluation for MSR fuel salts and TRISO pebble contents is not expected to differ from standard LWR fuels. In general, fresh fuel packaging requires little shielding due to unirradiated contents. This is not likely to change for MSR fuel types.

3.3.6 Criticality The criticality evaluation may need to carefully consider the potential for reconfiguring the MSR fuel contents assuming that fuel geometry is relied on to maintain subcriticality. Compared to an LWR fuel assembly, there are no structural components in the MSR fuel forms, whether TRISO pebbles or salts. Any structures used to prevent reconfiguration will be part of the packaging. A package design may likely incorporate single or multiple secondary containers for the MSR fuel forms to mitigate issues related to inadvertent chemical reactions during transport, particularly for fuel salts. These containers also could be relied on for moderator exclusion, in which the design prevents water ingress to the fuel under HAC scenarios. This licensing approach would necessitate that the design defines credible and bounding fuel configurations that maximize reactivity. NUREG-2216 describes review guidance for this approach when implemented for LWR spent nuclear fuel assemblies.

3.3.7 Materials Material interactions are important to consider in the package design review. The fuel materials may degrade the packaging, which can affect structural performance and the containment function of the package. Special care should be taken to evaluate elevated temperatures and long-term performance. Because packaging is usually in service for multiple shipments, interactions that occur over time must be considered.

Although TRISO pebbles are expected to be physically and chemically stable under NCT and HAC tests, MSR fuel salts will behave very differently due to their corrosive and oxidizing characteristics. The materials review may need increased consideration of potential reactions within the contents and with package components. The corrosion rates of metallic packaging components and welds should be justified per data for the expected loading and transport conditions, including temperatures, moisture availability, time of wetness, atmospheric contaminants, and oxidizing species. The use of coatings that may include nickel or chromium in the primary containment vessel and other secondary containers would potentially assist in mitigating corrosion and facilitate decontamination, which would necessitate a performance basis and evaluation of other inadvertent chemical reactions between the MSR salts and the coatings (e.g., generation of hydrogen or other combustible gases). It is important to note that NUREG-2216 states that these coatings should not be credited for protecting the substrate material or extending the useful life of the substrate material unless a periodic coating inspection and maintenance program is required for the coating.

It is unclear whether the temperatures reached by the MSR fuel salts during the HAC thermal test would lower the melting temperature of metallic materials of the salt containers. For instance, the reactivity of uranium with steel materials may need increased scrutiny because of its potential to lead to eutectic reactions. The potential for these eutectic reactions to

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 17 compromise the structural performance of structural and containment components would need to be examined to define bounding package contents.

Outgassing from the MSR salt fuel and subsequent pressure impacts may require an evaluation per the content chemical specification and temperatures for the NCT and HAC tests. The supporting technical basis may want to consider the criteria and procedures for loading the MSR salts in secondary containers and salt environments (e.g., residual water and oxygen). These are like considerations previously implemented for transporting vitrified high-level waste, from which lessons learned may be gathered.

The structural performance of the TRISO pebbles may be credited in demonstrating bounding fuel configurations and the extent of dispersible material, whereas this will not be the case for MSR salt fuels. Therefore, mechanical properties of the TRISO pebbles may be needed (e.g.,

pebble graphite using resin binder material, SiC of the TRISO particles) (Wells et al. 2021).

TRISO pebbles have a different structural configuration than LWR fresh fuel (i.e., a configuration of fuel particles enveloped in a composite material consisting of graphite and resin binder instead of the LWR configuration of uranium dioxide pellets with zirconium-based alloy cladding) (Wells et al. 2021). Therefore, defining the specific mechanical properties needed to adequately assess pebble performance may require additional consideration per the proposed acceptance criteria used to demonstrate pebble integrity.

Regarding other packaging materials (e.g., bolts and seals), the guidance in NUREG-2216 appears adequate. The performance of these components also may be affected by any chemical reactions and generated gases from the contents during transport. Review of the maintenance program for the package should be coordinated with the materials review to make sure activities are appropriate.

3.3.8 Operating Procedures The guidance in NUREG-2216 for the safety review of operating procedures is generic enough to apply to MSR fuel types. Operating procedures for package loading and unloading for MSR fuel salts will be different from LWR fuels. The procedures should provide acceptance criteria per the safety analyses of other disciplines, which would protect the salt contents and prevent or mitigate the interaction of the contents with other packaging components. The procedures for handling secondary containers for TRISO pebbles and salt fuel forms would need to identify special handling equipment, controls, or precautions during loading and unloading. The specifics of these activities will vary depending on the design basis and certification approach for each package.

