Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Accident”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 631 records · Page 35
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Boiling-Water Reactor Testing Capability in the Advanced Test Reactor

I-Loop is an irradiation facility that is currently being installed at the Advanced Test Reactor. It is a two-loop test facility capable of performing Light Water Reactor (LWR) irradiations in prototypic coolant conditions. The two loops are being installed to be capable of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) pressure, temperature, and chemistry environments. Each loop is nominally dedicated as a BWR or PWR for simplicity of operations. In-reactor water loop testing that an I-Loop provides is key to the deployment of new accident tolerant fuel technologies and other advanced LWR fuel concepts. Currently, pressurized water loops are the only testing facilities available to test BWR fuel concepts. Their test environments are non-prototypic at higher pressure/temperature and at single-phase fluid flow conditions. This void in the LWR test bed capabilities is one that the I-Loop is uniquely situated to provide. This report discusses the mechanical design, thermal hydraulic calculations, and neutronic calculations of a proposed standard experiment of accident tolerant BWR fuel concepts. Mechanical design examines the geometry and features of the main components. Thermal hydraulic calculations examine the modeling and results of the two-phase flow options available. Lastly, neutronic calculations examine the Monte Carlo analysis of enrichment, heat rates, and flux spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Transfer Factors for the FRMAC Assessment Manual and Turbo FRMAC to Improve Radiological Dose Assessment

The Turbo FRMAC analysis tool is used to perform complex calculations to quickly evaluate radiological consequences and aid in decision making during an emergency response by assessing impacts to the public, workers, and the food supply. Turbo FRMAC calculations are based on methods established by the Federal Radiological Monitoring and Assessment Center (FRMAC). To be able to assess impacts, input data called transfer factors that describe radionuclide uptake by local plants and animals are required. During the code application exercises in late 2016, identifying, finding, and validating nonstandard transfer factors proved to be time consuming and diverted the teams’ activity away from other critical tasks. As a result, a task was undertaken to dramatically expand the list of available transfer factors (food and non-food) and incorporate these factors into Turbo FRMAC. This will ultimately result in improved efficiency of the assessment team to perform calculations during times when the FRMAC is activated and provide more defensible, vetted data from which to calculate results. As a result of these and subsequent exercises, transfer factors were needed for the following items: bell peppers, Christmas tree, deer, flowers, fresh cucumbers, tomatoes, grapefruit, lichens, mushrooms, oranges, snap beans, squash, strawberries, sugarcane for sugar and seed, sweet corn, tea, tobacco, tree bark, and watermelon. To expand the applicability of the information tables, generic transfer data were also provided for common categories and recommendations were made for expanding the list of chemicals based on chemical similarity. The data presented in this report were compiled from recent literature with most of the data encompassing the period from 2000 to 2018. The following radionuclides were targeted during the literature search: elements associated with reactor accidents or nuclear detonations (Sr 89/90, Cs 134/137, Ce 141/144, Ru 103/106, I 129/131/133, Pu 238/239, Am 241, Zr 95, and Nb 95) and elements associated with industrial accidents or dirty bombs (Ir 192 and Co 60). Data from other elements were evaluated if they were identified during the literature search. In addition, reports were evaluated that were recommended by the research consultants. For each plant or animal transfer factor, the goal was to determine the geometric mean, the geometric standard deviation, the minimum, the maximum, and the number of measurements used. If only one measurement was available that was presented as the mean. Concentration ratios with large geometric standard deviations (GSDs) were generally the result of a paucity of measurements or a few disparate measurements. Measurement disparity was observed for data from different soil types. The geometric mean (GM) is a good reference value for planning and responses purposes, but the location-specific concentrations are unlikely to be similar to the model results. Recommendations for further work include: segregating the data to reflect the influence of soil type, developing approaches to incorporate animal data based on aggregated transfer measurements, and including data for foliar deposition on plants.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2024

