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At least 1,081 records · Page 60
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Effects of ZrN coating and heat treatment on U-Mo dispersion fuel systems under irradiation

The stability of U-Mo fuel particles embedded in an Al matrix under irradiation can be enhanced through ZrN coatings and/or heat treatment. Here, the present study investigates the irradiation behavior of fuel plates containing U-Mo fuel particles fabricated under various heat-treatment conditions and ZrN coating thicknesses. Different fission densities were also applied to each fuel plate to evaluate the effects of these variables. Results indicate that higher fission densities lead to more grain recrystallization and high burnup structure (HBS) development in the fuel particles. Heat treatment was found to mitigate the accumulation of fission gas bubbles in fuel particles at low fission densities by coarsening their grains. Fuel particles with ZrN coatings of a 1.2 μm thickness or above exhibited reduced formation of U-Mo/Al interaction layers, suggesting the existence of a critical ZrN coating thickness that minimizes the development of these layers. Fission gas bubbles were predominantly observed at grain boundaries of U-Mo fuel particles irradiated at low fission densities. Subgrain boundaries, which appeared to originate from the original grain boundaries containing fission gas bubbles or HBSs, were also observed, indicating the early stage of HBS propagation in the fuel particles.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Application of Molecular Catalysts for the Oxygen Reduction Reaction in Alkaline Fuel Cells

Here, the development of precious group metal-free (PGM-free) catalysts for the oxygen reduction reaction is considered as the main thrust for the cost reduction of fuel cell technologies and their mass production. Within the PGM-free category, molecular catalysts offer an advantage over other heat-treated PGM-free catalysts owing to their well-defined structure, which enables further design of more active, selective, and durable catalysts. Even though non-heat-treated molecular catalysts with exceptional performance have been reported in the past, they were rarely tested in a fuel cell. Herein, we report on a molecular catalyst under alkaline conditions: fluorinated iron phthalocyanine (FeFPc) supported on cheap and commercially available high-surface area carbon-BP2000 (FeFPc@BP2000). It exhibits the highest activity ever reported for molecular catalysts under alkaline conditions in half-cells and fuel cells.

25 ENERGY STORAGE↗
DOE OSTI2013

Progress on Plant-Level Modeling and Simulation of Used Nuclear Fuel Dissolution: Startup Condition with Quasi-Developed Flow and Varying Acid Concentration

An update to a previous report is presented here describing the startup condition of a continuous dissolver. The need for a careful implementation of a startup condition is twofold. First, by varing the used fuel composition and various other parameters of the model, it becomes difficult to solve the governing equations accurately or to obtain a converged solution for an arbitrary initial condition. This is evident when coupling the dissolver module with a third-party user simulator. The second impetus for improving the modeling of the startup process is that an improved realism is obtained and the model is more complete and useful for accounting material balances rigorously. The aforementioned previous report on this dissolver module is a required reading for the understanding of this update.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Complete Status Report Documenting the Processing, Microstructure and High-Temperature Mechanical Properties of the Nanostructured Ferritic Alloy OFRAC

This report provides a summary of the new nanostructured ferritic alloy, OFRAC (Oak Ridge Fast Reactor Advanced Fuel Cladding), that was developed for the advanced sodium fast reactor fuel cladding and core internals. The composition of OFRAC, which is Fe-12Cr-1Mo-03.Ti-0.3Nb + 0.3Y2O3 (wt. %), was chosen to achieve high temperature strength as well as high temperature creep performance and swelling resistance requirements that past structural materials such as austenitic stainless steels and ferritic/martensitic steels, i.e. HT-9, are not suitable as SFR cladding materials. The mechanical alloying procedures were described for producing OFRAC and the detailed characterization studies revealed uniformly distributed, high concentration of Y-Ti-O enriched oxide particles, or nanoclusters. The Nb addition that was intended to sequester interstitial C and N that are commonly introduced as contamination during ball milling of powder was found to be very effective in forming coarse size Nb carbonitrides along with excess Ti that removes these interstitial elements from the bcc-Fe matrix of OFRAC. The tensile tests on the OFRAC in the extruded condition showed excellent balance in strength and ductility from room temperature to 800°C. Both the microstructure and tensile properties were not significantly degraded after annealing the extruded OFRAC samples for 8 h at 1050°C and 1150°C. The latter temperature is ~76% of the melting point of OFRAC. These results demonstrated the stability of the microstructure and tensile properties to extremely high temperatures. The thermal creep resistance at elevated temperatures and stresses was investigated using strain rate jump testing. The results showed that OFRAC possessed high stress exponents at high stresses and temperatures of 550°C, 600°C, 700ºC and 800°C. This thermal dependence of minimum creep rate indicates that OFRAC will show quite low creep rates at higher temperatures and stresses compared to HT-9 and stainless steels. These significant improvements indicate that the NFA OFRAC alloy is a compelling and viable choice for sodium fast reactor fuel cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Impacts of LEU+ and ATF on Fresh Fuel Storage Criticality Safety

