ML24256A003
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| Issue date: | 09/17/2024 |
| From: | Wendy Reed NRC/RES/DE |
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Text
Technical Letter Report
[TLR-RES/DE/REB-2024-15]
Storage and Transportation of Molten Salt Reactor Wastes:
Identification of Technical Information Needs and Safety Implications for Safety Review Guidance Date:
September 2024 Prepared in response to Back-End Task 1 in User Need Request NMSS-2022-002, by:
Patrick LaPlante CNWRA Biswajit Dasgupta CNWRA Yi-Ming Pan CNWRA George Adams CNWRA 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
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.
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.
STORAGE AND TRANSPORTATION OF MOLTEN SALT REACTOR WASTES:
IDENTIFICATION OF TECHNICAL INFORMATION NEEDS AND IMPLICATIONS FOR SAFETY REVIEW GUIDANCEFINAL REPORT Prepared for U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Prepared by Patrick LaPlante, Biswajit Dasgupta, Yi-Ming Pan, and George Adams Southwest Research Institute Center for Nuclear Waste Regulatory Analyses September 2024
v EXECUTIVE
SUMMARY
To prepare for safety reviews and pre-licensing interactions for molten salt reactor (MSR) technologies, the U.S. Nuclear Regulatory Commission (NRC) staff is evaluating available information associated with possible future storage and transportation of fuel salt wastes generated from liquid-fueled MSR design concepts. Unlike light-water reactors (LWRs),
liquid-fueled MSRs would use molten salt as the coolant and fuel and would require management of unique waste streams. In addition, fuel salt waste processing may be needed to remove fission products or actinides, recover fissile and fertile material, and stabilize the waste.
Variations of liquid-fueled MSR design concepts involve different fuel characteristics and processing options that lead to a variety of potential waste compositions and forms. A previous report (Adams et al., 2023) described the historical experience with MSR technology, current MSR design and processing concepts, and applicable information from technical reports. This report extends that effort by focusing on a subset of expected waste materials and forms, identifying plausible scenarios for storing and transporting the waste, and identifying potential additional technical considerations and associated information needs for storage and transportation safety evaluations. Specifically, the safety evaluations involving the certification of storage casks and transportation packages within the context of existing regulatory frameworks (NRC, 2020a,b) are addressed.
To evaluate potential technical considerations and information needs associated with regulating aspects of the management of MSR waste, it is first necessary to describe the characteristics of wastes and waste forms that inform this evaluation. Because the possible wastes and waste forms vary depending on the processing options, this report summarizes MSR processing concepts and the characteristics of resulting wastes and waste forms that have been described in technical reports. This report also describes plausible storage and transportation scenarios and evaluates technical considerations and information needs within the context of applicable safety evaluation topics.
While various formulations for MSR fuel salt have been described (McFarlane et al., 2019),
design concepts and fuels are typically grouped by fuels composed of fluoride salts and fuels composed of chloride salts (NAS, 2022; Arm et al., 2020; McFarlane et al., 2019; Riley et al, 2018). Additionally, fluoride salt design concepts that have been described are typically intended to operate in the thermal neutron spectrum while chloride salt design concepts are typically intended to operate in the fast neutron spectrum (Adams et al., 2023). Therefore, these two broad groups of fuels cover a range of possible MSR design concepts. This report focuses on fuel salts and associated waste salts for these two MSR design concepts (fluoride-based thermal; chloride-based fast) as reasonably representative of the liquid-fueled MSR designs that are presently being pursued and that might be pursued in the future.
Treatment and immobilization of MSR salt wastes may be conducted to produce waste forms acceptable for waste management. The potential for different species present in the salts, including halides, alkalis, alkaline earths, rare earths, and actinides that could affect waste management strategies, is being addressed to some degree by ongoing research and analysis.
A combination of possible waste forms being considered are described, including glass waste forms, ceramics or mineral waste forms, ceramic-metal composite (cermet) waste forms, and halide-metal composites (halmet) waste forms (Riley et al., 2018, 2019; Riley, 2020; Arm et al.,
2020; Carlson et al., 2021).
This assessment of potential technical considerations and information needs associated with the certification of storage casks and transportation packages for MSR fuel waste involved
vi describing the characteristics of wastes and waste formsconsidering the possible ways wastes and waste forms could vary depending on the type of reactor, operational characteristics, and processing optionsand identifying plausible storage and transportation scenarios. The approach examined areas of review for key evaluations addressed by NRC Standard Review Plans (SRPs) for the certification of dry storage systems and transportation packages.
The assessment considered whether the area of review was applicable to dry storage systems (DSSs) or transportation packages in the context of the expected characteristics of MSR fuel wastes based on available technical information. Most of the areas of review were found to be applicable to MSR fuel waste, although there is a small subset of review topics related specifically to LWR spent fuel assemblies or related components that are not associated with MSR fuel waste (e.g., fuel assemblies, cladding, fuel pellets, hardware, and details of fuel condition). Many applicable areas of review were found to present no technical challenges (i.e., the existing review approach in the SRP appears valid and adaptable to MSR waste). A subset of areas of review were found to be applicable and present a potential for further technical assessment (e.g., notable differences in MSR fuel salt waste or its management relative to LWR spent nuclear fuel (SNF) may call for different review methods, additional information, or more NRC staff review time).
Regarding structural evaluation of an MSR DSS, the DSS components (e.g., can, container, canister and basket, as applicable) may be unique for an MSR DSS. In the absence of proposed MSR packaging for storage and disposal, MSR fuel waste is assumed to be placed in cans and multiple cans may be placed inside a canister separated by a metal fuel basket. It is also conceivable, that remote handled transuranic (RH-TRU) or similar canisters holding several fuel waste cans are inserted into storage canisters or overpacks. While there may be differences in structural configuration, inertial load, and inherent gaps between components compared to SNF DSS, similar approaches may be applicable to evaluating an MSR DSS and the integrity of the components. Although NUREG-2215 provides guidance for ensuring structural integrity of SNF cladding under drop accidents, the acceptance criteria for cladding would not be directly applicable to MSR cans. Therefore, a potential technical consideration was identified for verifying the structural integrity of MSR cans. No additional technical considerations were identified for evaluating an MSR basket.
The radionuclide inventory of the MSR fuel salt will impact several areas of the technical review in the SRP. Examples include thermal analysis (decay heat); shielding (defining the source);
criticality (material properties, design criteria, model specification, analysis, and burnup credit);
confinement and containment (isotopic inventory); and accident source terms. The radionuclide inventory determination may require specialized physics modeling that accounts for unique dynamics of the MSR reactor design and operation and may also require additional or different data or information (e.g., power rating, burnup, fueling approach, and processing) than a typical LWR SNF system.
MSR fuel waste has unique characteristics relative to LWR SNF. These characteristics, which can vary with potential processing options, include its chemistry, reactivity, potential for gas generation, mass, homogeneity, and structural stability. Potential technical considerations associated with the materials evaluation included the applicability of existing codes and standards to MSR salt waste and the evaluation of environmental degradation, corrosion, and other reactions on cans or other components. These cans or other components may be expected to provide structural and containment or confinement safety functions for the various MSR salt waste forms. Thermal evaluation also accounts for the effect of gas generation from
vii DSS or package contents affecting pressure and temperatures within the initial containment.
The reactivity, pressurization, and potential thermal effects of MSR fuel salt waste may need further assessment. The melting point of unprocessed MSR fuel waste could also present technical questions regarding reconfiguration of contents (criticality, confinement, and containment evaluations), determining accident release fractions, and atmospheric modeling of releases (accident evaluation).
This report documents an initial screening of topics pertaining to the safe storage and transportation of MSR fuel salt waste with the objective of highlighting those that may require further technical assessment. This includes key differences from typical LWR reviews that, for example, may require additional information, a modified approach, or take more time to complete. The information provided aims to further inform the NRC staff and management in evaluating whether additional preparation or guidance may be needed to address the increased industry interest in advanced reactors, including potential applications of MSR technology.
REFERENCES Adams, G., P. LaPlante, and Yi-Ming Pan. Assessment of the Current State of Knowledge on Storage and Transportation of Molten Salt Reactor WasteFinal Report. San Antonio, Texas:
Southwest Research Institute. Center for Nuclear Waste Regulatory Analyses. 2023.
Arm, S.T., D.E. Holcomb, R. Howard, and B. Riley. Status of Fast Spectrum Molten Salt Reactor Waste Management Practice. December 2020.
<https://www.osti.gov/servlets/purl/1761520/> (Accessed April 3, 2023).
Carlson, K., L. Gardner, J. Moon, B. Riley, J. Amoroso, and D. Chidambaram. Molten Salt Reactors and Electrochemical Reprocessing: Synthesis and Chemical Durability of Potential Waste Forms for Metal and Salt Waste Streams. International Materials Reviews. Vol. 66.
pp. 339-363. 2021.
McFarlane, J., P. Taylor, D. Holcomb, and W.P. Poore. Review of Hazards Associated with Molten Salt Reactor Fuel Processing Operations. ORNL/TM-2019/1195. 2019.
<https://info.ornl.gov/sites/publications/Files/Pub126864.pdf> (Accessed 19 September 2022).
NAS. Merits and Viability of Different Nuclear Fuel Cycles and Technology Options and the Waste Aspects of Advanced Nuclear Reactors. Washington, DC: The National Academies Press. 2022. <https://doi.org/10.17226/26500> (Accessed 10 May 2023).
NRC. NUREG-2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities, Final Report. ADAMS Accession No. ML20121A190. Washington DC: U.S. Nuclear Regulatory Commission. 2020a.
NRC. NUREG-2216, Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material, Final Report. ADAMS Accession No. ML20234A651. Washington DC:
U.S. Nuclear Regulatory Commission. 2020b.
Riley, B. Electrochemical Salt Wasteform Development: A Review of Salt Treatment and Immobilization Options. Industrial & Engineering Chemistry Research. Vol. 59.
pp. 9,760-9,774. 2020.
viii Riley, B., J. McFarlane, G. DelCul, J. Vienna, C. Contescu, and C. Forsberg. Molten Salt Reactor Waste and Effluent Management Strategies: A Review. Nuclear Engineering and Design. Vol. 345. pp.94-109. 2019.
Riley, B., J. McFarlane, G. DelCul, J. Vienna, C. Contescu, L. Hay, A. Savino, and H. Adkins.
Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors. NTRD-MSR-2018-000379. PNNL-27723. 2018.
ix ABBREVIATION/ACRONYMS ACU Abilene Christian University AI Aluminum ALARA as low as is reasonably achievable ANSI American National Standards Institute ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials B&PV Boiler and Pressure Vessel (ASME Code)
BWR boiling-water reactor CFR Code of Federal Regulations CNWRA Center for Nuclear Waste Regulatory Analyses CWF ceramic waste form DSF dry storage facility DSS dry storage system DSWF dehalogenated salt waste forms Fe-P-O iron phosphate GBS Glass-bonded sodalite GTCC greater-than-Class-C (waste)
HEPA high-efficiency particulate absorbing HLW high-level radioactive waste ISFSI independent spent fuel storage installation LLRW low-level radioactive waste LWR light-water reactor MSRE Molten-Salt Reactor Experiment MSR Molten Salt Reactor NRC U.S. Nuclear Regulatory Commission
x ORNL Oak Ridge National Laboratory Pu plutonium PWR pressurized-water reactor RH-TRU remote handled transuranic SAR Safety Analysis Report SNF spent nuclear fuel SRP Standard Review Plan SSC structures, systems, and components U
uranium USHYZ ultrastable H-Y zeolite
xi TABLE OF CONTENTS EXECUTIVE
SUMMARY
..................................................................................................ii ABBREVIATION/ACRONYMS.......................................................................................vi TABLE OF CONTENTS................................................................................................viii LIST OF TABLES............................................................................................................ix ACKNOWLEDGMENTS...................................................................................................x 1
INTRODUCTION................................................................................................1-1 2
MOLTEN SALT REACTOR PROCESSING CONCEPTS AND WASTES.......2-1 2.1 Unprocessed Salt Waste....................................................................................2-1 2.2 Stabilized Salt Waste..........................................................................................2-4 2.3 Separated Salt Waste.........................................................................................2-5 3
TECHNICAL CONSIDERATIONS AND INFORMATION NEEDS WITH STORAGE OF MOLTEN SALT REACTOR WASTES......................................3-1 3.1 Storage Scenarios..............................................................................................3-1 3.2 Assessment of Potential Technical Considerations and Information Needs......3-2 3.2.1 Structural Evaluation...............................................................................3-2 3.2.2 Thermal Evaluation.................................................................................3-4 3.2.3 Shielding Evaluation...............................................................................3-6 3.2.4 Criticality Evaluation................................................................................3-7 3.2.5 Materials Evaluation................................................................................3-8 3.2.6 Confinement Evaluation..........................................................................3-9 3.2.7 Accident Analysis Evaluation................................................................3-10 4
TECHNICAL CONSIDERATIONS AND INFORMATION NEEDS WITH TRANSPORTATION OF MOLTEN SALT REACTOR WASTES......................4-1 4.1 Transportation Scenarios....................................................................................4-1 4.2 Assessment of Potential Technical Considerations and Information Needs......4-2 4.2.1 Structural Evaluation...............................................................................4-2 4.2.2 Thermal Evaluation.................................................................................4-4 4.2.3 Shielding Evaluation...............................................................................4-5 4.2.4 Criticality Evaluation................................................................................4-6 4.2.5 Materials Evaluation................................................................................4-8 4.2.6 Containment Evaluation..........................................................................4-9 5
SUMMARY
AND CONCLUSIONS.....................................................................5-1 6
REFERENCES...................................................................................................6-1
xii LIST OF TABLES Page Table 2-1.