Report on RIA Relevant Modified Burst Testing of ATF Cladding Materials

The mechanical performance of accident-tolerant fuel (ATF) cladding candidates in light-water reactors (LWRs) must be similar to or better than that of current conventional nuclear fuel claddings to reduce dose to the public and ensure a that the core coolable geometry is maintained during a postulated reactivity-initiated accident (RIA) in light-water reactors (LWRs). During an RIA event, the rapid thermal expansion of nuclear fuel can deform the cladding once the fuel–cladding gap closes. In some cases, the pellet–cladding mechanical interaction (PCMI) can induce mechanical failure in ATF candidates. Thus, the mechanical response of ATF cladding must be investigated by mimicking the conditions of RIA and potentially performing Transient Reactor Test (TREAT) experiments to establish or verify the safety envelope. The work presented in this report investigated the failure behavior of as-received, hydrided, and chromium-coated (Cr-coated) Zircaloy-4 (Zry-4) cladding tube under strain-driven mechanical conditions, mimicking postulated RIA loading conditions. Mechanical testing was performed at 300°C via modified burst test (MBT) equipment with pulse width control previously developed under the Department of Energy’s (DOE’s) Advanced Fuel Campaign (AFC). The mechanical strains were determined using 2D digital image correlation (DIC) techniques. The base Zry-4 acquired by Cameco Inc. was in stress-relieved annealed (SRA) condition. Because of the observed large deformation of the cladding tubes, the failure strain definition was updated for the MBT, which can also be applied to other tube tests where significant bulging (out-of-plane deformation) is present. The failure strain was determined to be affected by the speed of the test or the RIA event. As the RIA-like event duration decreased from 75 to 15 ms, the failure strain decreased 5, 7 and 1% for as-received, hydrided, and Cr-coated specimens, respectively. Fractography on the Cr-coated specimens indicated the presence of two failure mechanisms: (i) crenulation at the outer surface of the coating due to the tensile tractional forces along with coatings grain microstructure and (ii) formation of critical defect at the coating/cladding interface that initiated coating rupture after severe plastic deformation of the Zry- 4 substrate. Based on the MBT results and fractography observations, performing mechanical property testing at high strain rates—in particular on Cr-coated tubes—and semi-integral TREAT experiments are required future efforts to ensure ATF cladding performance during transients. The testing recommended would also inform the development of long-term generalized cladding technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

NEAMS Burnup Extension Accomplishments and Remaining Modeling Gaps

The economic viability of light-water reactors (LWRs) in the United States is declining in heavily subsidized markets, and as a result, the nuclear industry is looking for opportunities to enhance the economic competitiveness of nuclear power. This is not a foreign concept to the nuclear industry: in the mid-2000s, the nuclear industry set out to achieve zero fuel failures by 2010. The goal in this effort was to drive down the cost of reactor shut down by replacing a pin or bundle in response to fuel rod failure. 2010 brought about the initiative to deliver the nuclear promise to reduce operating cost by 30% to improve nuclear energy’s economic competitiveness before 2020. The emergence of accident-tolerant fuel also offers the nuclear industry an opportunity to build on these past successes and deliver affordable, clean energy. Accident-tolerant fuel has been shown to provide superior performance compared to traditional Zircaloy/UO2 fuel concepts, offering the unique ability to remove operational limitations that inhibit the economic viability of nuclear power. This has led the industry to begin building a technical case to extend the peak rod average burnup beyond 62 GWd/tU to extend pressurized water reactor cycle lengths to 24 months and to develop more efficient boiling water reactor core designs. The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program mission is to develop advanced modeling and simulation tools and capabilities to accelerate the deployment of advanced nuclear energy technologies. The primary safety concern inhibiting the nuclear industry from extending burnup is related to high-burnup fuel fragmentation, relocation, and dispersal. Therefore, the NEAMS program developed a targeted 5-year plan to support the industry’s efforts to extend burnup. This milestone report summarizes the 5-year plan that was enacted in FY20, followed by a discussion of the ongoing activates required to fulfill the 5-year plan, as well as the approach to address the current modeling gaps. Additionally, an LWR stakeholder meeting was held to communicate work performed in the NEAMS program over the past three years, to assess the LWR community’s perspective on the impact of the program, and to identify remaining significant gaps in the NEAMS suite of capabilities. This engagement will be documented by the Electric Power Research Institute and used by NEAMS to redirect current LWR scope as needed and to develop the next phase for LWR research and development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Post-irradiation examination of legacy high burnup fuel to support safety testing