3.3.9 Acceptance Test and Maintenance Program The guidance in NUREG-2216 for the safety review of the acceptance test and maintenance program is generic enough to apply to MSR fuel types. The acceptance tests are developed based on the input parameters, assumptions, and conclusions from the analyses for the other safety disciplines of the package. NUREG-2216 notes that the simplicity of the information provided may vary, but initial package reviews for MSR fuel salts will likely necessitate significant detail since no consensus codes and standards are available specifically for these fuel types. Some acceptance tests will be like those per prior operating experience for standard LWR fuels or other radioactive material, such as visual inspections, weld examinations, structural and pressure tests, and shielding tests. Depending on the certification approach, testing of components and materials for packaging components (e.g., testing materials and leak

PNNL-33899, Rev. 1 Regulatory Frameworks for Molten Salt Reactor Front-End Operations 18 testing of secondary containers used for the MSR fuel salts within the primary containment vessel) may address the performance of coatings or corrosion resistance of the materials used in these containers. The maintenance program may also include activities to replace materials exposed to these fuel salts.

3.3.10 Quality Assurance The quality assurance requirements and review process are identical for all fissile material packages, regardless of contents. Therefore, no substantive changes are expected for this part of the package review.

3.4 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities and Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants There is high degree of uncertainty regarding the approaches that will be taken for fresh fuel management at MSR plants and the design features for fresh fuel storage and handling facilities. Further, whether fuel salt mixing operations will be conducted at a separate facility or integrated with the reactor is unclear. Under 10 CFR Part 50, these activities would require approved principal design criteria that likely would differ to some degree from criteria used for LWR operations. An assessment of considerations for principal design criteria for MSRs is not detailed in this report, as it is considered out of scope.

PNNL-33899, Rev. 1 Potential Approaches to Front-End Fuel Cycle Operations 19 4.0 Potential Approaches to Front-End Fuel Cycle Operations MSRs using liquid fuel, and to a lesser extent TRISO-particle-based fuels, require management of materials that present different hazards than traditional LWR UO2 fuel encapsulated in a zirconium alloy cladding, and the new fuels will require new safety assessments without regulatory precedent. The NRCs expectations for safety assessments of front-end operations will be based on the applicable requirements depending on the chosen licensing approach.

Torres et al. (2022) and Holcomb et al. (2022) have discussed potential pathways for licensing fuel cycle facilities supporting MSRs.

Table 6 provides applicable NRC regulatory frameworks for front-end operations for the fuel and carrier salt materials for both liquid and solid-fueled MSRs. Figure 7 describes the potential approaches for these operations.

4.1 Recovery and Conversion The regulatory framework for source material recovery and conversion (i.e., uranium) is provided in 10 CFR Part 40. Uranium ore processing (i.e., milling) into yellowcake concentrate is independent of downstream operations related to fuel conversion, enrichment, and production technologies. Therefore, uranium recovery operations supporting MSRs will not differ from those supporting the fuel cycle for current LWRs.

Transportation of uranium ore concentrate would be conducted with LSA-1 non-fissile or fissile-excepted packages (see 10 CFR 71.4 and 71.14). Transportation would be subject to 10 CFR Part 71 and DOT requirements (e.g., 49 CFR Part 173, Subpart I).

4.2 SNM Enrichment The regulatory framework for uranium enrichment is provided in several parts within Title 10 of the CFR, including 10 CFR Part 70, which addresses the possession and use of SNM, and 10 CFR Part 40, which addresses the possession and use of natural and depleted uranium. In the United States, the two facilities that have been licensed for 235U enrichment currently are operational at different scales. The Louisiana Energy Services, LLC gas centrifuge plant at Eunice, New Mexico, is the only operational enrichment plant with commercial capacity licensed to produce UF6 enriched up to 5.5 wt% 235U (NRC 2023b). The licensee, URENCO-USA, intends to pursue approval for a higher enrichment up to 10.0 wt% 235U (Freels 2021). The American Centrifuge Plant licensed Centrus Energy Corp. (Fitch 2022) has also been approved to operate a lead cascade of 16 AC-100M enrichment centrifuges to demonstrate the capability to produce up to 600 kg (1,323 lb) of HALEU UF6 at their Piketon, Ohio facility (Zimmerman 2021). The licensee has since requested approval to increase the volume in a phased approach (up to 900 kg [1,984 lb]) (Fitch 2022). The facility is anticipated to be used for near-term HALEU production.