Example Fuel and Coolant Chemistries for Reactor Design Concepts Considered in This Assessment.........................................................................2-2 Table 2-2.
Characteristics of MSRE Fuel Salt Waste and Flush Salt Waste in Three Postulated Cans Contained Within a Postulated Transuranic Waste Canister..............................................................................................................2-4
xiii ACKNOWLEDGMENTS This report was prepared to document work performed by the Center for Nuclear Waste Regulatory Analyses (CNWRA) for the U.S. Nuclear Regulatory Commission (NRC) under Contract No. 31310018D0001. The activities reported here were performed on behalf of the NRC Office of Nuclear Regulatory Research. The report is an independent product of CNWRA and does not necessarily reflect the views or regulatory position of the NRC.
The authors thank the NRC staffin particular, Dr. Wendy Reed and the staff from the Office of Nuclear Material Safety and Safeguardsfor constructive comments on a draft version of this report, Osvaldo Pensado for his technical review, and David Pickett for his programmatic review. The authors also thank Arturo Ramos for providing formatting and word processing support in preparation of this document.
QUALITY OF DATA, ANALYSES, AND CODE DEVELOPMENT DATA DATA: There are no original CNWRA-generated data in this report. Sources of other data should be consulted for determining the level of quality of those data.
ANALYSES AND CODES: No codes were used in the analyses contained in this report.
1-1 1
INTRODUCTION To prepare for safety reviews and pre-licensing interactions for Molten Salt Reactor (MSR) technologies, the U.S. Nuclear Regulatory Commission (NRC) staff is evaluating available information to identify potential additional technical considerations and information needs associated with possible future storage and transportation of irradiated fuel wastes generated from liquid-fueled MSR design concepts. Unlike light-water reactors (LWRs), liquid-fueled MSRs would use molten salt as the coolant and fuel and would require management of unique waste streams. In addition, salt processing may be needed to remove fission products or actinides, recover fissile and fertile material, and stabilize the waste. MSR technology was demonstrated in an experimental setting in the 1960s (Haubenreich and Engel, 1970), and several companies are currently pursuing development of new advanced reactors with molten salts in the design (Adams et al., 2023); however, many of the designs are preliminary and conceptual.
Variations of liquid-fueled MSR design concepts involve different fuel characteristics and processing options that lead to a variety of potential waste compositions and forms. A previous report (Adams et al., 2023) describes the historical experience with MSR technology, current MSR design and processing concepts, and applicable information from technical reports. This report extends that effort by focusing on a subset of expected waste materials and potential waste forms, identifying plausible scenarios for storing and transporting the waste, and identifying potential technical considerations and associated information needs for storage and transportation safety evaluations. Specifically, safety evaluations involving the certification of storage casks and transportation packages within the context of existing regulatory frameworks are addressed. This report does not address the adequacy or completeness of these existing regulatory frameworks.
Chapter 2 of this report summarizes the various MSR processing concepts pertaining to waste management, and the characteristics of the resulting wastes and waste forms that have been described in technical reports, which inform the evaluation made in subsequent chapters.
Chapters 3 and 4 describe plausible storage and transportation scenarios, respectively, and evaluate technical considerations and information needs within the context of applicable safety evaluation topics. Chapter 5 summarizes evaluation results and conclusions.
2-1 2
MOLTEN SALT REACTOR PROCESSING CONCEPTS AND WASTES Available technical information evaluated for this report indicates the characteristics of Molten Salt Reactor (MSR) salt waste would vary depending on the reactor design, operational considerations, and processing options that are implemented. Because processing can play a large role in determining the characteristics of the salt waste and waste form, the waste descriptions in this chapter are organized by processing option. Section 2.1 addresses unprocessed salt waste, Section 2.2 stabilized salt waste, and Section 2.3 separated salt waste.
Descriptions of wastes within these sections address important variations based on reactor design and operational considerations.
2.1 Unprocessed Salt Waste The degree of processing that has been described in technical reports varies depending on several characteristics, including reactor design, operational characteristics, and fuel cycle concept. Some MSR design concepts do not include online processing or consider removal for offsite processing or disposal (e.g., Terrestrial Integral Molten Salt Reactor and Abilene Christian University (ACU) Molten Salt Research Reactor). Because future proposals involving storage and transportation of unprocessed salt waste are possible, the characteristics of unprocessed salt waste are described here to inform the evaluation of potential technical considerations and information needs for storage and transportation.
While a variety of formulations for MSR fuel salt have been described (McFarlane et al., 2019),
design concepts and fuels are typically grouped by fuels composed of fluoride salts, and fuels composed of chloride salts (NAS, 2022; Arm et al., 2020; McFarlane et al., 2019; Riley et al.,
2018). Additionally, fluoride salt design concepts that have been described are typically intended to operate in the thermal neutron spectrum, while chloride salt design concepts are typically intended to operate in the fast neutron spectrum (Adams et al., 2023). Therefore, these two broad groups of fuels cover a range of possible MSR design concepts. Focusing on fuel salts and associated waste salts for these two MSR design concepts (fluoride-based thermal and chloride-based fast) is therefore considered reasonably representative of the liquid-fueled MSR designs that are presently being pursued and that could be pursued in the future. Example fuel salt chemistries based on the literature on fluoride-based and chloride-based salts are shown in Table 2-1.
Adams et al. (2023) summarized the experience with on-site storage of unprocessed MSR salt waste at Oak Ridge National Laboratory (ORNL) after the Molten-Salt Reactor Experiment (MSRE) was shut down in 1969. Potential safety issues associated with storage of the fuel salt waste were identified, including criticality safety issues, chemical and radiological issues, and material issues (Sautman, 1995; Peretz, 1996; Atz and Joseph, 2022; Price, 2022). Storage of the solidified fuel salt at the MSRE shows evidence of radioactive material migration that was detected throughout the drain tank piping and the off-gas system. The source of the radiation was volatile UF6 complexes formed via radiolytic decomposition of the solidified fuel salt.
Improper storage of MSR fuel salt could result in accumulation of fissile material in off-gas filters due to UF6 mobilization, which presented a criticality risk that prompted the removal of fissile material from the solidified fuel salt. In addition to criticality safety concerns, high radiation levels and reactive fluorine gas in the piping would also pose serious worker hazards during remediation activities. The presence of moisture and fluorine gas in the off-gas system could
2-2 create the potential for stress corrosion cracking of the off-gas system piping and the charcoal bed vessel made of Type 304 stainless steel. The buildup of corrosion products in the drain line Table 2-1.
Example Fuel and Coolant Chemistries for Reactor Design Concepts Considered in This Assessment Reactor Design Concept Example Reactor Fuel Chemistry Coolant Chemistry Reference Fluoride-based Thermal Spectrum Molten Salt Reactor Experiment 7LiF-BeF2-ZrF4-UF4 (65-29-5-1 mol%,
respectively) 2(7LiF)-BeF2 Riley et al.
2018; Serp et al., 2014 Chloride-based Fast Spectrum Molten Chloride Reactor Experiment PuCl3-NaCl (36-64%,
respectively)
PuCl3-NaCl (36-64%,
respectively)
Latkowski (2021); DOE (2023)
Riley, B., J. McFarlane, G. DelCul, J. Vienna, C. Contescu, L. Hay, A. Savino, and H. Adkins. Identification of Potential Waste Processing and Waste Form Options for Molten Salt Reactors. NTRD-MSR-2018-000379.
PNNL-27723. 2018.
Serp, J., M. Allibert, O. Bene, S. Delpech, O. Feynberg, V. Ghetta, D. Heuer, D. Holcomb, V. Ignatiev, J.L. Kloosterman, L. Luzzi, E. Merle-Lucotte, J. Uhlír, R. Yoshioka, and D. Zhimin. The Molten Salt Reactor (MSR) in Generation IV: Overview and Perspectives. Progress in Nuclear Energy 77 (Supplement C):308-19. 2014.
<https://doi.org/10.1016/j.pnucene.2014.02.014>
Latkowski, J. TerraPowers Molten Chloride Fast Reactor (MCFR). TerraPower.
February 22, 2021.
<https://www.nationalacademies.org/documents/embed/link/LF2255DA3DD1C41C0A42D3BEF0989ACAECE3053 A6A9B/file/DB0D308269688B2BD7B1AF60BAA143D48890C2DE80BB?noSaveAs=1> (Accessed 22 September 2022).
DOE Final Environmental Assessment for the Molten Chloride Reactor Experiment (MCRE) Project, INL/RPT 68976, August 2023.
also caused a blockage in the drain line and postponed the effort to transfer the salt waste from the existing drain line to salt canisters. Additional information related to unprocessed MSR fuel salt waste characteristics, including inventory, dose rates, fissile material content, and thermal power, was examined for this current assessment. Inventory estimates are included in studies by Krall et al. (2022); NAS (2022); Wheeler et al. (2021); EPRI (2021); and Choe et al. (2018).
Choe et al. (2018) provided an analysis of fuel cycle options for the Terrestrial Energy Integral Molten Salt Reactor that included information on generated wastes. EPRI (2021) estimated fuel salt fission product concentrations for a chloride fast reactor concept. NAS (2022) evaluated wastes generated (total annual masses of fission products, minor actinides, uranium, and plutonium) from a Thorium fueled thermal MSR with continuous processing. These provide general (e.g., non-radionuclide-specific activity or mass inventories) or relative or normalized results without itemizing detailed inventories.
Peretz (1996) described the surface dose rate, thermal output, and fissile mass in postulated canisters of MSRE fuel salt waste as part of an evaluation of waste management alternatives.
This information is summarized in Table 2-2. MSRE was a low power (8 MW thermal) experimental MSR reactor that operated for 1.3 equivalent full power years (Haubenreich and Engel, 1970). MSRs with a higher power rating and longer operational period than the MSRE would be expected to generate larger inventories; however, as previously noted, documentation of detailed reactor physics calculations estimating salt waste radionuclide inventories is limited.
Additionally, some of the values in Table 2-2 account for removal of uranium by fluorination and consider decay during decades of storage. Although the MSRE analysis indicated criticality
2-3 metrics were within acceptable limits, this conclusion was applicable for fuel salt waste where the uranium had been removed by fluorination.
2-4 Table 2-2.
Characteristics of MSRE Fuel Salt Waste and Flush Salt1 Waste in Three Postulated Cans Contained Within a Postulated Transuranic Waste Canister Characteristic Fuel Salt Waste Flush Salt Waste Volume of salt in canister (m3) 0.235 0.24 Mass of salt in canister (kg) 581 533 Specific activity (Ci/L)*
13.4 0.22 239Pu equivalent activity (PE-Ci) 36 0.05 239Pu fissile gram equivalents (g)**
95 4
Surface dose rate (R/hr)***
10 0.36 Thermal power (W)****
<14.2
<0.25 Transuranic content (nCi/g) 61,900 1,240 Source: Peretz, F.J. Identification and Evaluation of Alternatives for the Disposition of Fluoride Fuel and Flush Salts from the Molten Salt Reactor Experiment at Oak Ridge National Laboratory, Oak Ridge, Tennessee.
ORNL/ER-380. Oak Ridge, Tennessee: Oak Ridge National Laboratory. 1996.
<https://www.osti.gov/scitech/servlets/purl/441122> (Accessed 9 October 2022).
- Activity assumes decay to 1999 (fuel was discharged in 1969)
- Calculated contact dose at side of salt can with 2 in. steel shielding. Unshielded surface doses are 270 R/hr for the fuel salt and 4.5 R/hr for the flush salt.
- Thermal power not reduced to account for uranium migration or fluorination.