Safety/transient testing to evaluate performance under off-normal conditions is an essential pillar for both the development of Accident Tolerant Fuels (ATF) and the optimization of fuel operation economics beyond current discharge burnups. Among other factors, the successful interpretation of the transient testing results relies upon the knowledge of the initial conditions of the test, including the characteristics of the fuel system under scrutiny. When testing pre-irradiated material, the assumptions that the fuel and the cladding still have the same properties as in the pre-irradiation stage is obviously wrong and could affect the results of the test. This is particularly true the more burnup accumulates in the fuel rod and irradiation progresses. The knowledge of the initial microstructure of both fuel and cladding allows a clearer interpretation of the subsequent transient testing results, provides validation of the physical phenomena underlying the model predictions and eliminates the uncertainties related to the limited knowledge of the sample status before the test. One example is the phenomenon of fine fragmentation that occurs in Light Water reactor (LWR) fuel. During a Loss of Coolant Accident (LOCA) or Reactivity Initiated Transient (RIA) the fuel can severely fragment. During LOCA, high burnup fuel tend to finely fragment, which has raised safety concerns due to the increased likelihood of dispersal of such small particles once the cladding has burst and due to the increased fission gas release. Therefore, efforts have been devoted to the assessment of a pulverization threshold that could determine the conditions under which fine fragmentation is predominant. However, the lack of information regarding the initial conditions of the fuel, and the connections between those conditions and the pre-transient irradiation history, have hindered the development of a fully mechanistic fragmentation and pulverization criterion. The empirical relationships rely on conservative estimations, due to the lack of information on critical material properties and characteristics. More generally, experimental evidence of the irradiation-induced modifications at microstructural scale are necessary to determine the behavior of the material at the macroscopic scale, with the latter being the one of technological interest. Significant progress has been made in the last two decades in the developments of analytical materials science techniques that can be applied to highly radioactive materials, such as high burnup fuels. The availability of new techniques and the improvement of existing ones has enabled investigations previously not possible that can deepen the understanding of the fuel characteristics and properties at high burnup. The better knowledge of material behavior and irradiation-induced phenomena could help the prediction of its performance. In this context, the scope of the present work is to apply a wide portfolio of advanced characterization techniques to determine properties that are relevant for safety and performance. The results are interpreted in the context of engineering scale post-irradiation examinations and available information on the irradiation conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Benchmarking MELCOR's NAC Package to ABCOVE Tests AB5 and AB6

This report presents analyses of the AB5 and AB6 ABCOVE sodium spray fire experiments with the MELCOR code. This code simulates the progression of accident events for analysis and auditing purposes of nuclear facilities during accident conditions. Historically, the ABCOVE experiments have contributed to the validation of aerosol physics and related phenomena. Given advancements in sodium-cooled reactor designs, characterization of the sodium spray combustion may further the review and validation of newly incorporated sodium properties and physics packages, namely, the sodium equations of state (EOS) and the sodium combustion (NAC) package within MELCOR. By analyzing the AB5 and AB6 experiments with and without the NAC package, sodium specificity for spray combustion and aerosol formation as well as speciation of the combustion products are reviewed with the new packages. This effort provides code users with a demonstration of the current code capabilities. This report provides the current best practices for the NAC package as well as a discussion of any issues observed while performing the presented analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Residual Stress Measurements of Cr-coated Cladding

Accident-tolerant fuel concepts for light-water reactor applications have been developed and tested in diverse research programs around the world. Industry teams have developed coated cladding concepts using different coating application methods, such as physical vapor deposition and coldspraying. Recently, variations in thickness and application method were investigated, and their effects on the microstructure and mechanical properties were reported. The results suggest an increase in performance in an unirradiated state, however, it remains unclear what phenomenon drove this performance. One explanation to this performance is associated with the coating process imparting a residual stress. Residual stress analysis was performed on coatings produced with various process parameters to understand how a thin layer of Cr impacts the mechanical properties after coating application and after simulated operating conditions. The gathered information is intended to address safety concerns or support margin identification related to accident-tolerant fuels and to support subsequent modeling and simulation efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Recent High Burnup LOCA Testing at Oak Ridge National Laboratory