Conversion and enrichment of 233U may not be conducted in an NRC-licensed facility under 10 CFR Part 70. The DOE holds 233U enriched material in the form of oxides (i.e., UO2 and triuranium octoxide at ORNL), which could be processed at a DOE facility not subject to regulations at 10 CFR Part 70.

PNNL-33899, Rev. 1 Potential Approaches to Front-End Fuel Cycle Operations 20 Table 2 Applicable Regulatory Frameworks for Front-End Activities per MSR Fuel or Material Types(a, b, c)

MSR Type Fuel or Material Type Target Fuel Enrichment Source Material Recovery/ Extraction Operations Enrichment Activities Production/

Fabrication Activities Mixing/Handling/Storage Activities Transportation Activities Liquid (Salt)

Fueled 233UF4 Unknown 233U specification DOE inventory 10 CFR Part 70 (Category 2) 10 CFR Part 70 (Category 2)

Away from reactor:(d) 10 CFR 40, 70 (Category 2)

At-reactor: (d) 10 CFR 40, 50/52, 70 (Category 2) 10 CFR Part 71 235UF4 or 235U chlorides Low-enriched uranium (LEU) <5 wt% 235U 10 CFR Part 40 10 CFR Part 70 (Category 3) 10 CFR 70 (Category 3)

Away from reactor: (d) 10 CFR 70 (Category 3)

At-reactor: (d) 10 CFR 50/52, 70 (Category 1) 10 CFR Part 71 235UF4 or 235U chlorides LEU <20 wt% 235U 10 CFR Part 40 10 CFR Part 70 (Category 2) 10 CFR Part 70 (Category 2)

Away from reactor: 10 CFR 70 (Category 2)

At-reactor: 10 CFR 50/52, 70 (Category 2) 10 CFR Part 71 Nonradioactive Carrier Salts Not applicable Not applicable Not applicable Commercial sourcing Away from reactor: (d) 10 CFR 40, 70 At-reactor: (d) 10 CFR 40, 50/52, 70 Commercial transport Solid-Fueled TRISO compacts LEU <20 wt% 235U 10 CFR Part 40 10 CFR Part 70 (Category 2) 10 CFR Part 70 (Category 2) 10 CFR Part 50/Part 52, Part 70 10 CFR Part 71 (a) 10 CFR Part 20 requirements generally apply to all NRC-licensed facilities.

(b) Environmental Protection Agency, DOT, and other regulatory frameworks may be applicable.

(c) Facilities possessing or using SNM subject to the requirements in 10 CFR Parts 73 and 74.

(d) Mixing/handling operations for fuel salt materials involving 232ThF4 may need a 10 CFR 40 license for source material possession and use.

PNNL-33899, Rev. 1 Potential Approaches to Front-End Fuel Cycle Operations 21 Figure 3 Candidate Approaches for Molten Salt Reactor Fuel Front-End Operations

PNNL-33899, Rev. 1 Potential Approaches to Front-End Fuel Cycle Operations 22 Transportation of enriched UF6 (235U or 233U) would be conducted per the requirements of 10 CFR Part 71 in acceptable cylinders. It is important to note that shipments between an enrichment facility and a fuel production facility, each located at a DOE site, may not require NRC approval. The DOT, in 49 CFR 173.7(d), authorizes the use of packaging made by or made under the direction of the DOE for the transportation of Class 7 materials when evaluated, approved, and certified by the DOE against packaging standards equivalent to those specified in 10 CFR 71. However, the DOE generally seeks approval from the NRC for packages unrelated to its environmental management activities or national defense mission.

4.3 Special Nuclear Material Fabrication and Away-From-MSR Salt Mixing The term fuel fabrication, as used in NRC regulations in 10 CFR Part 70, includes the conversion from enriched uranium to another chemical form and also includes the mechanical assembly of material into a metal-clad cylindrical or plate structure. With respect to MSRs, the fuel described in Table 5 and Table 6 are either liquid fueled or use TRISO, which could be in a cylindrical or a spherical structure, and use enriched uranium for the fissile material.

A recommendation in this report is to use terminology inclusive of MSR technology. Three NRC-licensed Category 3 fuel fabrication facilities have been licensed to produce fuel based on LEU (i.e., fuel in which the wt% of 235U in the uranium is less than 20%; see 10 CFR 50.2 and 10 CFR 110.2), as well as two Category 1 facilities that can process HEU for research reactors and military requirements (i.e., fuel with 235U in the uranium of 20 wt% or greater; see 10 CFR 50.2 and 10 CFR 110.2) (NRC 2020a).