Notz (1985) presented results of ORIGEN, physics-based calculations of the time-dependent fuel salt inventory for MSRE at the time of discharge and at later timesteps, by radionuclide, and cumulatively for all radionuclides. The thermal power of all fuel and flush salts from MSRE at discharge was reported as 5,700 watts. Because the fuel salt controls the inventory and thermal power, if this thermal power at discharge is assumed to correspond to the volume of fuel salt and is scaled to match the volume of the presumed canister volume in Table 2-2, the resulting approximate thermal power of a presumed canister of fuel salt at discharge would be 710 watts and at 10 years after discharge it would be 21 watts (which approaches the <14.2 watts value in Table 2-2). These values fall within maximum heat load values specified for LWR spent nuclear fuel (SNF) storage casks and transportation packages (NRC, 2015; 2014).
More recently, Wheeler et al. (2021) conducted initial source term modeling for molten salt reactors and provided general inventory estimates for a hypothetical 1 GW of power fluoride thermal MSR after one year of operation. They also report the fuel salt volume equal to 1.47 x 105 liters and a full core inventory of 3.02 x 1010 Ci. For comparison, the reported MSRE volume of fuel salt is 1.88 x 103 liters and the inventory of fuel salt at the time of discharge was reported as 1.8 x 106 Ci (Notz, 1985). Therefore, for the hypothetical 1 GW of power MSR reactor Wheeler estimated a fuel salt volume that was 78 times greater than the reported 8 MWp MSRE and a full core fuel salt radioactivity inventory that was over 4 orders of magnitude larger than the reported MSRE fuel salt inventory at discharge.
2.2 Stabilized Salt Waste Although unprocessed fuel salt waste is solid at lower temperatures (below the solidus temperature), its stability can be improved by additional processing. Arm et. al (2020) 1 Flush salt is the salt that was run through the reactor to remove impurities and any oxidation products from the reactor vessel prior to fueling.
2-5 highlighted the lack of a state-of-the-art approach to treatment and stabilization of used fuel salt, although describes several options that exist in concept. Specialty phosphate glass and glass-bonded mineral or ceramic waste forms are described as common across all MSR fuel cycles and salt types being investigated. Some of these waste forms have been experimentally demonstrated, especially for chloride salt. Riley et al. (2018) evaluated a variety of possible waste form options and highlights glass-bonded sodalite for immobilizing mixed chloride-based salt wastes and glass-bonded apatite as a better option for fluoride-based salt wastes. Arm et al. (2020) noted that salt-metal composites have been postulated and have broad applicability but need practical demonstration as waste forms. Leveraging experience with treating waste streams arising from the molten chloride salt in electrorefiners used in pyroprocessing spent metal nuclear fuel is suggested to inform options for stabilizing MSR fuel waste. Stabilization methods and waste form options applicable to electrochemical reprocessing wastes are described in more detail in Section 2.3 because they can be used to stabilize mostly unprocessed fuel salt waste as well as waste streams resulting from separation of constituents from the fuel salt waste.
2.3 Separated Salt Waste Available technologies are being explored for separating various constituents in MSR fuel salt waste for various reasons including reuse of materials, improving or maintaining reactor efficiency, volume reduction of higher hazard materials, and facilitation of waste management.
Adams et al, 2023 summarized electrochemical reprocessing (pyroprocessing) as a technology that allows separation of fissile material from fission products for reuse (e.g., uranium and plutonium). The separated fission products could be incorporated into a ceramic waste form (e.g., sodalite) or other forms and then be disposed. Separation can result in a relatively larger quantity of residual salt (that could possibly be stabilized and disposed as low-level radioactive waste (LLRW) as some authors have suggested).
Many processes are available for treating salt wastes from electrochemical reprocessing. Salt separation techniques include vacuum distillation, oxidative precipitation, reactive precipitation, melt crystallization, dehalogenation, phosphorylation, and zone freezing (Riley, 2020). Salt processing and waste form options for different MSR waste streams are reviewed by Riley et al.
(2018, 2019). Arm et al. (2020) noted the selected equipment, methods, and procedures for separating solids from fuel salts have not been disclosed by developers to date.
Treatment and immobilization of MSR salt wastes to produce waste forms acceptable for waste management will have to be capable of containing all the different species present in the salts, including halides, alkalis, alkaline earths, rare earths, and actinides. A combination of possible waste forms is being considered, including glass waste forms, ceramics or mineral waste forms, ceramic-metal composite (cermet) waste forms, and halide-metal composites (halmet) waste forms (Riley et al., 2018, 2019; Riley, 2020; Arm et al., 2020; Carlson et al., 2021).
Waste form options applicable to immobilization of MSR fuel salts with delineation between fluoride and chloride salts are tabulated in detail in Arm et al. (2020). The glass-bonded sodalite (GBS), ceramic waste form (CWF), and the iron phosphate (Fe-P-O) glass and ultrastable H-Y zeolite (USHYZ) waste forms following dehalogenation of the chloride salts, have been experimentally demonstrated for MSR application, but these waste forms are not demonstrated for fluoride salts. Waste form properties that may be relevant to the safety review regulatory frameworks for storage and transportation include reactivity, durability, resistance to thermal or radiological effects, compatibility with future processing/re-canning if needed. Descriptions of
2-6 the waste form options that have been demonstrated for MSR application with emphasis on relevant properties are provided below.
Glass-bonded sodalite ceramic waste form The GBS-CWF is one of the most well-studied waste form options for immobilizing chloride-based salt wastes and has been demonstrated as a waste form for salt wastes from electrochemical reprocessing of Experimental Breeder Reactor II fuel (Riley et al., 2019). The GBS-CWF is a multi-phase material where alkalis and halides from the salts are immobilized in sodalite, while the remainder fission products such as alkaline earths and rare earths are incorporated into the glass phase as oxides (Priebe and Bateman, 2008; Riley et al., 2019; Riley, 2020). This waste form is fabricated by mixing the salt and zeolite 4A and heating to 500 °C [932 °F] to occlude the salt into the structure of the zeolite. A borosilicate glass binder is added at a loading of 26 mass %. The mixture is then heated to 925 °C [1,697 °F] to convert the zeolite to the final GBS waste form. The waste forms have been made at full-scale (i.e., 400 kg monolith that is 0.679 m diameter x 3.12 m tall), and production of radiological waste forms has also been experimentally demonstrated in a hot cell on a smaller scale (Carlson et al., 2021). Since the sodalite structure can readily incorporate alkalis and halides, glass-bonded apatite is a better option for fluoride-based salt wastes because the apatite structure can incorporate alkalis, alkaline earths, and halides (Riley et al., 2018).
Priebe and Bateman (2008) measured chemical durability of the GBS-CWF using the standard product consistency test, according to American Society for Testing and Materials (ASTM)
C1308 Test Method A. The test results revealed that the release rates of all the matrix elements (i.e., silicon, aluminum, boron, and potassium) are at least an order of magnitude lower than a comparable environmental assessment glass that is used as the baseline for repository durability analyses. The performance of accelerated leach tests suggests that the GBS-CWF is a suitable option for immobilizing unprocessed chloride-based salt wastes.
Iron phosphate glass waste form The iron phosphate (Fe-P-O) glass waste form has been developed to immobilize salt wastes from the electrochemical reprocessing of spent nuclear fuel (Carlson et al., 2021; Riley et al.,
2021). The process for producing the waste form with electrochemical salt wastes involves the reaction of chloride salts with the phosphate precursors at moderate temperatures less than 600 °C [1,112 °F] that drives off the halides from the salt as a volatile product and produces a Fe-P-O glass. The Fe-P-O glass is not chemically durable, and iron oxide is added and vitrified at higher temperatures less than 1,200 °C [2,192 °F] to generate a glass waste form with acceptable chemical durability. This process results in dehalogenation of the waste salt, capture of the gaseous halide-bearing species, and immobilization of the residual salt components in the Fe-P-O glass waste form.
Chemical durability studies of the Fe-P-O glass waste form have demonstrated its performance as a matrix for radioactive salt cations (Ebert and Fortner, 2019; Carlson et al., 2021).
Accelerated leach tests were conducted using a modified ASTM C1308 method with four sequential complete solution exchanges made at 24-h intervals. The test results revealed that Fe-P-O glasses exhibit much lower dissolution rates compared to the GBS-CWF. The Fe-P-O glasses have been demonstrated as waste forms for immobilizing chloride-based salt wastes and may potentially be suitable for the treatment and immobilization of fluoride-based salt wastes.
2-7 Ultrastable H-Y zeolite waste form A dechlorination process that employs the ion exchange properties of microporous zeolite frameworks has been developed for immobilizing salt wastes from electrochemical reprocessing (Wasnik et al., 2019; Gardner et al., 2020). Development of this two-step approach process is driven due to the lack of high-salt-loaded waste forms available for salt immobilization. This process is referred to as the ultrastable H-Y zeolite (USHYZ) process. The first step in waste processing involves the removal of chlorine as hydrogen chloride gas via ion exchange reactions between USHYZ and halide salts. In the second step, the salt cation-loaded zeolite is consolidated to produce a dense, monolithic assemblage of silicate-based phases for storage or final disposal. The dehalogenation and consolidation processing steps result in a substantial reduction of 35% in required storage volume for waste salt. Wasnik et al. (2019) evaluated several processing variables to improve the degree of ion exchange and achieve dechlorination efficiencies of up to 95%.
Gardner et al. (2020) performed chemical durability experiments using modified ASTM C1308 method in deionized water and different silica solutions. A comparison of release rates for the GBS-CWF and the USHYZ-derived dehalogenated salt waste forms (DSWF) showed that the cumulative releases for the DSWF are less than 33% of those measured for the GBS-CWF.
Unlike the USHYZ process for dechlorination of chloride salt wastes, defluorination using the USHYZ process has not yet been demonstrated with MSR-relevant fluoride salt mixtures.
A comparison of the physical properties for the GBS, Fe-P-O glass, and USHYZ waste forms is provided by Riley (2020). It should be noted that little information is available on waste form properties relevant to the safety reviews for storage and transportation of MSR fuel-related wastes, such as resistance to thermal or radiological effects. Therefore, further investigation on waste form options and their properties for MSR salt wastes is warranted.
3-1 3
TECHNICAL CONSIDERATIONS AND INFORMATION NEEDS WITH STORAGE OF MOLTEN SALT REACTOR WASTES Cask systems used for the storage of spent nuclear fuel are certified by the U.S. Nuclear Regulatory Commission (NRC) in accordance with the regulations in Title 10 Code of Federal Regulations (CFR) Part 72. NUREG-2215 (NRC, 2020a) provides guidance to the NRC staff for reviewing Safety Analysis Reports (SARs) for a Certificate of Compliance for a dry storage system (DSS). To help the NRC staff prepare for the possibility of certifying storage casks for MSR fuel salt waste, an assessment of the topics addressed in the standard review plan (i.e., areas of review) was conducted considering the potential differences in the characteristics of MSR fuel salt waste relative to typical light-water reactor (LWR) spent fuel to identify potential technical considerations and associated information needs. This assessment is an initial screening of topics with the objective of highlighting those that may need additional consideration. The assessment considered key differences from typical LWR reviews that, for example, may require additional information, a modified approach, or take more time to complete. Increased understanding of technical aspects pertaining to MSR spent fuel storage and associated wastes is expected to further inform the NRC staff in evaluating whether additional preparation or guidance may be needed.
The assessment approach considers the descriptions of MSR fuel salt wastes by processing option that are documented in Chapter 2 of this report and develops scenarios for how these wastes could be stored. Consistent with the focus on storage cask certification, the storage scenarios provide general information and assumptions about what an MSR fuel salt waste dry storage system (DSS) might look like to allow the applicable review areas to be assessed for potential technical considerations and information needs. The scenarios considered in this assessment appear plausible and are consistent with available information but should not be considered as an endorsement of any technology or approach.
3.1 Storage Scenarios In the evaluation of Molten-Salt Reactor Experiment (MSRE) fuel salt waste disposition alternatives, Oak Ridge National Laboratory (ORNL) (Peretz, 1996) described a packaging concept with general design information related to MSR fuel salt waste transportation which is also broadly applicable to storage. To allow for possible additional processing, handling, and transportation, ORNL proposed filling an inner Hastelloy-N can with molten salt and allowing it to cool and solidify. The can would be placed inside a shielded container, and three shielded waste containers could then be placed inside a remote handled transuranic (RH-TRU) canister.
It is noted that the ORNL transportation study, which presumed an RH-TRU 72-B transportation package could be used for transportation of the loaded RH-TRU canisters, did not conduct a detailed evaluation or address whether the proposed approach would meet package specifications. The storage scenario for the assessment in this report assumes a similar can and canister system could be used for both storage and transportation.
Waste forms resulting from all three MSR fuel waste processing options (unprocessed, stabilized, separated) described in Chapter 2 are assumed to be placed in or solidified in a can or similar component that provides, for example, structural support for handling and an initial containment or confinement barrier; therefore, the conceptual canning approach described by ORNL is assumed to be broadly applicable to wastes associated with the three processing options. Additional details are provided as follows.