To improve fuel cycle economics, increasing the fuel burnup limit in light-water reactors requires a solid technical foundation. Observations from experiments at Halden and Studsvik highlighted severe fuel fragmentation during loss-of-coolant-accident (LOCA) conditions, indicating a need for further technical considerations. These experiments suggest that the fragmentation threshold for high-burnup fuel might be influenced by pre-transient power levels. Consequently, additional LOCA test data are essential to complement existing findings and to deepen our understanding. Oak Ridge National Laboratory’s Severe Accident Test Station has been instrumental in advancing knowledge about high-burnup fuel fragmentation, relocation, and dispersal. This milestone report details two high-burnup LOCA tests designed to evaluate the effects of terminal temperature and grid spacers (or cladding restraints) on fuel fragmentation and relocation susceptibility. In addition, BISON fuel performance modeling and out-of-cell benchmark testing were conducted to better interpret the in-cell test results. These tests were developed in collaboration with fuel vendors to ensure that the results provide valuable data for topical reports and support the Nuclear Regulatory Commission’s review.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Community Data Contribution to M.E.T.A. with ATF-relevant Hydrided Zr cladding (Coated and Uncoated)

Since the aftermath of the Fukushima Daiichi loss-of-coolant accident, accident-tolerant fuel (ATF) claddings have been developed to improve the coping times in such events. However, the mechanical performance of ATF cladding is crucial in ensuring that it does not negatively impact the mechanical integrity during all other stages of the nuclear fuel cycle, and the validity of the existing safe operating margins must be verified. However, due to the cladding’s tube geometry and textured anisotropy, determination of apparent mechanical properties under certain deformation paths is challenging. In the uniaxial hoop direction, for instance, the measured mechanical stresses include frictional forces caused by loading mandrels or varying deformation paths in the sample during traditional ring tensile testing. This experimental difficulty is exacerbated by the specimen size. However, addressing these challenges enables irradiation separate-effects investigations in which the materials can be inserted in reactors like the High Flux Isotope Reactor, and reducing the material consumption of commercially irradiated material allows for further post-irradiation examinations. Despite the advantages of reduced-scale mechanical testing, any drawbacks from new specimen geometries must be evaluated, and uncertainties from specimen preparation, setup, and analysis methodologies must be understood.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Microreactor Transportation Emergency Planning Challenges

This report discusses microreactor transportation emergency planning and response challenges. The emphasis of the report is on transporting a microreactor containing irradiated fuel by highway and rail. The report first discusses microreactors in general, including microreactor fuel types and the use of hazardous material in microreactor designs. The report presents a microreactor concept of operations for transport highway, rail, and vessel and presents several transportation regulatory challenges associated with microreactors. Updates contained in this report include the transportation implications of revising the uranium lung absorption category for tri-structural isotropic (TRISO) particle fuel, discussion of exercises and drills including the 2024 Naval Spent Nuclear Fuel Transportation Accident Demonstration with the Shoshone-Bannock Tribes and the State of Idaho, and Tribal and State perspectives on microreactor transportation emergency response planning. Additional microreactor transportation external engagement activities conducted during FY2024 are also discussed. The report discusses the implications of the recent East Palestine, Ohio, rail accident for transportation emergency response planning. Included in this update is a discussion of H.R. 8996, the Railroad Safety Enhancement Act of 2024.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

HERA M&S Exercise Problem Description Report

The Nuclear Energy Agency (NEA) Framework for Irradiation Experiments (FIDES) program includes the High burnup Experiments for Reactivity initiated Accident (HERA) Joint Experimental Program (JEEP). The HERA project is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity Initiated Accident (RIA) conditions. The HERA experiment plan includes analytical integral experiments using test specimens tailored to investigate specific conditions of relevance as well as prototypic integral experiments focused on irradiated fuel from prototypic origin. Modeling & simulation (M&S) is a key component of any experiment program, and the HERA JEEP is coordinating a M&S exercise. The purpose of this document is to provide problem descriptions to support the HERA M&S exercise based on fuel performance modeling. The HERA M&S exercise is expected to evolve into multiple efforts in outyears. This document may be revised and expanded to incorporate those evolutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Benchmarking MELCOR's NAC Package to ABCOVE Test AB7

This report presents analyses of the AB7 ABCOVE sodium spray fire experiment with the MELCOR code. This code simulates the progression of accident events for analysis and auditing purposes of nuclear facilities during accident conditions. Historically, the ABCOVE experiments have contributed to the validation of aerosol physics and related phenomena. Given advancements in sodium-cooled reactor designs, characterization of the sodium spray combustion may further the review and validation of newly incorporated sodium properties and physics packages, namely, the sodium equations of state (EOS) and the sodium combustion (NAC) package within MELCOR. Previously, the AB5 and AB6 experiments were analyzed with and without the NAC package. This work builds on the previous analyses with a demonstration of the current code capabilities of MELCOR with a more mild Na spray and pool fire scenario.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