The use of increased fuel enrichment, which is still in the realm of low-enriched uranium (LEU) fuel, has been of interest to commercial light water reactor operators as part of the next iteration in fuel cycle technological advances and research and development. Using increased enrichment fuel, or high-assay LEU (HALEU), in power plants has clear benefits for being able to load cores with additional power-producing fuel. Although HALEU enrichments can range up to 20%, the more guarded approach of investigating enrichments above current fuels within 10% enrichment is referred to as LEU plus (LEU+) to reflect the less drastic change in operating conditions and requirements and similarity to current fuel cycles. Of additional interest and increasing maturity is the incorporation of accident-tolerant fuel (ATF) concepts, which are also applicable to the current fleet. This class of technologies involves changes such as cladding (e.g., chromium coating or FeCrAl) and fuel composition (e.g., chromia dopant) alterations to demonstrate improved fuel performance under accident scenarios. The ability to properly store fuel before and after residence time in the reactor is crucial to plant operation. Typically, this is done in either a new fuel vault (NFV) or spent fuel pool (SFP). Storing, loading, and unloading dozens of fuel assemblies within the same general area provides opportunities for obvious criticality concerns. These concerns are addressed with regulations to the subcriticality margin that the NFV and SFP must maintain in certain conditions. Adopting LEU+ fuel results in inherent reactivity increases, which are extremely relevant for safe fuel storage. Therefore, a clear understanding of the effects of LEU+ fuel and ATF on criticality safety margins to regulatory limits is required, as well as an understanding of the degree of absorber crediting under normal and accident conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Second Target Station Project: STS Cross-Directorate Workshop on Hydrogen Fuel

Hydrogen, particularly green hydrogen produced through electrolysis using renewable energy, is poised to play a critical role in decarbonizing the global economy. Its ability to address the intermittency of renewable energy sources and decarbonize hard-to-electrify sectors positions it as a vital component of a sustainable energy future. Interest in hydrogen as a clean energy carrier is in a decade of unprecedented growth, with dozens of countries having released national hydrogen strategies contributing to a global hydrogen economy. In 2021, the Department of Energy (DOE) announced the Hydrogen Shot, the first of the Energy Earthshot Initiatives, which aims to lower the cost of clean hydrogen to $\$$1/kg by 2031. This initiative was followed by a considerable increase in funding for hydrogen technologies through the Bipartisan Infrastructure Law (BIL), with $\$$18 billion recently announced for regional demonstration projects (Hydrogen Hubs) and another $\$$11.5 billion aimed at research and development of electrolyzers, fuel cells, manufacturing, and recycling. The 2023 U.S. National Clean Hydrogen Strategy and Roadmap identifies the cost of clean hydrogen as a critical challenge for achieving economic scale. This includes the cost of hydrogen production by electrolysis, delivery and dispensing, onboard storage, and end-use technologies like fuel cells. Fundamental research and development into catalysts, component architectures, and material durability are critical to lower the costs of these vital technologies aimed at achieving a net-zero carbon emission economy by 2050.

08 HYDROGEN↗
DOE OSTI2024

Elucidating the Radiation-Induced Redox Chemistry of Plutonium Under Used Nuclear Fuel Reprocessing Conditions

Plutonium plays a critical role in the development of sustainable nuclear fuel cycles, and yet, our fundamental understanding of this element’s inherent radiation-induced redox chemistry and associated impacts on nuclear fuel cycle technologies is limited. Unanticipated changes in oxidation state distribution can influence the speciation and transport of plutonium in a given process. Control of these parameters is especially important for used nuclear fuel reprocessing technologies, wherein the separation and recovery of plutonium is typically achieved by the selective formation, maintenance, and complexation of specific oxidation states. Furthermore, plutonium’s inherent radiation-induced redox chemistry has the capacity to influence the radiolytic behavior of its complexes, the longevity of which are critical in the design of efficient and cost-effective advanced reprocessing technologies. These radiation-induced processes are unavoidable under fuel cycle conditions owing to the inherency of ionizing radiation fields to the decay of plutonium’s isotopes and to the various other radioisotopes generated by nuclear fission and neutron-capture process and the subsequent radioactive decay of their products. As such, mechanistically understanding the response of plutonium’s multiple oxidation states to multi-component ionizing radiation fields is essential for predicting the behavior of this critical element under used nuclear fuel reprocessing conditions. Here, through a combination of time-resolved (electron pulse) and steady-state (alpha and gamma) irradiation experiments complemented by quantitative, multiscale modeling calculations, we present advances in our understanding of radiation-induced plutonium redox chemistry!