3-2 Peretz (1996) provided general design information that was considered by ORNL. The salt waste Hastelloy-N can would have 1.27 cm [0.5 in] thick walls. This thickness was thought to provide an adequate corrosion allowance should uranium be separated from the salt by fluorination in the can. The can would then be inserted into a shielded container with the approximate outer dimensions of a 55-gallon drum. This container would provide 5.1 cm
[2.0 in] of steel in each direction as a radiation shield. ORNL expected three of these shielded salt waste containers could be placed into an RH-TRU canister. An RH-TRU canister is made of 0.64 cm [0.25-in] thick carbon steel and is 307 cm [121 in] long and 66 cm [26 in] in diameter.
For a fluoride salt containing the gamma source from cesium or the 232U daughter chain, ORNL noted the evolution of radiolytic fluorine must be controlled (Peretz, 1996). ORNL proposed a getter, such as soda-lime (a mixture of sodium hydroxide and calcium oxide), to control fluorine gas. To facilitate a getter system, each salt can would be filled with fuel salt waste to 75% of its internal volume. The remaining internal volume would be used to vent the fluorine gas through the getter and a high-efficiency particulate absorbing (HEPA) filter. ORNL noted the quantity of fluorine gas could be calculated and verified, thereby allowing the amount of getter to be determined. ORNL indicated additional effort was needed to determine whether such a transportation package could be certified, and similar considerations also would apply to certifying a DSS.
The solidification of molten radioactive waste in steel canisters has been demonstrated in other programs. U.S. Department of Energy (DOE) has longstanding experience vitrifying high-level radioactive waste (HLW) for storage in stainless steel canisters in preparation for eventual disposal (NWTRB, 2017). HLW is vitrified by mixing it with silica sand and other glass-forming chemicals, heating the mixture to very high temperatures [approximately 1,150°C (2,100°F)]
until it melts, and pouring the molten material into stainless steel canisters where it cools to form a glass. Some glass waste forms that are being evaluated for processed MSR waste streams are described in Sections 2.2 and 2.3.
Storage of MSR fuel waste could occur onsite or offsite. Additionally, the timeframe for storage would depend on the waste management approach, including whether waste processing is planned, and if so, if it will occur onsite or offsite.
3.2 Assessment of Potential Technical Considerations and Information Needs The following sections describe the results of an assessment to determine the technical considerations and potential information needs regarding certification reviews of storage casks for MSR fuel waste. The assessment was conducted for review areas documented in the Standard Review Plan (SRP) for storage cask certification reviews in NUREG-2215 (NRC, 2020a). The assessment determined that the review plans for all areas of review are broadly applicable to cask certifications for a wide variety of contents [e.g., LWR SNF and greater-than-Class-C (GTCC) wastes], including MSR fuel waste. The following sections highlight a subset of narrower topics within review areas where possible differences in the MSR fuel waste relative to LWR spent fuel may present additional technical considerations and information needs.
3.2.1 Structural Evaluation The structural review of a DSS and dry storage facility (DSF) in Chapter 4 of the SRP (NUREG-2215) (NRC, 2020a) evaluates the structural integrity of structures, systems, and components (SSCs) important to safety and other SSCs. The SSCs important to safety for
3-3 storage systems are those that are relied on to provide confinement, subcriticality, radiation shielding, support, and retrievability safety functions under normal, off-normal (anticipated occurrences), and accident conditions including natural phenomena effects. The structural evaluation is coordinated with review of information in other chapters of the SRP, namely, thermal, shielding, material, criticality, confinement, and accident analysis. This section focusses on the areas of NUREG-2215 with any expected important differences associated with the various MSR wastes relative to typical wastes [e.g., LWR SNF or low-level radioactive waste (LLRW)/GTCC] that could affect the safety review.
A DSS for spent fuel typically uses a canister within an overpack as components that store SNF in a dry environment. A DSS provides confinement, radiological shielding, sub-criticality control, structural support, and passive cooling of SNF during normal, off-normal, and accident conditions. A similar storage system is envisioned for MSR salt storage.
Characterization of MSR waste, and description and design of disposal packages for storage of MSR salt is currently not available (Riley, 2018). However, it is known that the MSR waste material would be either untreated solid salt from the reactor, or in some other processed forms (e.g., glass, glass-ceramic, or ceramic). Adams et al. (2023) discussed the concept of a packaging system for transportation of ORNL MSR fuel salt waste proposed by Peretz (1996) that is also described in Section 3.1 of this report. In absence of a proposed MSR packaging for storage and disposal, we assume that the MSR waste either as untreated or in processed form would be in cans and multiple cans may be placed inside a canister separated by a metal fuel basket. It is also conceivable, that RH-TRU or similar canisters holding several fuel waste cans are inserted into a storage canister or overpack. Typically, at independent spent fuel storage installation (ISFSI) sites, the DSF consists of concrete pads supported on a firm foundation and the DSS is placed either horizontally or vertically on the concrete pad as a free-standing structure.
For evaluation of structural design and integrity of SSCs important to safety for an MSR DSS, the areas of review include: (a) confinement container or canister, including shell, lid and associated bolts or welds, and the internals, including fuel basket and MSR cans instead of fuel rods and cladding; (b) storage overpack; and (c) concrete storage pads (if classified as important to safety). Other SSCs associated with transfer and handling operations of the DSS used at the DSF may be part of the structural review subject to NRC approval. The review focus on each SSC important to safety and other SSCs (if approved by NRC) is broadly categorized in the SRP as (i) description, (ii) design criteria, (iii) loads associated with normal, off-normal and accident conditions, (iv) analytical approaches, (v) normal and off-normal conditions, and (vi) accident conditions. The MSR DSS design is envisioned to include a certified canister and overpack design with potentially a fuel basket configured to the dimensions and number of MSR cans to be inserted in the canister. Currently, there are no certificates of compliance for an MSR fuel waste DSS.
To assess any expected differences associated with the various MSR wastes (discussed in Section 2.0) relative to typical wastes (e.g., LWR SNF or LLRW/GTCC) that could affect the safety review, technical requirements under all the subsections associated with SRP Section 4.5 (Review Procedure) were applied to all DSS and DSF SSCs. The evaluation resulted in focusing on the structural design of the fuel basket, MSR can, and MSR container, or canister.
The storage casks and transfer casks that are certified for spent nuclear fuel require evaluation and demonstration of suitability for handling and storing MSR waste for a site-specific or general license. The NRC staff has significant experience in the review of technical information,
3-4 justifications, and analyses to verify that the structural design meets the acceptance criteria for certification and subsequent site-specific evaluations of the DSS under normal, off normal, and accident conditions in accordance with the SRP.
A DSF involves SSCs providing functional requirements associated with handling and transfer operations, storage, and site infrastructure designed to mitigate external hazards. DSF for an ISFSI site also includes concrete storage pads where storage overpacks are located either free standing or anchored. The storage pads, in accordance with the SRP, may or may not be important to safety. Staff experience exists for site-specific review of the DSF facilities at storage sites in accordance with the SRP. Unique technical considerations are not anticipated to be associated with the review of a DSF because facility design and operations for storage of MSR waste are expected to be similar to design and operations for LWR spent fuel storage casks.
MSR can, fuel basket, and container or canister The guidance in NUREG-2215 has not been applied to a review of an MSR DSS or its components. If the SSCs of an MSR DSS have important to safety functions that are covered in the SRP, then the SRP should be applicable to the review of this system. Currently there are no proposed DSS designs for dry storage of MSR waste. The configuration and design of the fuel basket will be dependent on the dimensions, shape, and number of MSR waste cans inserted in a canister, taking into consideration thermal and criticality control.
The MSR cans are likely to be placed freely within the basket slot inside the container or canister cavity, similar to the placement of SNF LWR fuel assemblies for typical spent fuel waste storage. Because of the several contacts and interfaces between various components in the MSR DSS, structural analyses considering material nonlinearities are expected to be used to calculate the impact between the components under accidental drop and tipover conditions to evaluate the structural integrity of the MSR DSS. NRC staff has significant experience reviewing analytical calculations and structural simulations of storage casks under accident conditions, considering the SNF assemblies and basket. While there may be differences in structural configuration, inertial load, and inherent gaps between components compared to an SNF DSS, similar modeling approaches are applicable to evaluating an MSR DSS and the integrity of its components. A potential technical consideration was identified for verifying the structural integrity of MSR cans because MSR fuels (for fluid-fueled reactor types) do not have cladding. If the MSR cans are assumed to be designed, fabricated, and tested in accordance with applicable divisions and sections of ASME Boiler and Pressure Vessel (B&PV) code, the guidance, and acceptance criteria in NUREG-2215 would be applicable to review of the can structural integrity. It is also possible that applicants may propose and demonstrate alternative criteria. In such instances, a more detailed review and evaluation by NRC staff would determine whether the applicants technical bases and information support the evaluation of structural integrity of MSR cans including their closure system.
3.2.2 Thermal Evaluation As documented in the SRP for storage cask evaluations (NRC, 2020a), the thermal evaluation addresses heat transfer and flow characteristics and decay heat removal systems of a DSS.
This evaluation ensures that the storage container and fuel material temperatures of a DSS remain within the allowable limits for normal, off-normal, and accident conditions. The review, when applied to storage of LWR fuel, is intended to confirm that the temperatures of the fuel cladding as a fission product barrier will be maintained throughout the storage period to protect
3-5 the cladding against degradation that could lead to gross rupture. The review also confirms that the applicant uses acceptable analytical and testing methods, as applicable, as part of the evaluation of the DSS thermal design. The review of the decay heat removal system verifies that the system operates to limit the temperatures of materials used for SSCs ITS, and solidified HLW containers within the allowable limits under normal, off-normal, and accident conditions.
The NRC staff also evaluates the wet and dry fuel assembly loading and canister transfer systems for adequate decay heat removal under normal, off-normal, and accident conditions.
Specific areas of review include the decay heat removal system, material and design limits, thermal loads and environmental conditions, and analytical methods, models, and calculations, and surveillance requirements. The analytical methods, models, and calculations area of review evaluates configuration, material properties, boundary conditions, computer codes, temperature calculations, pressure analysis and confirmatory analysis.
The methods applied to evaluations of heat transfer, flow characteristics, and decay heat removal systems would be typical of current cask systems once the heat source has been adequately defined. Some details of the guidance that are not directly applicable to MSR fuel waste include references to LWR fuel assemblies and characteristics that are only applicable to LWR fuel (e.g., cladding, fuel pellets, and wet loading). Therefore, thermal analyses relating cladding and fuel pellets to heat transfer or that are needed to demonstrate LWR fuel assemblies remain within design limits would not directly apply. Similar analyses may be needed for MSR fuel waste containers (e.g., cans) or salt fuel waste forms if relied upon for safety. Applicable MSR fuel waste form material characteristics would be expected to be provided by the applicant and could potentially be described in guidance, although the present range of options may favor case by case review. Additionally, some processed MSR fuel salt waste streams would have different compositions and therefore would vary regarding decay heat, potentially involving separate or bounding decay heat calculations. Note that the design decay heat information would be calculated by the vendor for the container.
The potential need for further assessment of thermal evaluation methods was identified for a subset of review areas applicable to the analytical methods, models, and calculations area of review regarding boundary conditions, pressure analysis, and confirmatory analysis Regarding pressure analysis, typical fission product gasses identified in the SRP such as krypton, xenon, and tritium are expected to be removed during MSR operations (Riley et al.,
2018); however, other reactive gases may apply to fuel salt waste depending on reactor design and operation (e.g., fluorine and uranium hexafluoride; Section 2.1). Reactive gases that are different than gases in LWR spent fuel may involve different considerations including additional design or components (e.g., use of getters) and would have different properties and potentially different thermal characteristics and effects (Section 3.1). NUREG-2216 SRP (Section 4.4.2.3) indicates that no credit should be taken for getters as it relates to hydrogen generation; it is unknown as this point how getters for fluorine gas generation, for example, would be addressed.
Areas related to the decay heat removal system, material and design limits, thermal loads and environmental conditions, analytical methods, models, and calculations (configuration, material properties, computer codes, temperature calculations) and surveillance requirements are thought to be broadly applicable to MSR waste. It is considered that the methods applied to evaluations of heat transfer, flow characteristics, and decay heat removal systems would be typical of current cask systems once the heat source has been adequately defined.
Further analysis of possible waste forms and their characteristics, including thermal characteristics, would help to reduce uncertainties about likely options that may be pursued in
3-6 future applications. Improved characterization of MSR fuel waste radionuclide inventories for various design and operation options would help to bound the range of possibilities and facilitate comparisons with LWR spent fuel and other wastes that can inform preparations for safety reviews.