Several MSR technologies will use HALEU as their fissile fuel material. HALEU is not defined in NRC regulations, but it has been defined in Title 42 of the U.S. Code (42 USC § 16281[d][4]) as uranium having an assay greater than 5.0 wt% and less than 20.0 wt% of 235U. There are no currently licensed Category 2 fuel cycle facilities for HALEU production, although TRISO-X submitted an application (NRC 2022c) for a fuel fabrication plant to produce TRISO compacts for advanced reactors.

SNM fuel fabrication activities will likely involve the conversion of enriched UF6 to the forms needed for MSR operation (e.g., UF4, uranium chlorides, or TRISO compacts). These operations would likely be conducted at a separate facility. The conversion facility also could receive nonradioactive carrier salts to mix and prepare MSR salt fuels. Alternatively, the enriched UF4 or uranium chlorides could be shipped to another stand-alone facility or the MSR site, licensed per 10 CFR 70, which would separately receive the carrier salts for mixing. TRISO compacts would be shipped directly to the MSR site without additional processing.

Transportation of MSR fuel forms and salt mixtures would likely be conducted using Type B (unirradiated, fissile) packages subject to the 10 CFR Part 71 requirements.

4.4 Nonradioactive Carrier Salts Nonradioactive carrier salts (e.g., FLiBe) are currently expected to be procured from commercial sources, and their transportation to a facility for mixing (stand-alone or MSR site) would not be subject to 10 CFR Part 71 requirements.

PNNL-33899, Rev. 1 Potential Approaches to Front-End Fuel Cycle Operations 23 Operations at the MSR site will include storing and handling SNM and source material, which may include mixing fissile materials with the nonradioactive carrier salts. The applicable regulatory frameworks for these activities would be 10 CFR Parts 50 or 52 (reactor licensing),

10 CFR Part 40 (possession and use of source material), and 10 CFR Part 70 (possession and use of SNM). It is noted that the NRC has been developing an alternative technology-inclusive and risk-informed regulatory framework for advanced reactors under draft 10 CFR Part 53.

Because that framework has not been finalized and is subject to change, it is not further considered in this report.

MSR fuel salt mixing may be performed in a separate facility at the MSR site or integrated into the reactor fuel system. Both approaches will require 10 CFR Part 40 and 10 CFR Part 70 licenses, which may be pursued jointly or independently from the 10 CFR Part 50 or 10 CFR Part 52 license. An integrated 10 CFR Part 70 and 10 CFR Parts 50/52 license would require normalization of the safety classification of systems, structures, and components (SSCs) and radiological performance requirements. The regulations in 10 CFR Part 70 require controls for high-consequence event workers (where a worker is defined as an individual who receives an occupational dose as defined in 10 CFR 20.1003). Consideration of how these controls would need to be harmonized with 10 CFR Parts 50 and 52 may be needed, which include different categorizations, such as those associated with control room operators. Proposed licensing approaches for advanced reactors (e.g., NEI 2019) may be flexible given that harmonizing key requirements and expectations can be achieved to avoid redundancy in requirements of 10 CFR Parts 50 and 70. Some advantages and challenges associated with a separate 10 CFR Part 70 license from a reactor license are listed in Table 7 (Holcomb et al. 2022).

Table 3 Advantages and Challenges of Separate License for Front-End Fuel Cycle Operations from a Reactor License Advantages Challenges Simplification of reactor safety analysis Demonstrating a strong technical basis for the logical and physical separation between the reactor and fuel cycle facility Separation of fuel cycle from reactor-related SSC safety functions and safety classifications Developing a separate ISA for the fuel cycle facility and reactor license application in which some duplication may be necessary Avoidance of technical specifications and limiting conditions of operation of fuel cycle SSCs with respect to reactor safety Operating and maintaining two separate facilities for which the site and total facility design may require some duplication (e.g.,

control rooms, electrical distribution areas, etc.)

Fuel cycle facility operations and trained operator qualifications could be simplified and separated from reactor operator training and qualifications Need for potential additional interfacing system components (e.g., two pumps may then be necessaryone for the reactor and one for the fuel cycle facilityinstead of just one pump)

Inspections, testing, and maintenance activities are not tied to reactor safety and can be developed in accordance with 10 CFR Part 70 requirements and safe operation of the fuel cycle facilities

PNNL-33899, Rev. 1 Potential Approaches to Front-End Fuel Cycle Operations 24 The final decision for a separate 10 CFR Part 70 license may result from several factors. These may include the quantity of material, material form, procedures required to incorporate the material into the reactor system, and potential other unknowns at this time. Material enrichment will also play a role for certain reactor designs, as a differentiator in 10 CFR Part 70 is the split above and below 10 wt% enrichment of 235U for determining quantity. Unique hazards associated with onsite fuel mixing include potential inadvertent criticality, chemical hazards, and release pathways and process interactions, which may differ depending on where the operations are conducted.