U.S. Efforts in Support of Examinations at Fukushima Daiichi - September 2024 Meeting Notes

Information obtained from Fukushima Daiichi Nuclear Power Station (Daiichi) is required to inform future Decontamination and Decommissioning (D&D) activities, improving the ability of the Tokyo Electric Power Company Holdings, Incorporated (TEPCO Holdings) to characterize potential hazards and to ensure the safety of workers involved with cleanup activities. This information also has important implications for the safety and operation of U.S. Commercial nuclear power plants. A collaborative U.S. and Japanese effort was initiated in 2014 by the Department of Energy Office of Nuclear Energy to identify Daiichi examination needs and evaluate recent Daiichi examination data to address these needs. This document summarizes information presented at and findings, action items, and recommendations by U.S. and Japanese experts in reactor safety and plant operations during the September 2024 Forensics Effort meeting. Significant safety insights were obtained in several areas: system and component performance, radionuclide surveys and sampling, debris end-state location, combustible gas effects, and plant operations and maintenance. In addition to reducing uncertainties and knowledge gaps in severe accident modeling progression, these insights continue to be used to assess whether additional updates are needed in guidance for severe accident prevention, mitigation, and emergency planning. Furthermore, Daiichi-related activities, such as code modeling improvements and analysis, testing, and new technology deployment efforts, have the potential to offer additional safety and economic benefits to the operating fleet and new light water reactor (LWR) and non-LWR designs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Credible Criticality Safety Margin in the 30B Package with LEU+ UF 6 and Hypothetical Water Ingress

The commercial nuclear industry is pursuing advancements in fuel and reactor design that increase the uranium enrichment above 5 wt. % 235 U. These advancements will necessitate the ability to transport bulk quantities of UF 6 at increased enrichments. Currently, the 30B cylinder is the primary container used by the industry for UF 6 storage and transportation and has a long history of successful shipments. This container can support up to 2,277 kg of UF 6 at a maximum enrichment of 5 wt. % 235 U. Previous evaluations have assessed the potential impact of criticality safety for 30B transport at higher enrichments but assumed moderator intrusion would not require evaluation. Although current regulations allow for the exception of moderator intrusion for UF 6 packages through the design and quality control of the package content, this exception is limited to enrichments up to 5 wt. % 235 U. Thus, an investigation of moderator intrusion into a 30B cylinder should be performed. Moderator intrusion into a 30B cylinder is a unique condition for criticality safety evaluation in transportation because of the violent chemical reactions that occur between UF 6 and H 2 O. The resulting intrusion is strongly dependent on the breach size, breach location, breach interface solid/ullage of the UF 6 content, the UF 6 distribution (which is temperature dependent), and temperature/pressure conditions which are dynamic in accident conditions. Additionally, the complexity of the HF-UO 2 F 2 -H 2 O interface that occurs during the event can influence the potential solubility of uranium in the system, as well as influence the amount of UF 6 reaction with H 2 O. With the operating experience from Orano Federal Services LLC, the UF 6 chemical expertise from Oak Ridge National Laboratory (ORNL), and the criticality safety expertise from ORNL, this paper evaluates the neutronic conditions (i.e., k eff ) that apply the understood chemistry and experimental conditions that occur during moderator intrusion of a breached 30B cylinder under postulated accidents. This report examines the historical evaluations of UF 6 transport and expands these evaluations for the enrichments expected for nuclear industry advancement. These simulations primarily feature a homogeneous mixture of UF 6 and H 2 O as an infinite media system and as a sphere with water reflection given an impurity limit of 0.5 wt. % UF 6 . This report demonstrates that this H/U limit is valid up to 8 wt. % enrichment for both H 2 O and HF as the moderating mixture. The water-reflected homogeneous mixed spheres evaluated demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. Additionally, this report evaluates moderator ingress scenarios of a 30B cylinder. The 30B cylinder simulations vary the cylinder orientation, the mass of UF 6 in the system, the mass of H 2 O in the system, and how much H 2 O has reacted with UF 6 . The moderator ingress of a 30B cylinder was evaluated for the water-reflected homogeneous mixed spheres to demonstrate the amount of safety margin applied by restricting the 30B cylinder impurity limit to 0.5 wt. %. These simulations are standard practice and are independent of the UF 6 -H 2 O reaction. This work also explored more complex simulations that layer the H 2 O over the UF 6 , which incorporates the amount of reacted UF 6 as a separate layer and considers solubility limits of UO 2 F 2 in H 2 O as it is produced. This approach is intended to simulate the postulated event of a large hairline crack occurring on a submerged cylinder. Finally, this work considers the potential of H 2 O mixing into the UF 6 solid heterogeneously using a sponge-like model. The water ingress into the UF 6 is treated as random size spheres, and the reaction products form into a layer over the spheres.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Success Path Method: Conformance with Safety Management Systems