3.2.3 Shielding Evaluation As documented in the SRP for storage cask evaluations (NRC, 2020a), the shielding evaluation ensures shielding design features provide adequate protection against direct radiation from the DSS contents. The shielding features should limit the direct radiation dose to the operating staff and members of the public so that the total dose (i.e., due to direct radiation and any effluents or releases) remains within regulatory requirements during design-basis normal operating, off-normal, and accident conditions. The review ensures that the shielding design is adequately defined and evaluated to support the evaluation of compliance against dose limits for direct radiation from an independent spent fuel storage installation (ISFSI) in 10 CFR 72.236(d),
10 CFR 72.104 and 10 CFR 72.106. The review also ensures adequate consideration of as low as reasonably achievable (ALARA) in the DSS design and operations. The areas of review for the shielding evaluation include (i) the shielding design description; (ii) radiation source definition (initial enrichment, computer codes for radiation source definition, gamma sources, neutron sources, other parameters affecting the source term); (iii) shielding model specification (configuration of shielding and source, material properties); and (iv) shielding analyses (computer codes, flux-to-dose-rate conversion, dose rates, confirmatory analyses).
The methods applied to evaluations of shielding design, radiation source definition, shielding models and analyses would be typical of a current DSS except for determination of the inventory of its contents. Also, some processed (e.g., separated, or concentrated) MSR fuel salt waste streams would have different compositions than the fuel salt waste that was removed from the reactor and therefore processing methods and the characteristics of resulting waste streams and forms may be more important than typical LWR characteristics such as initial enrichment, cooling time, and burnup to define the radiation source for processed waste streams.
The radionuclide inventory of MSR fuel waste would impact radiation source definition; shielding model specification (material properties); and shielding analyses (applicant and confirmatory analyses) and would be potentially unique to MSR-specific reactor design and operational considerations that affect inventory (e.g., power rating, burnup, fueling approach, and processing). Specialized reactor physics modeling may be needed to determine or bound the inventory for a specific design and operational approach. Radionuclide inventory in the salt waste would also affect shielding calculations. While detailed radionuclide inventory estimates are available for MSRE waste (Section 2.1) and capabilities exist to conduct such calculations, comparable detailed estimates of MSR fuel salt inventories for more recent designs or design concepts were not available for reactors with higher power ratings and longer expected operational periods. As described in Section 2.1, for context, the dose rate from a postulated shielded can of MSRE fuel salt waste was estimated to be 10 R/hr (which increased to 270 R/hr for an unshielded can). These dose rates are less than dose rates associated with some HLW and greater-than-Class-C (GTCC) wastes. Additionally, if the MSR fuel salt waste for specific designs or operational approaches were found to have an unusually high radionuclide inventory contributing to high dose rates, mitigation strategies are available, such as increasing the cooling time or diluting the waste with compatible clean materials to reduce the inventory of radioisotopes.
3-7 For other areas of review, namely related to the shielding design description; shielding model specification (configuration of shielding and source, material properties other than inventory);
and shielding analyses (computer codes, flux-to-dose-rate conversion, dose rates), the approach applied to these evaluations would be typical of a current DSS once the inventory for the MSR fuel salt has been adequately defined. Regarding configuration of the source, a potential theoretical concern would be the potential for radioactive materials in liquid fuel salt waste to segregate during cooling, as the phase changes from liquid to solid. ORNL previously evaluated the potential for segregation of uranium in the context of criticality (Sections 3.2.4 and 4.2.4), however, did not report similar studies for other elements. This could affect assumptions of source homogeneity in shielding calculations and therefore may warrant further consideration in preparations for future DSS certification reviews.
Further analysis of possible waste forms and their characteristics, including radiological characteristics, would help to reduce uncertainties about likely options that may be pursued in future applications. Improved understanding of MSR fuel waste radionuclide inventories for various design and operation options would help to bound the range of possibilities and facilitate comparisons with LWR spent fuel and other wastes that can inform preparations for safety reviews.
3.2.4 Criticality Evaluation As documented in the SRP for storage cask evaluations (NRC, 2020a), the criticality evaluation ensures SNF proposed to be placed into dry storage under 10 CFR Part 72 remains subcritical under normal, off-normal, and accident conditions involving handling, packaging, transfer, and storage. The areas of review for the criticality evaluation include the criticality design criteria and features; fuel specification (fuel type, non-fuel hardware, fuel condition); model specification (configuration, material properties); criticality analysis (computer codes and cross section data, neutron multiplication factor, benchmark comparisons); and burnup credit (limits for the licensing basis, licensing-basis model assumptions, code validationisotopic depletion, code validation keff determination, loading curve and burnup verification).
It was determined that the criticality areas of review in the SRP include some aspects that are only applicable to reviews of LWR fuel storage. The concepts and methods applied to these criticality evaluations would be typical except for determining the inventory of the contents and some implementation details. Areas of review that were found to be not directly applicable to MSR fuel waste includes specifications for non-fuel hardware and fuel condition (both are LWR fuel assembly-specific) and consideration of the loading curve and burnup verification (also based on the potential for variation in burnup among fuel assemblies that might be stored). Also, some processed (e.g., separated, or concentrated) MSR fuel salt waste streams would have different composition than the fuel salt waste that was removed from the reactor and therefore may not contain fissile material and, therefore, the criticality review would not apply.
The potential need for further assessment of criticality evaluation methods was identified for all directly applicable review areas including criticality design criteria and features; fuel specification (fuel type); model specification (configuration, material properties); criticality analysis (computer codes and cross section data, neutron multiplication factor, benchmark comparisons); and burnup credit (limits for the licensing basis, licensing-basis model assumptions, code validationisotopic depletion, code validationkeff determination). Most of these review areas are affected by the fuel salt waste inventory which has unique technical considerations, previously described in Section 3.2.3. These review areas also include technical bases and references to existing data that are specific to pressurized-water reactor (PWR) SNF.
3-8 While criticality analyses are available for MSRE waste (Section 2.1), comparable detailed analyses of MSR fuel salt waste for more recent designs or design concepts were not available for reactors with higher power ratings and longer than experimental operational periods.
Variation in possible reactor designs and operational characteristics suggest variation in fissile material inventories. It is noteworthy that the SRP explains recommendations relating to burnup credit are limited to PWR SNF because a technical basis for burnup credit applicable to boiling-water reactor (BWR) SNF had not been developed. Along these lines, it may be reasonable to consider the potential need to develop a technical basis for accepting a burnup credit approach for MSR fuel waste.
Additionally, for MSR fuel salt wastes with higher fissile material content that may contribute to criticality concerns, potential mitigation strategies that might be considered or developed include volume reduction in cans, processing to remove fissile material, adding neutron poisons to cans, incorporating neutron poisons into the salt matrix or waste form, or diluting the waste with compatible clean materials to reduce the inventory of fissile material per can.
A related potential theoretical concern would be the potential for uranium in liquid fuel salt waste to segregate during cooling, as the phase changes from liquid to solid. For the MSRE, ORNL reported the results of an experiment where simulated fuel salt was solidified at rates that would approximate the cooling rate in the reactor drain tanks after reactor shutdown (Peretz, 1996).
The objective of this experiment was to determine whether uranium segregation could take place as crystalline phases began to form, but most of the salt was still liquid. Uranium samples taken from the solidified salt plug ranged from 2.94 weight-percent near the top to 5.74 weight-percent at the bottom. The degree of segregation was considered by ORNL to be small and not affect criticality safety in the drain tanks. This may not necessarily be the case for future fuel salt, especially considering that some MSR designs are considering uranium fuel at much higher enrichment (~20 percent) and a criticality analysis would likely be needed.
Further analysis of possible waste forms and their physical characteristics relevant to criticality safety (e.g., structural integrity and neutronic poisons) would help to reduce uncertainties about likely options that may be pursued in future applications. A more detailed evaluation of the technical bases and information in the SRP that are limited to PWR SNF may help to clarify whether additional guidance is needed to support reviews of MSR fuel waste. Improved characterization of MSR fuel waste radionuclide inventories for various design and operation options would help to bound the range of possibilities and facilitate comparisons with LWR spent fuel and other wastes that can inform preparations for safety reviews.
3.2.5 Materials Evaluation As described in NUREG-2215 (NRC, 2020a), the materials evaluation addresses materials performance of the SSCs used for dry storage. The materials review ensures that the materials performance of storage system SSCs must be adequate under all credible loads and environments for normal off-normal, and accident conditions, and will not pose operational problems with respect to its removal from storage under normal and off-normal conditions.
Specific areas of review include the system and facility design (drawings, codes and standards, weld design, inspection, and testing), material properties (mechanical properties of metals, thermal properties, radiation shielding materials, criticality control materials, concrete and reinforced steel, bolt applications, seals), environmental degradation and corrosion and other reactions (corrosion resistance, protective coatings, content reactions, management of aging degradation) and fuel cladding integrity (fuel classification, uncanned spent fuel, canned spent fuel).
3-9 The areas of review address fundamental aspects of DSS materials evaluation that are adaptable to variations in proposed DSS contents. Further assessment may need to be considered regarding the application of codes and standards for MSR fuel waste storage system design and construction. The review method in NUREG-2215 calls for verification of the codes and standards for storage container components important to safety. The applicability of codes and standards for storage system design and fabrication with new materials will likely need to be evaluated. In the absence of specific codes and standards, the design and fabrication details for MSR fuel waste storage systems, including the technical basis for the adequacy of existing codes and standards, will likely need to be provided by the applicant. NRC technical readiness plans for evaluating non-LWR designs describe the NRCs intention to apply the established process for incorporating codes and standards into the regulatory framework (NRC, 2017).
For the review of the environmental degradation and corrosion and other chemical degradation processes caused by reactions among the contents and between the contents and the storage container components, the review method in NUREG-2215 calls to examine whether corrosion wastage could lead to a loss of intended functions. The current review method is focused on corrosion of materials such as hardware components of stainless steel and zirconium alloy-clad UO2 fuels. Therefore, the chemical interactions of air and water with MSR fuel waste or associated cans or other components that are relied upon for structural support and containment may need to be evaluated. For the review of the fuel cladding integrity regarding uncanned spent fuel, the safety analyses of uncanned spent fuel rely on the integrity of the fuel cladding to maintain the analyzed configuration. The current review method is limited to guidance on the mechanical properties of zirconium alloy cladding. The performance of MSR fuel waste under storage environments may need evaluation.
The general acceptance criteria in NUREG-2215 (NRC, 2020a) pertaining to materials performance of storage system SSCs were determined to be not impacted by specifics of the technologies for MSR fuel waste storage.
Further analysis of possible waste forms and their performance would help to reduce uncertainties about the likely options that may be pursued in future applications. A more detailed evaluation of the technical bases and information in the SRP that are limited to LWR SNF may help to clarify whether additional guidance is needed to support reviews of MSR fuel waste.
Improved characterization of the chemical interactions between MSR fuel waste and the storage container components under storage environments would help to bound the range of possibilities and facilitate comparisons with LWR SNF and other wastes that can inform preparations for safety reviews.
3.2.6 Confinement Evaluation For confinement review, the SRP (NUREG-2215) (NRC, 2020a) provides guidance to evaluate whether the confinement features of the DSS and DSF are designed to ensure that radiological releases to the environment would be within the limits established by the regulations and that the waste material will be sufficiently protected against degradation. The review areas include confinement design characteristics; confinement monitoring capability; nuclides with potential for release; confinement analyses addressing normal, off-normal, and accident conditions; identification of release events; and evaluation of release estimates.
The applicant should provide information on confinement boundary including confinement vessel, welding or bolt closure, penetrations, valves, seals, and closure devices, material, leak
3-10 testing etc. Confinement boundary failure under normal, off-normal, or accident conditions is not acceptable, hence structural integrity during design-basis conditions is confirmed by the structural analysis. Guidance in NUREG-2215 addresses confinement capabilities and review of confinement design under normal, off-normal, and accident conditions.
For review of radionuclides for potential release, NUREG-2215 (NRC, 2020a) provides guidance to determine radionuclide inventory available for release from SNF considering initial enrichment, burn-up and cool time including verifying that the analysis of total activity is based on storage container design loading. The SRP provides guidance in detail to review fractions of radioactive material available for release and fractions that are not releasable to the environment under credible normal, off-normal, and accident conditions accounting for radionuclides trapped in the fuel matrix and radionuclides that exist in a chemical or physical form.
NUREG-2215 provides guidance on confinement analysis and resulting dose for normal, off-normal, and accident conditions. The review includes verification that calculation of the specific activity (activity per unit of volume) for each radioactive isotope is based on rod breakage fractions, release fractions, isotopic inventory, and cavity free volume. The NRC staff also evaluates input parameters, adjusted maximum confinement boundary leakage rates under normal, off-normal, and accident conditions, calculation of isotope specific leak rates (curies per second) and determination of doses for inhalation and immersion exposures at the controlled area boundary considering atmospheric dispersion.