The quantity of material stored for online refueling will play a role in the requirements for a 10 CFR Part 70 license, as certain thresholds require licensing. If 235U is the only fissile material in the fresh fuel, then this quantity will be specific to the amount of stored 235U. Quantities of materials that meet the thresholds of SNM of Moderate or Low Strategic Significance would meet this requirement.

Some designs have proposed using materials containing 233U and 239Pu after initial start-up.

These isotopes would also require licensing under 10 CFR Part 70 if the material added during operation contains more than 15 g (0.5 ounces) of 233U and 239Pu combined mass.

Subpart H to 10 CFR Part 70 applies to each applicant or licensee that is or plans to be authorized to possess greater than a critical mass of SNM, and engaged in enriched uranium processing, fabrication of uranium fuel or fuel assemblies, uranium enrichment, enriched UF6 conversion, plutonium processing, fabrication of mixed-oxide fuel or fuel assemblies, scrap recovery of SNM, or any other activity that the Commission determines could significantly affect public health and safety (10 CFR 70). Mixing salts and SNM would likely require a license under Subpart H to 10 CFR Part 70 because the process would require more than a critical mass of SNM and is performing fuel fabrication by mixing the SNM and salts.

PNNL-33899, Rev. 1 Fresh MSR Fuel Feed Extraction and Concentration Operations 25 5.0 Fresh MSR Fuel Feed Extraction and Concentration Operations For this report, information was collected and reviewed regarding experience producing, storing, and transporting fresh or unirradiated liquid MSR fuel materials and precursors. These activities included an assessment of the state of knowledge of DOE plans for the production or recovery of HALEU fuel feed material, including the chemical processes and characteristics (e.g.,

potential residual contaminants) being pursued by the DOE as well as the potential use of these processes at an NRC-licensed facility.

The following sections provide general considerations during the production, storage, and transportation of MSR liquid fuel precursors, intermediate products, and final products awaiting use in a nuclear reactor. The list was developed from the perspective of the existing front end of the fuel cycle to produce UO2 fuel fabrication supporting the current fleet of light-water reactors.

This fuel cycle may evolve, and the list is not exhaustive.

5.1 Production Facilities Production facilities associated with MSR liquid fuel synthesis have some commonalities with existing Category 3 commercial fuel cycle facilities producing UO2-based fuel forms and many distinct differences. Common areas are summarized in the next section, followed by a review of potential distinctions. Then, general considerations for MSR fuel cycle facilities are provided.

5.1.1 Baseline LEU/UO2 Domestic Experience All LEU fuel fabrication in the United States is performed in a commercial Category 3 facility where UF6 is converted to UO2. The origin of the material is either virgin uranium or HEU down-blended material that typically meets broadly accepted uranium isotopic specifications for LEU material. In the United States, no organization currently works with re-enriched reprocessed uranium LEU material (i.e., uranium recovered in a Plutonium Uranium Extraction Plant process, converted to UF6, and enriched to a higher level). This material currently can be found in Europe and Russia; however, this process was done in the past by U.S. companies. Such material requires additional handling requirements for radiological reasons.

5.1.2 HALEU Domestic Experience Legacy commercial experience in working with HALEU in the United States does exist. One example was the Training, Research, Isotopes, General Atomics Nuclear Research Reactor (TRIGA) fuel fabrication facility in San Diego, California, which has long since been decommissioned. This capability now is provided by the TRIGA International fuel fabrication facility in Romans, France. That facility is proposed to provide HALEU fuel to Idaho National Laboratory (INL) to support projects such as Microreactor Applications Research Validation and Evaluation using feed material from the Y-12 facility at ORNL. Isotopic feed material specifications and transportation plan specifications are defined in the INL report INL/RPT 66550 (Johnson et al. 2022).

Today, facilities owned by BWX Technologies, Inc. in Lynchburg, Virginia, and Erwin, Tennessee, provide the only domestic source of finished fuel product with >10% enrichment, including for the U.S. Navy, High Flux Isotope Reactor, and HALEU fuel used in the United States High-Performance Research Reactor conversion program. The facilities are rated

PNNL-33899, Rev. 1 Fresh MSR Fuel Feed Extraction and Concentration Operations 26 Category 1, so experiences are different from that expected for a HALEU focused facility.

However, there is existing documented information related to transportation and HALEU isotopic values that provide relevant knowledge.