All industrial facilities deal with safety hazards such as equipment failures, chemical and toxic releases, and fires and explosions just to name a few. A disciplined framework for managing the integrity of operating systems and processes that handle hazardous substances by applying good design principles, engineering, and operating practices is called process safety. The goal of such framewoks is to prevent the release of energy or material that could cause harm to people or damage to equipment and/or environment. Process safety covers all aspects of facility operation also including design, maintenance, and human and organizational factors that could possibly have effect on process safety. As it will be seen from the discussion below, process safety is just one of the pieces of the much bigger matter, a safety culture. Many industries have long recognized the importance of safety culture in their day-to-day operation. Although the definition of safety culture can slightly differ from organization to organization, in general, a safety culture is how things are done to demonstrate a commitment to safety by everyone involved. An organization acquires safety culture over time as the product of individual and group values, actions, and behaviors toward overall safety. It is important to note that safety culture should not be viewed as some static state that an organization wants to reach. It is more like a constantly evolving level of “how things are done when nobody is watching.” Safety culture is an inherent characteristic of an organization, it is always present, but the level can range on a continuum from undesirable to desirable or more commonly used, from negative to positive. An example for an undesirable, negative safety culture would be a company in which accidents resulting in harm (physical and/or emotional) of its employees, equipment or surrounding environment and community occur frequently. At the other end of the spectrum would be a company in which such accidents are rare or do not occur at all (desired or positive safety culture). Every company/industry exists somewhere within this spectrum.

42 ENGINEERING↗
DOE OSTI2025

Chemical Process Safety at TRISO-Based, Metal-Based, and Salt-Based Fuel Fabrication Facilities: Technical Assessment and Guidance Assessment

As part of efforts to prepare for potential and ongoing safety reviews for licensing of advanced non-light-water reactor fuel cycles, the U.S. Nuclear Regulatory Commission (NRC) tasked Pacific Northwest National Laboratory to prepare an assessment on the state of knowledge of potential chemical processes at fuel cycle facilities supporting the front end of these fuel cycles, and to assess the associated regulatory guidance. This report provides a technical assessment of chemical process safety considerations to support NRC licensing reviews of fabrication processes for tri-structural isotropic (TRISO) based, metallic-based, and salt-based fuels. The assessments involved collecting publicly available information on the fuel fabrication processes to (i) identify the operational process steps, characteristics and chemicals involved, (ii) identify the physical safety considerations and health safety considerations during licensing reviews of the various process steps, and (iii) collect information to support assessments of severity of accidents and potential mitigative measures to be implemented. The assessment provides a foundational basis on chemical process safety considerations for advanced fuel fabrication activities, although it is recognized that licensing reviews may necessitate design-specific considerations. The specific conditions under which chemical hazards emerge will require process-specific considerations, highlighting the importance of process-informed interpretation. The assessment also determined that exposure guidelines and limits to assess the consequences of acute exposures are limited for some chemicals, although alternative limits and supplementary information from databases or safety data sheets provide sufficient information to evaluate consequences of acute exposures. In addition, it was identified that metallic and salt fuel fabrication processes may involve beryllium, which is an exposure hazard. The regulatory framework for the licensing of advanced fuel cycle facilities, per 10 CFR Part 70 Domestic Licensing of Special Nuclear Material, is deemed robust and flexible to address the chemical safety considerations in this report. A review was conducted on various regulatory guidance and technical basis documents. This included reviewing NUREG-1520, Revision 2, Standard Review Plan for Fuel Cycle Facilities License Applications – Final Report and the process descriptions in Appendix A of NUREG/CR-6410, Nuclear Fuel Cycle Facility Accident Analysis Handbook, to address advanced fuel types. As new fuels will involve process-specific chemical uses, process-specific considerations are provided in this report. Additionally, it is noted that the U.S. Department of Energy protective action criteria database includes Temporary Emergency Exposure Limits (TEELs) for process-specific chemicals. This report provides technical information to support chemical safety assessments of new advanced fuel cycle facilities and identifies technical and safety information to support licensing reviews. No regulatory barriers were identified for the licensing of advanced fuel cycle facilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Basis for Dose and Reactor Safety Design Criteria for Army Regulation AR 50–7 and DA Pamphlet