Differences in the properties and radionuclide inventory of the MSR fuel waste material and its response to thermal conditions (e.g., melting and crystallization/solidification) may require different considerations, data, or information related to estimating release fractions in typical LWR SNF evaluations.
3.2.7 Accident Analysis Evaluation The SRP (NUREG-2215; NRC, 2020a) provides guidance to review an applicants identification of off-normal and accident conditions including site-specific natural hazards that may directly or indirectly affect SSCs important to safety. NUREG-2215 includes a minimum list of the off-normal and accident conditions that the applicant must evaluate; however, other hazards that are unique to facility site and design should be considered by the applicant. The focus of this section is on the off-normal conditions and accidents impacting storage of MSR DSS considering any expected important differences associated with the various MSR wastes relative to typical wastes (e.g., LWR SNF or LLRW/GTCC) that could affect the safety review.
The SRP stipulates a minimum list of both the off-normal events and the accident conditions that the applicant must consider in the Safety Analysis Report (SAR). In the case of accident events and conditions, this list includes storage container tipover, storage container drop, flood, fire and explosion, lightning, earthquake, loss of shielding, adiabatic heat up, tornadoes and missiles generated by natural phenomena, accidents at nearby sites, building structural failure and collapse onto SSCs. The accident analysis section evaluates information on the operating environment and the physical parameters of the accident, the analysis methodology, and the actual analysis performed. The evaluation is coordinated with other chapters (e.g., confirming that the structural design bases of SSCs include loading from all expected off-normal conditions, accidents, or natural hazards). The evaluation also verifies that the structural integrity, thermal performance, shielding capability, and criticality safety of the DSS are not impaired by off-normal and accident events.
3-11 NUREG-2215 provides guidance to review a hypothetical cask drop and non-mechanistic tipover of the cask. Cask handling and transfer operations can result in a vertical end drop, horizontal side drop or tipover. Potential for cask tipover also exists during natural hazards, specifically earthquake, tornado wind and missile, and flood. Cask drop and tipover should be analyzed in order to address the potential impact on the structural integrity of all components of the DSS. The review ensures that the applicants analysis includes evaluation of stress intensity level and buckling of each component of storage container, deformation of container internal members, and deformation and damage to waste material. Cask drop and tipover may also impact shielding capability and criticality safety of the storage container. Additional consideration is discussed in Section 3.2.1 regarding structural analyses for drop and tipover events of storage container with MSR salt waste.
The applicant is required to evaluate an earthquake accident using the design ground motion of the site. NUREG-2215 (NRC, 2020a) provides the approach to review the accident analysis for the DSS in accordance with codes and standards including NRC guidance. The analysis should confirm no excessive sliding or overturning of DSS during shaking, excluding the possibility of impact between casks triggered by the design earthquake. The applicant is also required to evaluate the effects of site-specific tornado and tornado generated missile accident on a DSS precluding translation and overturning due to drag force resulting from high wind and ensuring that shielding, criticality safety, and heat removal capability is not adversely affected from potential physical damage from the tornado missile impact. NUREG-2215 provides guidance to review the effects of potential flooding at the storage site requiring demonstration of stability of DSS subject to water flow from flooding.
No specific technical considerations are identified for the application of SRP guidance under site specific natural hazards because the review focuses on kinematic stability of the cask system and not on structural integrity of individual components. Review of explosion hazards should confirm that maximum external pressure from a credible explosion does not cause a breach of the confinement boundary of the DSS, and the stress-intensity level is below the ASME B&PV stress limit. Structural review of the DSS under postulated fire (e.g., wildfire) includes evaluation of effects of increased internal pressure in the confinement boundary, changes in material properties, thermal stress, and spalling of concrete.
The SRP provides guidance for evaluating the thermal response of the DSS to ensure that limiting temperature of the fuel cladding is not reached due to off-normal and accident conditions. These conditions are partial vent blockage, adiabatic heat-up and fire accidents. The overpack is designed to provide natural circulation of air during storage for passive cooling of the canister. Partial blockage of the vent under off-normal conditions can cause the fuel cladding temperature to rise. For an adiabatic heat accident scenario, NUREG-2215 (NRC, 2020a) requires an evaluation of thermal response of the DSS components assuming no heat loss to the environment. Adiabatic heat results from design basis decay heat of the waste content. The applicant must evaluate the fire accident and show that the maximum temperature of MSR DSS components including fuel is below the allowable limit. The limiting temperature of MSR solid salt waste has not been established for dry storage. Beyond the melting point, phase change occurs in MSR salt from solid to liquid and the SRP guidance does not address storage of radioactive waste in liquid form. Additionally, 10 CFR 72.2 lists the scope of the material for a Part 72 license and often mentions solid form waste but not liquid.
Furthermore, it should be noted that finite element modeling of liquid behavior subjected to various accident loads is not straightforward for structural evaluation. A potential technical consideration is identified because the allowable temperature of MSR salt waste is needed to review accidents related to the thermal response of an MSR DSS. A more detailed evaluation of
3-12 technical bases and information on the physical characteristics of the fuel salt, including its liquidus temperature, may help bound allowable temperatures of MSR salt during storage.
Further work may be needed based on thermal evaluation during fire accidents and other scenarios to determine whether melting of MSR salt would be a potential issue.
Loss of shielding of DSS during storage can result from postulated accidents such as tornado missiles, explosions, fires, and container drop. NUREG-2215 guidance on loss of shielding accidents includes review of applicants evaluation of the maximum reduction of the radiation shielding thickness, material shielding effectiveness, or loss of temporary shielding in DSS or DSF SSCs resulting from the postulated accidents. No additional technical considerations are anticipated in reviewing loss of shielding for MSR salt waste storage.
4-1 4
TECHNICAL CONSIDERATIONS AND INFORMATION NEEDS WITH TRANSPORTATION OF MOLTEN SALT REACTOR WASTES Packages used for the transportation of spent nuclear fuel (SNF) must be certified by the U.S. Nuclear Regulatory Commission (NRC) in accordance with the regulations in Title 10 of the Code of Federal Regulations (10 CFR) Part 71. NUREG-2216 (NRC, 2020b) provides guidance to the NRC staff for reviewing Safety Analysis Reports for a Certificate of Compliance for transportation packages. To help the NRC staff prepare for the possibility of certifying transportation packages for Molten Salt Reactor (MSR) fuel salt waste, an assessment of the topics addressed in the standard review plan (i.e., areas of review) was conducted considering potential differences in the characteristics of MSR fuel salt waste relative to typical light-water reactor (LWR) spent fuel to identify potential technical considerations and associated information needs. This assessment is expected to further inform the NRC staff in evaluating whether additional preparation or guidance may be needed.
The assessment approach considers the descriptions of MSR fuel salt wastes by processing options that are documented in Chapter 2 of this report and develops scenarios for how these wastes could be packaged for transportation. Consistent with the focus on transportation package certification, the transportation scenarios provide general information and assumptions about what an MSR fuel salt waste transportation package might look like. The scenarios considered in this assessment are reasonably consistent with available information but should not be considered as an endorsement of any technology or approach.
4.1 Transportation Scenarios ORNL described a packaging concept with general design information related to MSR fuel salt waste transportation as part of an evaluation of Molten-Salt Reactor Experiment (MSRE) fuel salt waste disposition alternatives (Peretz, 1996). To allow for possible additional processing, handling, and transportation, Oak Ridge National Laboratory (ORNL) proposed filling an inner Hastelloy-N can with molten salt and then allowing the salt to cool and solidify in the can. The can would be placed inside a shielded container, and three shielded waste containers could then be placed inside a remote handled transuranic (RH-TRU) canister. ORNL presumed an RH-TRU 72-B transportation package could be used for transportation of the loaded RH-TRU canisters but did not conduct a detailed evaluation nor addressed whether the proposed approach would meet package specifications. The transportation scenario for this assessment assumes a similar can and canister system could be used for both storage and transportation.
Peretz (1996) provided general design information. For example, the salt waste can would have 1.27-cm [0.5-in] thick walls. This thickness was thought to provide an adequate corrosion allowance should uranium be separated from the salt by fluorination in the can. The can would then be inserted into a shielded container with the approximate outer dimensions of a 55-gallon drum. This container would provide 5.1 cm [2.0-in] of steel in each direction as a radiation shield. ORNL expected three of these shielded salt waste containers could be placed into an RH-TRU canister. An RH-TRU canister is made of 0.64-cm [0.25-in] thick carbon steel and be 307-cm [121-in] long and 66-cm [26-in] in diameter.
For a fluoride salt containing the gamma source from cesium or the 232U daughter chain, ORNL noted the evolution of radiolytic fluorine must be controlled (Peretz, 1996), though other sources in the spent fuel would probably cause radiolysis (not discussed in Peretz). ORNL proposed a getter, such as soda-lime (a mixture of sodium hydroxide and calcium oxide), to capture fluorine
4-2 gas. The containers would then be vented through the getter and a high-efficiency particulate absorbing (HEPA) filter, if necessary. ORNL noted the quantity of fluorine gas could be calculated and verified, thereby allowing the amount of getter to be determined. ORNL indicated additional effort was needed to determine whether such a transportation package could be certified.
Transportation of MSR fuel waste could occur from reactor sites as well as potential offsite storage or processing facilities and the timing and location of transportation depends on the waste management approach including whether waste processing is planned to occur onsite or offsite. Transportation could occur from reactor sites after a period of onsite storage, depending on available processing and disposal options. As of now, there is no information from MSR vendors to suggest how long these periods of storage could potentially be or whether waste processing will occur onsite or offsite.
Waste forms resulting from all three MSR fuel waste processing options (unprocessed, stabilized, separated) described in Chapter 2 are assumed to be placed in or solidified in a container that provides, for example, structural support for handling and an initial containment or confinement barrier; therefore, the conceptual canning approach described by ORNL is assumed to be broadly applicable to wastes associated with the three processing options.
4.2 Assessment of Potential Technical Considerations and Information Needs The following sections describe the results of an assessment to determine the technical considerations and information needs for certification reviews of transportation packages for MSR fuel waste. The assessment is conducted for review areas documented in the SRP for transportation package certification reviews in NUREG-2216 (NRC, 2020b). The assessment determined that the review plans for all areas of review are broadly applicable to package certifications for a wide variety of contents (e.g., LWR SNF and greater-than-Class-C (GTCC) wastes), including MSR fuel waste. The following sections highlight a subset of narrower topics within review areas where possible differences in the MSR fuel waste relative to LWR spent fuel may present additional technical considerations and information needs.
4.2.1 Structural Evaluation The objective of the structural review in Chapter 2 of the Standard Review Plant (SRP)
(NUREG-2216; NRC, 2020b) is the evaluation of structural design and performance of transportation packages. The structural components of a transportation package system provide containment, subcriticality, and radiation shielding functions under normal conditions during transportation and hypothetical accidents. The structural evaluation is coordinated with the review of information in other chapters of the SRP, namely, thermal, shielding, material, criticality, and containment. For evaluation of structural design and integrity of transportation packages the review in the SRP includes (i) description of structural design, (ii) general requirements of packages, (iii) lifting and tie-down devices, (iv) general considerations for structural evaluation, (v) normal conditions of transport, and (vi) hypothetical accident conditions. The performance of a package during air transport was not considered in this report.
Although many variations exist depending on the form and quantity of radioactive material, the typical transportation package for spent nuclear fuel and other high-level nuclear waste generally consists of an inner canister and an outer, lead-lined shell. The transportation package also includes impact limiters used for absorption of energy under normal and accident conditions. The canister is a cylindrical shell made of structural steel and placed inside a
4-3 transportation package which is a thick cylindrical vessel. The canister contains the fuel basket and fuel assemblies. A general description of a package for transportation of MSR fuel waste is currently not available and this is identified as a potential technical consideration. As discussed in Section 4.1, several MSR canisters or containers may be placed inside a canister. The MSR transportation package is envisioned to include a certified canister and outer shell design potentially with a fuel basket configured to the dimensions and number of MSR cans in the canister. Design of a transportation package with MSR salt waste is assumed to conform to applicable divisions and sections of American Society of Mechanical Engineers (ASME)
Boiler and Pressure Vessel (B&PV) and reviewed in accordance with SRP guidance.
Structural evaluation is performed either by analysis or physical testing to evaluate transportation packages under normal conditions and hypothetical accidents. The SRP provides detailed guidance for review of the structural evaluation to verify the appropriateness of the analysis models and computer simulations. The evaluation that results from the analysis should demonstrate that the structural performance of individual components of the transportation package meets acceptance criteria. For structural evaluation by testing, the review involves the description of test procedures and interpretation of test results showing adequate structural margin of safety in the transportation package components.