5.1.3 Possible Advanced Fuel Pathways Like existing LEU facilities, virgin enriched material from a HALEU enrichment capability will enter the MSR fuel cycle as UF6 and then be converted to another form. Conversion to metal, uranium mononitride, UF4, or UO2 is well understood for LEU enriched <5 wt% and that experience can be extended to HALEU enrichments where geometries of processing vessels, piping, and processes are adapted for more stringent criticality concerns. Waste streams, scrap recovery, effluents, and chemical hazards are well understood and generally scalable from LEU experience. Figure 8 illustrates numerous pathways for fuel forms starting from UF6. The TRISO and molten chloride fuel forms may be derived in a multi-step process involving a dry conversion or using ammonium diuranate consistent with current LEU/UO2 conversion processes.

Figure 4 Typical Feedstock Flows to Support Advanced Reactor Fuel Fabrication Needs Using UF6 Feed (Zbib et al. 2021)

PNNL-33899, Rev. 1 Fresh MSR Fuel Feed Extraction and Concentration Operations 27 5.1.4 Fabrication/Fuel Synthesis Starting with existing LEU/UO2 conversion processes, a rudimentary gap analysis highlights some considerations for a fuel cycle facility built for advanced reactor fuel forms. However, without detailed process flow sheets, it is not possible to identify specific anticipated differences from existing fuel cycle facilities:

Several of the proposed salts for use as fuel or coolant contain known hazardous chemicals. In the case of FLiBe (which was also used in the MSRE), beryllium and beryllium compounds are listed as known human carcinogens (U.S. Department of Health and Human Services 2011).

Hydrogen fluoride and chloride gas, which could potentially be used in the synthesis and purification of fuel salts, have noted health impacts. Additionally, pyrophoric materials involved in fuel processing may require evaluation in the context of impacts on an industrial scale fire suppression system design. Halide-based fuel and coolant salts also can be highly corrosive under oxidizing conditions. Therefore, chemical degradation of process piping and storage containers may need to be considered.

Other considerations are:

x High-temperature exothermic reactions may be used for synthesis of compounds or substituents (e.g., to make actinide metal); consequently, potentially novel facility design will need to be considered.

x Gaseous byproducts (e.g., molecular hydrogen [H2]) produced by the process (intentionally or not) may need to be considered regarding off-gas system requirements.

x The presence of fissile isotopes and forms with less developed criticality analyses, those outside existing standards, and the introduction of new moderators not previously considered (or not addressed in depth) in existing standards will need to be considered.

Criticality analyses of the geometry, material compositions, and neutron moderation to determine the conditions under which a self-sustaining nuclear chain reaction might occur may need to be fully developed.

x Generation of mixed low-level wastes (mixed chemical and radioactive risks) during processing (for example, radioactive wastes and contaminated articles such as booties and gloves or filters that also contain beryllium contamination).

x Fuel cycle facility decontamination and decommissioning challenges (e.g., mixed wastes, unusually tenacious contaminants, etc.) that would require additional planning and costs.

x Recovery is an aqueous or hazardous chemical process to recover fissile material and high-worth chemicals from an off-specification product, such as uranium. Recovery may introduce new hazards, such as potential dissolution of the SiC layer of off-specification TRISO particles was conducted.

Effects of a short-term or long-term facility shutdown are addressed with storage considerations below.

5.2 Storage Below are considerations for storing and handling intermediate and finished products for use at the fuel cycle facility or an interim or new fuel-receipt area (stationary material not contained within a transportation package). The intermediate product may be stockpiled as equivalent to UO2 powder awaiting pelletizing or pellets waiting to be loaded into cladding for LWR fuel.

PNNL-33899, Rev. 1 Fresh MSR Fuel Feed Extraction and Concentration Operations 28 Stockpiling could be part of a load-leveling effort, awaiting analytical chemistry results, etc.

Material held up in vessels and processing equipment during the temporary shutdown could fall into this category.

5.2.1 Short-Term Storage Short-term accumulation of finished products can occur while awaiting final product certification, packaging, transportation, etc. (DOE 2000). Some possible differences from current experience include:

1. Fissile isotopes and forms with less characterized criticality analysis methods and those outside existing benchmarks, and introduction of new moderators (perhaps beryllium-bearing compounds) not previously considered or addressed in depth in existing standards.

2. Secondary containment methods for aqueous solution spills, including criticality considerations.

3. Accumulated material should have multiple barriers to criticality. New or different barriers may be used.

5.2.2 Long-Term Storage Below are additional considerations for extended (multiyear, decadal) storage beyond short-term considerations. These considerations may come into play if a fuel cycle facility builds ahead of current needs for first or initial reactor cores and then a start-up is delayed.