This report describes the basis used to develop the radiological dose acceptance and design criteria contained in the draft updates to Army Regulation 50–7 (AR 50–7) Army Reactor Program and its accompanying draft Department of the Army (DA) Pamphlet (PAM), Army Reactor Program Procedures. These criteria will apply to Army nuclear reactors that fall under AR 50–7 and its accompanying DA PAM and ensure alignment with the overall objectives of the Army Reactor Program. The development basis for the radiological dose and design criteria supports a modern, technology-neutral, risk-informed, and performance-based approach to Army regulation of reactors and the demonstration of “adequate protection of the public.” To establish these criteria that support the Army’s unique operational requirements, multiple well-known and well-established standards and their supporting documentation were reviewed to ensure consistency with existing regulatory safety levels, including guidance from U.S. and international sources. These include the U.S. Nuclear Regulatory Commission’s (NRC’s) regulations and policy, the Canadian Nuclear Safety Commission’s (CNSC’s) regulatory documents, the International Atomic Energy Agency’s (IAEA’s) safety standards, as well as industry input that is tailored specifically to advanced microreactors. This report walks through the key definitions and associated references used for these criteria, which are outlined in Section 2.0. Based on these definitions, the dose acceptance criteria were established for various receptors for routine reactor operations (Section 3.2), design basis accidents (Section 3.3), and beyond design basis accidents (Section 3.4). Comparisons of multiple national and international dose limits are provided in these sections. Lastly, Section 4.0 outlines the reactor safety design criteria contained in the draft DA PAM and their associated bases.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Gas-Cooled High-Temperature Pebble-Bed Reactor Reference Plant Model Updates

This work presents the latest improvements to, and investigations performed with, the pebble-bed high-temperature gas-cooled reactor (PB-HTGR) reference plant models for the United States Nuclear Regulatory Commission. These models serve as the foundation for the future development of detailed design evaluation models based on license applications. The reference plant models have been developed with the Comprehensive Reactor Analysis Bundle, or BlueCRAB, which is the code suite proposed for non-light-water reactor systems safety analysis. It incorporates various simulation tools developed by the Nuclear Energy Advanced Modeling and Simulation program, including the Griffin code for reactor physics, the Pronghorn and SAM codes for core thermal fluids, the BISON code for solid conduction and fuel performance, and the SAM code for system analysis. The primary objective of the work that was performed was to assess BlueCRAB’s level of readiness for modeling a PB-HTGR. To do so, we first developed numerical models in BlueCRAB that include the key physics for this technology to ensure an adequate level of fidelity for modeling PB-HTGR core performance and for performing multiphysics simulations for equilibrium core conditions and different accident scenarios. Then we simulated transient scenarios, including depressurized and pressurized loss of forced cooling accidents, over-cooling, and control rod withdrawal events with delayed and prompt supercritical reactivity insertions. The analysis in this report includes comparisons of the 2D thermal fluid porous media models in Pronghorn and SAM, and comparisons of coupled SAM/Griffin/SAM and coupled Pronghorn/Griffin for depressurized and pressurized loss of forced cooling, over-cooling, and control rod withdrawal events. In addition, we compare 3D, 2D, and 0D/PKE neutronic models for the two control rod withdrawal scenarios with coupled Pronghorn/Griffin. The comparisons show that the BlueCRAB models lead to physically intuitive solutions for the scenarios examined. The changes in the various scalar and vector fields, such as neutron flux, power, temperature, density, pressure, and velocity, are within the expected ranges, and their distributions can be explained by the system response of the transients and the geometric and material variations. Several comparisons suggest that the porous media models in Pronghorn and SAM can lead to similar solutions, even though they are based on different methodologies. This work further highlights the need for flexible tools with various levels of fidelity to cover the breadth and depth of needs that may arise in future technical evaluations of the PB-HTGR. We believe that the BlueCRAB capabilities will be a significant asset for confirmatory analyses in order to resolve important safety questions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