Under normal conditions of transport, NUREG-2216 (NRC, 2020b) provides guidance in Section 2.4.5 to confirm that the package is designed accounting for the following: (i) during the heat and cold conditions of transport, the thermal stresses of the packaging due to thermal expansion of all constituent components are within the allowable stress limits, (ii) stresses induced in the packaging from reduced external pressure and increased external pressure between inside and outside the packaging meet the stress limit criteria, (iii) vibration and fatigue are considered for any individual component including the fuel basket, MSR container, and MSR can (iv) packaging integrity is demonstrated under free drop and corner drop test loadings in combination with internal pressure, thermal, and other stresses; and (v) structural integrity of packaging is demonstrated under penetration test by dropping a vertical steel cylinder on the package surface. Note that these tests also include a water spray test for normal conditions of transport.
Guidance on review of hypothetical accident conditions is provided in SRP Section 2.4.6. The hypothetical accident conditions are based on a sequential application of tests (either analytical or physical) to evaluate that adequate structural integrity of the package components is cumulatively demonstrated. The hypothetical accident tests include (i) free drop from a height of 30 ft [9 m] onto an essentially unyielding surface considering combined effects of internal pressure and thermal stresses and impacted at an orientation to induce maximum damage, (ii) puncture test by free drop from a height of 40 inches [1 m] onto a vertically-oriented cylindrical mild steel bar mounted on an essentially unyielding, horizontal surface, and (iii) thermal test with test specimen exposed to fully engulfing fire at an average temperature of 800ºC [1,475ºF] for 30 minutes or its equivalent. The test results are verified in accordance with SRP guidance to confirm that the containment, thermal performance, shielding, and subcriticality requirements are met during transportation.
Significant NRC staff experience exists to review analytical calculations and structural simulations of transportation packages under normal and accident conditions, considering SNF assemblies and the basket. However, structural simulation of a transportation package with MSR salt waste may involve additional technical considerations because of differences in structural configuration, physical nonlinearities, inertial load, and material properties of MSR containers compared to SNF assemblies. A more detailed evaluation of the technical bases and
4-4 information in the SRP that are limited to LWR SNF may help to clarify whether additional guidance is needed to support reviews of structural integrity of transportation packages with MSR fuel waste. The NRC staff has experience in review of physical testing of certified transportation packages for SNF and no additional technical considerations are anticipated in reviewing structural integrity by testing.
4.2.2 Thermal Evaluation As documented in the SRP for transportation package certifications (NRC, 2020b), the thermal evaluation with regard to heat transfer and flow ensures the applicant adequately evaluated the thermal performance of the transportation package design based on thermal tests specified under normal conditions of transport, short-term operations (e.g., drying and backfilling), and hypothetical accident conditions, and that the package design meets the thermal performance requirements of 10 CFR Part 71, Packaging and Transportation of Radioactive Material.
Specific areas of review include description of the thermal design (packaging design features, codes and standards, content heat load specification, summary tables of temperatures, summary tables of pressures in the containment vessel); material properties and component specifications (material thermal properties, specifications of components, thermal design limits of package materials and components); general considerations for thermal evaluations (evaluation by analyses, evaluation by tests, confirmatory analyses, effects of uncertainties, conservatisms); evaluation of accessible surface temperatures; thermal evaluation under normal conditions of transport (heat and cold, maximum normal operating pressure); thermal evaluation under hypothetical accident conditions (initial conditions, fire test, maximum temperatures and pressures).
The methods applied to these evaluations of heat transfer and flow characteristics would be typical of a current transportation package once the heat source has been adequately defined.
Some details of the guidance that are not directly applicable to MSR fuel waste include references to LWR fuel assemblies and characteristics that are only applicable to LWR fuel assemblies (e.g., cladding and cladding temperature limits). Additionally, some processed MSR fuel salt waste streams would have different compositions and therefore would vary regarding decay heat and therefore may involve separate or bounding decay heat calculations.
Technical considerations and information needs were identified for a subset of review areas applicable to the description of thermal design (content heat load specification, summary tables of temperatures, summary tables of pressures in the containment vessel); general considerations for thermal evaluations (evaluation by analyses, confirmatory analyses, effects of uncertainties); thermal evaluation under normal conditions of transport (heat and cold, maximum normal operating pressure); thermal evaluation under hypothetical accident conditions (initial conditions, fire test, maximum temperatures and pressures).
Regarding pressure analysis, typical fission product gasses mentioned in the SRP (e.g., krypton, xenon, and tritium) are expected to be removed during MSR operations (Riley et al., 2018); however, other reactive gases may apply to fuel salt waste depending on reactor design and operations (e.g., fluorine and uranium hexafluoride; Section 2.1). Reactive gases that are different than gases in LWR spent fuel may involve different considerations including additional design or components (e.g., use of getters) and would have different properties and potentially different thermal characteristics (Section 4.1).
For the remaining areas of review in the thermal evaluation it was determined that the methods applied to these evaluations of heat transfer and flow characteristics would be typical of a
4-5 current transportation package once the heat source has been adequately defined. These areas are the description of thermal design (packaging design features, codes and standards);
material properties and component specifications (material thermal properties, specifications of components, thermal design limits of package materials and components); general considerations for thermal evaluations (evaluation by tests); general considerations for thermal evaluations (conservatisms); thermal evaluation of accessible surface temperatures; thermal evaluation under normal conditions of transport.
Further analysis of possible waste forms and their characteristics, including thermal characteristics, would help to reduce uncertainties about likely options that may be pursued in future applications. Improved characterization of MSR fuel waste radionuclide inventories for various design and operation options would help to bound the range of possibilities and facilitate comparisons with LWR spent fuel and other wastes that can inform preparations for safety reviews.
4.2.3 Shielding Evaluation As documented in the SRP for transportation package evaluations (NRC, 2020b), the shielding evaluation verifies that the design of Type B transportation packages meets the external radiation requirements of 10 CFR Part 71. The areas of review for the shielding evaluation include description of shielding design (shielding design features, codes and standards, summary tables of maximum external radiation levels); radioactive materials and source terms (source-term calculation methods, gamma sources, neutron sources); shielding model and model specifications (configuration of source and shielding, material properties); shielding evaluation (methods, code input and output data, fluence-rate-to-radiation-level conversion factors, external radiation levels, confirmatory analyses).
The methods applied to these evaluations of shielding design, radioactive materials and source terms, shielding model and model specifications and shielding evaluation would be typical of a current transportation package, except in the determination of the radionuclide inventory of the contents and a potential need for principal design criteria for MSR fuel salt waste cans if used.
Also, some processed (e.g., separated, or concentrated) MSR fuel salt waste streams would have different compositions than the fuel salt waste that was removed from the reactor, and therefore processing attributes may be more important than typical LWR characteristics such as initial enrichment, cooling time, and burnup for defining the radiation source for processed waste streams.
Technical considerations and information needs were identified for a subset of review areas relating to description of shielding design, radioactive materials and source terms, shielding model and model specifications (material properties), and shielding evaluation (code input and output data, confirmatory analyses). These review areas are affected by the fuel salt waste inventory which has unique technical considerations that have been previously described in Section 3.2.3. Radionuclide inventory in the salt waste would also affect both applicant shielding calculations and staff confirmatory calculations.
Shielding design features are expected to be typical of current transportation packages, however, principal design criteria for MSR cans would need to be defined (e.g., internal pressure, temperature, material specs, and applicable codes and standards). The SRP may need to be revised to include specific review methods and acceptance criteria for MSR cans.
4-6 For the remaining areas of review in the evaluation corresponding to the shielding model and model specifications (configuration of source and shielding); shielding evaluation (methods, fluence-rate-to-radiation-level conversion factors, external radiation levels), no additional technical considerations were identified because the approach applied to these evaluations would be typical of a current transportation package once the inventory for the MSR fuel salt has been adequately defined. Regarding configuration of the source, a potential theoretical concern would be the potential for radioactive materials in liquid fuel salt waste to segregate during cooling, as the phase changes from liquid to solid. ORNL previously evaluated the potential for segregation of uranium in the context of criticality (Sections 3.2.4 and 4.2.4),
however, did not report similar studies for other elements. This could affect assumptions of source homogeneity in shielding calculations and therefore may warrant further consideration in preparations for future transportation package certification reviews.
Further analysis of possible waste forms and their characteristics, including radiological characteristics, would help to reduce uncertainties about likely options that may be pursued in future applications. Improved understanding of MSR fuel waste radionuclide inventories for various design and operation options would help to bound the range of possibilities and facilitate comparisons with LWR spent fuel and other wastes that can inform preparations for safety reviews.
4.2.4 Criticality Evaluation As documented in the SRP for transportation package certification (NRC, 2020b), the criticality evaluation verifies that the transportation package design meets the nuclear criticality safety requirements in 10 CFR Part 71. The areas of review for the criticality evaluation include description of criticality design (packaging design features, codes and standards, summary table of criticality evaluations, criticality safety index); contents; general considerations for criticality evaluations (model configuration, material properties, analysis methods and nuclear data, demonstration of maximum reactivity, confirmatory analyses, moderator exclusion under hypothetical accident conditions); single package evaluation (configuration, results); evaluations of package arrays (package arrays under normal conditions of transport, package arrays under hypothetical accident conditions, package arrays results and criticality safety index); benchmark evaluations (experiments and applicability, bias determination); burnup credit evaluation for commercial LWR SNF (limits for the certification basis, model assumptions, code validation isotopic depletion, code validationkeff determination, loading curve and burnup verification).
It was determined that the criticality areas of review in the SRP include some aspects that are only applicable to reviews of LWR fuel storage. Otherwise, the concepts and methods applied to criticality evaluations of MSR salt waste packages would be typical of current criticality evaluations, except in determining the inventory of the contents and some implementation details. An area of review found to be not directly applicable to MSR fuel waste included burnup credit evaluation for commercial LWR SNF (loading curve and burnup verification), which considers the potential for variation in burnup among fuel assemblies that might be loaded into a transportation package. Also, some processed (e.g., separated, or concentrated) MSR fuel salt waste streams would have different compositions than the fuel salt waste removed from the reactor and therefore they may not contain fissile material rendering criticality areas of review not applicable.
The potential need for further assessment of criticality evaluation methods was identified for all directly applicable review areas identified including description of criticality design (packaging design features, codes and standards, summary table of criticality evaluations);
4-7 contents; general considerations for criticality evaluations (model configuration, material properties, analysis methods and nuclear data, demonstration of maximum reactivity, confirmatory analyses); benchmark evaluations (experiments and applicability, bias determination); burnup credit evaluation for commercial LWR SNF (limits for the certification basis, model assumptions, code validationisotopic depletion, code validationkeff determination). Most of these review areas are affected by the fuel salt waste inventory, which has unique technical considerations that have been previously described in Section 3.2.3.
These review areas also include technical bases and references to existing data that are specific to PWR SNF (e.g., burnup credit). Along these lines, it may be reasonable to consider the potential need to develop a technical basis for accepting a burnup credit approach for MSR fuel waste.
Review of criticality design features is expected to broadly involve considerations typical of existing transportation packages; however, principal design criteria for MSR cans may need to be defined (e.g., internal pressure, temperature, material specifications, and applicable codes and standards). The SRP may need to be revised to include specific review methods and acceptance criteria for MSR cans. Additional potential technical considerations relating to configuration and maximum reactivity (general considerations for criticality evaluations) for unprocessed MSR fuel waste, especially owing to its relatively low melting point relative to thermal tolerances for LWR SNF and related SSCs were identified. For example, the reported liquidus temperature for MSRE Li2BeF4 fuel salt is 434 °C [813 °F]. For context, the SRP maximum cladding temperature threshold for all burnups under hypothetical accident conditions is 570 °C [1,058 °F]. Therefore, consideration of the potential for reconfiguration under accident conditions may be indicated if credible temperatures inside an MSR fuel waste can or container exceed the liquidus temperature. A related potential theoretical concern would be the potential for uranium in liquid fuel salt waste to segregate during cooling, as the phase changes from liquid to solid. For the MSRE, ORNL reported the results of an experiment where simulated fuel salt was solidified at rates that would approximate the cooling rate in the reactor drain tanks after reactor shutdown (Peretz, 1996). The objective of this experiment was to determine whether uranium segregation could take place as crystalline phases began to form, but most of the salt was still liquid. Uranium samples taken from the solidified salt plug ranged from 2.94 weight-percent near the top to 5.74 weight-percent at the bottom. The degree of segregation was considered by ORNL to be small and not affect criticality safety in the drain tanks. This may not necessarily be the case for future fuel salt, especially considering that some MSR designs are considering uranium fuel at much higher enrichment (~20 percent) and a criticality analysis would likely be needed.
Benchmark evaluations (i.e., review of applicant benchmarking of computer codes for criticality calculations against appropriate experiments) may necessitate development of benchmark experimental data. The SRP indicates the benchmark experiments should have, to the maximum extent possible, the same materials, neutron spectrum, and configuration(s) as the package evaluations for each type of content.