LWR fuel is generally stable in the final fabricated form. There are mechanical considerations on structure materials, such as spacer grid springs deformation if placed horizontally in containers.

But overall, the UO2 form, cladding, and structure are chemically, mechanically, and radiologically stable. However, consideration for new processes, such as complex salt chemistry that is stored indefinitely, may be warranted. Examples include:

1. Chemical attack on storage vessel, cladding, piping, etc., by the primary constituent

2. Chemically unstable compounds that undergo changes during storage:

x gaseous hydrogen generation with resulting flammability considerations x gas generation (H2, He, etc.) from chemical decay may lead to an overpressure x UO2 and hydrofluoric acid generation if fuel salts combine with atmospheric water vapor x new minor constituents with corrosive properties

3. Decay of radioactive isotopes

a. Radiation hazards

b. gas generation (e.g., fluorine gas, UF6) from radioactive decay or radiolysis effect on hydrogenous materials leading to an overpressure

c. Build-in of isotopes with adverse neutronic properties affecting future reactor operation along with new radiation hazards (different energy or radiation hazard category)

d. Self-mobile isotopes that result in the dispersion of materials Understanding the effects of prolonged deferment of facility operations and storage of materials is helpful in the regulatory space from the viewpoint of understanding consequences when a

PNNL-33899, Rev. 1 Fresh MSR Fuel Feed Extraction and Concentration Operations 29 facility license to operate is suspended. For example, the consequence of not properly addressing material holdup and in-process chemicals when a facility is urgently shut down may not be immediately clear for novel or new processes.

5.3 Summary of Safety Hazards and Constraints The hazards associated with liquid fuel processing were extensively documented by McFarlane et al. (2019). They are briefly summarized in Table 8 (taken from Table 3a in McFarlane et al.

[2019]). Each row refers to an activity, such as a chemical process or monitoring activity. The activity is briefly described in the first column. The second column describes the salt or component being processed and the goals of the process. The third column lists important hazards associated with the activity. The fourth column lists ways to mitigate the hazards (McFarlane et al. 2019).

Table 4 Safety Hazards and Constraints Associated with Fresh MSR Fuel Feed Extraction and Concentration Operations. Reproduced from McFarlane et al. (2019)

Physical or chemical process Salt type and process objective Key hazards Mitigation strategies x

Salt synthesis x

Colex(a) 7Li production x

Mercury x

Caustic (LiOH) x Alternatives to using mercury are being assessed x

Salt synthesis x

Electromigration 7Li production x

Heat x

Salt synthesis x

Crown ether 7Li production x

Organic chemicals x

Caustic (LiOHaq) x Fire x

Salt synthesis x

Liquid diffusion Chlorine-37 isotope (37Cl) production x

Carbon tetrachloride (CCl4), chloroform x

Pressure: 5 bar x

Salt synthesis x

Anion exchange chromatography 37Cl production x

Nitric acid (HNO3) and nitrogen oxide gases x

AgNO3 x

Salt synthesis x

NaCl crystallization 37Cl production x

20-80 millitesla (mT) magnetic field x

Salt synthesis x

Reduction with CCl4 Conversion: UF6 to UF4 x

U contamination x

CCl4, molecular chlorine (Cl2),

phosgene (COCl2) x Standard contamination control for unirradiated materials x

Neutralize acidic and toxic gases x

Salt synthesis x

Reduction with NOx Conversion: UF6 to UF4 x

U contamination x

NOx, nitrosyl fluoride (NOF), HF, H2 x

Standard contamination control for unirradiated materials x

Neutralize acidic and toxic gases

PNNL-33899, Rev. 1 Fresh MSR Fuel Feed Extraction and Concentration Operations 30 Physical or chemical process Salt type and process objective Key hazards Mitigation strategies x

Salt synthesis x

Fluorination Conversion: oxides to fluorides x

HF, H2 x

UO2+x, contamination x

Standard contamination control for unirradiated materials x

Neutralize acidic and toxic gases x

Salt synthesis x

Reaction with ammonium bifluouride Conversion: oxides to fluorides x

400qC x

Ammonium fluoride (NH4F), ammonium bifluouride x

Standard contamination control for unirradiated materials x

Salt purification x

Hydrofluorination Removal of oxides/hydroxides from fluoride salts x

H2/HF x

Be if present as the BeF2 salt component x

Heat (650qC) x Contamination if actinides present x

Standard contamination control for unirradiated materials x

Neutralize acidic and toxic gases x

Salt purification x

Chlorination Removal of oxides/hydroxides from chloride salts x

CCl4, Cl2, COCl2, hydrochloric acid (HCl) x Heat x

Contamination if actinides present x

Standard contamination control for unirradiated materials x

Neutralize acidic and toxic gases x

Packaging and transportation Actinides and non-fissile components x

Contamination if actinides present x

Air-sensitive x

Be (if present as BeF2) x Double barrier container (moderator exclusion under all operating conditions)

(a)

COLEX is the Li hydroxide-mercury amalgam column exchange-based separation process and was used to separate 6Li from 7Li.