For the remaining areas of review in the evaluation, related to the criticality safety index; moderator exclusion under hypothetical accident conditions; single package evaluation (results);
evaluations of package arrays (package arrays under normal conditions of transport, package arrays under hypothetical accident conditions, package arrays results and criticality safety index), it was determined that these evaluations would be typical of a current transportation package once the inventory for the MSR fuel salt has been adequately defined and other technical considerations previously described have been adequately addressed.
4-8 4.2.5 Materials Evaluation As described in NUREG-2216 (NRC, 2020b), the materials evaluation addresses materials performance of the transportation package. The materials review ensures that the materials meet applicable codes, standards, and specifications to support the intended functions of the components under normal conditions of transport and hypothetical accident conditions. Specific areas of review include the drawings, codes and standards (usage and endorsement, ASME code components, code case use and acceptability, non-ASME code components), weld design and inspection, mechanical properties (tensile properties, fracture resistance, tensile properties and creep of aluminum alloys at elevated temperatures, impact limiters), thermal properties, radiation shielding (neutron-shielding materials, gamma-shielding materials),
criticality control (neutron-absorbing material specification, computation of percent credit for boron-based neutron absorbers, qualifying properties not associated with attenuation), corrosion resistance (environments, carbon and low allow steels, austenitic stainless steel), protective coatings (review guidance, scope of coating application, coating selection, coating qualification testing), content reactions (flammable and explosive reactions, content chemical reactions, outgassing, and corrosion), radiation effects, package contents, fresh fuel cladding, spent nuclear fuel (spent fuel classification, uncanned spent fuel, canned spent fuel), bolting material and seals (metallic seals, elastomeric seals).
The areas of review in the materials evaluation in NUREG-2216 (NRC, 2020b) address fundamental aspects of transportation package materials evaluation that are adaptable to variations in proposed transportation package contents. Review of the codes and standards used for transportation package design and construction in the context of MSR fuel waste may need to be considered. The review method in NUREG-2216 calls for verification of the codes and standards for packaging components important to safety; however, the applicability of codes and standards for transportation package design and fabrication for transport of MSR fuel salt waste has not been evaluated. In the absence of specific codes and standards, the design and fabrication for transportation package design for MSR fuel waste, including the technical basis for the adequacy of existing codes and standards, will likely need to be provided by the applicant. A potential challenge was also identified pertaining to the review of the thermal properties of materials. The review method in NUREG-2216 calls for verification of the thermal properties and the change in these properties due to material degradation. However, data characterizing thermal properties of MSR fuel waste are limited. Therefore, the thermal properties and the effect of degradation of MSR fuel waste should be evaluated.
For the review of chemical reactions, outgassing, and corrosion among the contents and between the contents and the transportation package components, the review method in NUREG-2216 (NRC, 2020b) calls for examining whether corrosion wastage could lead to a loss of intended functions. The current review method is limited to guidance for the examination of corrosion of materials such as stainless steel and zirconium alloy-clad UO2 fuels. The chemical interactions of air and water with the MSR fuel waste or associated cans or other components that are relied upon for structural support and containment should be evaluated. A similar potential challenge was also identified pertaining to the review of the package contents. For the review of the spent nuclear fuel regarding uncanned spent fuel, the safety analyses of uncanned spent fuel rely on the integrity of the fuel cladding for maintaining the analyzed configuration.
The current review method is limited to guidance for the performance of zirconium alloy, aluminum alloy, or stainless steel-clad UO2 fuels. Therefore, the performance of MSR fuel waste under transportation environments should be evaluated.
4-9 The general acceptance criteria established in NUREG-2216 (NRC, 2020b) pertaining to materials performance of the transportation package were determined to be not impacted by specifics of the technologies for transport of MSR fuel waste.
Further analysis of possible waste forms and their performance would help to reduce uncertainties about likely options that may be pursued in future applications. A more detailed evaluation of the technical bases and information in the SRP that are limited to LWR SNF may help to clarify whether additional guidance is needed to support reviews of MSR fuel waste.
Improved characterization of the chemical interactions between MSR fuel waste and the transportation package components under transportation environments can inform preparations for safety reviews.
4.2.6 Containment Evaluation The guidance for reviewing the containment evaluation of transportation packages is documented in Chapter 4 of the SRP (NUREG-2216; NRC, 2020b). The areas of review include description of containment system, general considerations, containment under normal conditions and hypothetical accident conditions. The SRP addresses Type B packages that can be used for the transportation of SNFs. The review description of the containment system ensures that the applicant provided information on containment boundary design features describing vessel, welds, penetrations, and closure devises (e.g., O-rings, seals, lids, and cover plates). The information should also include identification of appropriate codes and standards associated with the design, fabrication, testing (e.g., helium leakage rate testing),
inspection, and certification of the containment system.
Under normal conditions of transport, the SRP calls to verify that the maximum permissible release rate and maximum permissible leakage rate are calculated in accordance with ANSI N14.5-2022, Leakage Tests for Radioactive Materials, using reasonable source terms.
The maximum temperature and maximum operative normal pressure in these analyses should be verified to be consistent with thermal analysis and pressure in the containment vessel based on normal transport conditions including temperature, release of gases and volatiles from fuel rod cladding breaches, and vaporization of contents. The SRP provides guidance to verify that the source term calculations are consistent with constituents of the SNF including fuel type, fuel amount, percent enrichment, burnup, cool time, and decay heat. The SRP provides guidance to verify that the package meets the containment requirement demonstrated by (i) leak tests, (ii) analysis, and (iii) by acceptance testing.
The containment evaluation for hypothetical accident conditions is similar to the evaluation of normal conditions of transport with noted differences including: (i) the pressure in the containment vessel is based on hypothetical accident; (ii) releasable source terms, maximum permissible release rate and maximum permissible leakage rate is based on containment requirements for hypothetical accident conditions; (iii) reference air leakage rate calculated for hypothetical accident conditions is greater than normal conditions of transport; and (iv) containment requirements for hypothetical accident conditions is applied individually for krypton-85 and the other radioactive materials.
As previously described, source terms will need to be determined for analysis. Differences in the properties of the MSR fuel waste material and its response to thermal conditions may require different considerations, data, or information related to estimating release fractions and rates relative to typical LWR SNF evaluations.
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SUMMARY
AND CONCLUSIONS To assist the U.S. Nuclear Regulatory Commission (NRC) staff in preparing for the possibility of certifying dry storage systems (DSSs) and transportation packages for Molten Salt Reactor (MSR) fuel salt waste, the Center for Nuclear Waste Regulatory Analyses (CNWRA) assessed potential technical considerations and information needs associated with the certification of these systems under the existing regulatory framework. This assessment involved describing the characteristics of wastes and waste forms considering possible ways wastes and waste forms could vary depending on the type of reactor, operational characteristics, and processing optionsand identifying plausible storage and transportation scenarios. A transportation packaging concept previously documented by Oak Ridge National Laboratory (ORNL) for Molten-Salt Reactor Experiment (MSRE) salt waste remediation planning formed the basis for developing a broad storage scenario and a broad transportation scenario. These assumed scenarios involved solidifying molten fuel salt waste in portable cans placed in shielded containers that were sized for compatibility with existing canister or cask/package systems. The approach then evaluated areas of review for key evaluations addressed by NRC Standard Review Plan (SRPs) for the certification of dry storage systems and transportation packages.
The review documented in this report considered whether areas of review in the SRPs were applicable to DSSs or transportation packages in the context of expected characteristics of liquid-fueled MSR fuel salt waste based on available technical information.
Most of the areas of review were found to be applicable to MSR fuel waste, although a small subset was specific to light-water reactor (LWR) spent fuel assemblies or related components that cannot be generalized to MSR fuel waste (e.g., fuel assemblies, cladding, fuel pellets, hardware, and details of fuel condition).
For many applicable areas of review the existing review approach in the SRP is considered valid and adaptable to MSR fuel waste.
A subset of the areas of review were found to be applicable but further technical assessment may need to be considered (e.g., notable differences in MSR fuel salt waste or its management relative to LWR SNF may call for different review methods, additional information, or more NRC staff review).
Regarding structural evaluation of an MSR DSS, the DSS components (e.g., can, container, canister, and basket, as applicable) may be unique for an MSR DSS. In the absence of proposed MSR packaging for storage and disposal, MSR fuel waste is assumed to be placed in cans and multiple cans may be placed inside a canister separated by a metal fuel basket. It is also conceivable, that remote handled transuranic (RH-TRU) or similar canisters holding several fuel waste cans are inserted into storage canisters or overpacks. While there may be differences in structural configuration, inertial load, and inherent gaps between components compared to an SNF DSS, similar modeling approaches may be applicable for evaluating an MSR DSS and the integrity of the components. Although NUREG-2215 provides guidance for ensuring structural integrity of SNF cladding under drop accidents, the acceptance criteria for cladding would not be directly applicable to MSR cans. Therefore, a potential technical consideration was identified related to verifying the structural integrity of MSR cans. No additional technical considerations were identified for evaluating an MSR basket. The need to determine the radionuclide inventory of the MSR fuel salt was identified as having an impact on several areas of review. The inventory is considered determinable with available tools, but it may require specialized physics modeling that accounts for unique dynamics of the MSR
5-2 reactor design and operation and may also require additional or different supporting data or information or more review time than a typical certification involving LWR SNF. Examples of review areas that consider inventory include thermal analysis (decay heat); shielding (defining the source); criticality (material properties; design criteria, model specification, analysis, burnup credit); confinement and containment (isotopic inventory), and accident source terms.
Technical considerations were associated with unique material characteristics of MSR fuel waste relative to LWR SNF. These unique characteristics, which can vary with potential processing options, include its chemistry, reactivity, potential for gas generation, mass, homogeneity, and structural stability. Regarding the materials evaluation, further assessment may be needed to determine the applicability of existing codes and standards to MSR salt waste, and the evaluation of environmental degradation, corrosion and other reactions on containers or cans. These containers or cans may be expected to provide structural and containment or confinement safety functions for the various MSR salt waste forms. The thermal evaluation also accounts for the effect of gas generation from DSS or package contents affecting pressure and temperatures within the initial containment. The additional potential for gas generation from MSR fuel salt waste may present additional technical challenges regarding reactivity, pressurization, and potential thermal effects. The melting point of unprocessed MSR fuel waste could also affect reconfiguration of contents (criticality, confinement, and containment evaluations), determining accident release fractions, and atmospheric modeling of releases (accident evaluation).
The assessment of technical considerations documented in this report is an initial screening of topics with the objective of highlighting those that may be considered for further assessment.
This includes key differences from typical LWR reviews that, for example, may require additional information, a modified approach, or take more review time to complete. Increased understanding of these differences is expected to further inform the NRC staff and management in evaluating whether additional preparation or guidance may be needed to address the increased industry interest in advanced reactors including potential applications of MSR technology.
Sufficient technical information to support license applications would be expected to be provided by applicants, but some internal NRC efforts may be appropriate to prepare for or enhance the staff capabilities to conduct future licensing reviews. The following general and specific recommendations are provided on addressing the needs identified in this report:
Continue ongoing interactions with technical staff responsible for safety reviews and obtain feedback on technical considerations and whether additional preparations or guidance are needed on specific topics. If additional guidance or supplementation is needed to address MSR-specific technical considerations, this could be accomplished by developing new appendices or focused reports.
Focus future MSR related research activities on evaluating technologies that are most likely to be used by early adopters of MSR technology. There is a wide variety of potential design and operational options associated with MSR technology. Focusing preparations on technologies most likely to be proposed by future applicants can help limit the potential options and associated information needs.
5-3 Engage potential applicants during pre-licensing interactions regarding plans for waste management activities involving storage and transportation to improve and narrow the likely approaches that would be used and the associated information needs.
Conduct further research or monitoring of technical developments regarding potential waste forms to narrow the field of potential options expected to be utilized in future projects.
Conduct further research or monitoring of technical developments regarding the chemical interactions of air and water with the MSR fuel waste or associated cans or other components that are relied upon for structural support and containment or confinement.
Conduct further research or monitoring of technical developments regarding radionuclide inventory and source terms for MSR technologies to help clarify potential technical considerations.
Conduct further research or monitoring developments regarding MSR specific accident scenarios involving fire to improve risk insights related to the potential for melting of solid fuel salt waste.
Evaluate the applicability of codes and standards for transportation package design and fabrication with new materials. Specific codes and standards to be used for transportation package design and fabrication for transport of MSR fuel waste may be defined or developed, or the technical basis for the adequacy of alternative codes and standards may be provided by applicants.
Evaluate the thermal properties and the effect of degradation of MSR fuel waste. The review method in NUREG-2216 calls for verification of the thermal properties and the change in these properties due to material degradation. However, data characterizing thermal properties of MSR fuel waste are limited.
Evaluate the performance of MSR fuel waste under transportation environments with regard to maintaining configuration.
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