PNNL-33899, Rev. 1 Summary 31 6.0 Summary This report provides the results of assessing the state of hazards and safety impacts related to front-end activities for potential MSR fuel cycles for five MSR technologies. The assessment was performed by comparing publicly available information to NRC regulations and guidance.

Information needs were identified to conduct licensing and other regulatory activities relating to the uranium recovery, production (synthesis), transportation, and processing (e.g., mixing) of fresh (unirradiated) MSR fissile fuel materials. Although there are five publicly available MSR technology designs, the information review was limited when considering licensing activities.

Specifically, the publicly available specific design information related to the strategy for the front-end fuel cycle is generally vague, likely to protect proprietary information, or underdeveloped as MSRs are an emerging reactor technology.

From this assessment, current regulations and guidance appear to be appropriate for most MSR fuel types, and no necessary updates were identified.

Four operational areas were identified: (1) recovery, (2) fuel cycle facilities, (3) transportation, and (4) at-reactor fuel. Specific information needs in each area are provided below.

6.1 Recovery Recovery is the process to extract and concentrate the uranium. The assessment concludes that the 10 CFR Part 40 regulations appear to adequately address uranium-related activities as source material. Uranium recovery operations supporting MSRs are not expected to deviate from those supporting the fuel cycle for current LWR technologies. Therefore, no gaps were identified for current safety review guidance for these facilities.

6.2 Fuel Cycle Facilities Several MSR vendors are designing their MSRs to use HALEU as fuel (i.e., material with uranium having an assay greater than 5.0 wt% and less than 20.0 wt% of 235U). However, there are no NRC-licensed Category 2 fuel cycle facilities for HALEU. One fuel vendor, TRISO-X LLC (NRC 2022c), has applied for a construction and operating license for a fuel fabrication facility to manufacture TRISO-based fuels using HALEU.

The assessment concludes that the safety regulations in 10 CFR Part 70 are largely performance-based, and no gaps were identified concerning regulations of activities related to uranium enrichment and conversion for MSR fuel materials. The fabrication of liquid fuels and mixing of SNM and salt mixtures would require a license issued under 10 CFR Part 70 and would be subject to the requirements in Subpart H of 10 CFR Part 70.

The assessment identified beryllium hazards, stability for stored synthesized molten salt fuel, including gas generation, and off-gas systems in fuel cycle facilities as novel characteristics of some MSRs. Beryllium is a known industrial hazard (10 CFR 850 2023; OSHA n.d.), but it does not have much application in current fuel cycle facilities. Existing beryllium safety guidance (29 CFR 10910.1024 2022; OSHA n.d.) could be adopted or specific NRC beryllium safety guidance could be developed within the MSR fuel cycle facilities.

PNNL-33899, Rev. 1 Summary 32 6.3 Transportation As mentioned above, the assessment concludes that safety regulations in 10 CFR Part 71 are largely performance-based, and no gaps were identified concerning activities related to transporting fresh (unirradiated) MSR fuel materials. However, NRC-approved package designs for these fuels are not available. Likewise, NRC-approved package designs specifically for transporting a factory-built and fueled reactor are unavailable.

A specific consideration is the potential for gas generation from any fuel constituents, which may cause difficulty for the package containment design. When designing a package, consideration of the potential for over-pressurization after sealing and subsequent loss of containment may need to be considered.

6.4 At-Reactor Fuel Processing One novel characteristic of fuel processing for MSRs is the ability to collocate the fuel cycle facility at the reactor site. Normally, the fuel cycle facility is not collocated at the reactor site, and the fresh fuel is shipped from the fuel cycle facility to the reactor site, and two separate licenses are required. However, an applicant for a collocated fuel cycle facility and MSR may consider combining their applications for a fuel cycle facility and an MSR operating license using NRC regulations at 10 CFR 50.31 or 10 CFR 52.8. This arrangement of having a fuel processing facility at the reactor site represents a different operational configuration than typically found for LWRs.

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