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Flow field and scalar measurements in a series of turbulent partially-premixed dimethyl ether/air jet flames

Here, we present a series of benchmark flames consisting of six partially-premixed piloted dimethyl ether (DME)/air jet flames. These flames provide an opportunity to understand turbulence-flame interactions for oxygenated fuels and to develop predictive models for these interactions using a canonical burner geometry. The development of accurate models for DME/air flames would establish a foundation for studies of more complex oxygenated fuels. The flames are stabilized on a piloted jet burner similar to that of the partially-premixed methane/air jet flames that have been studied extensively within the context of the TNF Workshop. This series of six jet flames spans jet exit Reynolds numbers, Re D , from 29,300 to 73,300 and stoichiometric mixture fractions, ξ st , from 0.35 to 0.60. Flame conditions range from very low probability of localized extinction to a high probability of localized extinction and subsequent re-ignition. Measurements in the flames are compared at downstream locations from 5 to 25 diameters above the nozzle exit. Mean and fluctuating velocity components are measured using stereo particle image velocimetry (SPIV). Simultaneous laser-induced fluorescence (LIF) imaging of OH and CH 2 O provides insights into the distribution of these intermediate species in partially-premixed DME/air flames. OH LIF imaging is also combined with SPIV to investigate the strain rate field across the reaction zone.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2014

Fuel-Flexible Combustion System for Refinery and Chemical Plant Process Heaters

This project culminated in the demonstration of a full-scale industrial burner which allows a broad range of “opportunity” gaseous fuels to be cost-effectively and efficiently utilized while generating minimal emissions of criteria air pollutants. The burner is capable of maintaining a stable flame when the fuel composition changes rapidly. This enhanced stability will contribute significantly to improving the safety and reliability of burner operation in manufacturing sites. Process heating in the refining and chemicals sectors is the primary application for this burner. The refining and chemical sectors account for more than 40% of total industrial natural gas use. Prior to the completion of this project, an enabling technology did not exist that would allow these energy-intensive industries to take full advantage of opportunity fuels and thereby reduce their natural gas consumption. Opportunity gaseous fuels include biogas (from animal and agricultural wastes, wastewater plants, and landfills) as well as syngas (from the gasification of biomass, municipal solid wastes, construction wastes, and refinery residuals). The primary challenge to using gaseous opportunity fuels is that their composition and combustion performance differ significantly from those of conventional fuels such as natural gas and refinery fuel gas. An effective fuel-flexible burner must accept fuels that range widely in quality and change in composition over time, often rapidly. In Phase 1 of this project, the team applied computational fluid dynamics analysis to optimize the prototype burner’s aerodynamic, combustion, heat transfer, and emissions performance. In Phase 2, full-scale testing and refinement of two prototype burners were conducted in test furnaces at Zeeco’s offices in Broken Arrow, OK. These tests demonstrated that the full range of conventional and opportunity fuels could be utilized by the project’s burner while achieving robust flame stability and very low levels of air pollutant emissions. In Phase 3, the team retrofitted three fuel-flexible burners into a fired heater at a Shell plant and demonstrated the project’s technology over a 6-month period. The project burners performed well during this period. They remain in commercial service at the Shell plant. Through this work, an improved understanding of flame stabilization mechanisms was gained. Also, methods for accommodating a wide range of fuel compositions were developed. This knowledge facilitated the commercialization of a new generation of burners that are suitable for the fuels of the future.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2025

Mesoscale Modeling for Restructuring and Fragmentation in High Burnup UO 2

This report summarizes the mesoscale modeling conducted in fiscal year 2025 under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, focusing on the microstructural evolution and restructuring in high burnup UO 2 nuclear fuel and its impact on fuel fragmentation. We developed a pioneering phase-field model to simulate restructuring behavior across different regions of high burnup fuel, including the dark zone and rim region. A grand-potential-based phase-field model is employed to concurrently evaluate subgrain formation and the growth of fission gas bubbles within the fuel. An energy-based subgrain formation criterion was introduced to simulate the restructuring process. The effects of temperature and burnup rate were studied to capture how each of these parameters influences the characteristics of the restructured fuel. Subgrain formation was observed to initiate around existing fission gas bubbles and proceed toward triple junctions, grain boundaries, and grain interiors. Under a given subgrain formation rate, the rate of restructuring increases with rising fuel temperature. The restructuring occurs faster with higher burnup rate. A restructuring bias was observed within the microstructure, due to the variation in defect accumulation when comparing different grains. Microstructures corresponding to the dark zone and rim region can be obtained by parameterizing the model with the appropriate defect production rate, as determined based on the burnup rate and temperature. The predicted microstructures are consistent with experimental observations of the restructured regions. Based on the mesoscale simulations, a mechanistic model for restructuring and grain size evolution was implemented in BISON. Thus, this work provides a first-of-its-kind restructuring model for different regions of high-burnup fuel to BISON. It was shown that the model predicts appropriate grain size evolution along fuel radius as those observed in experiments. This model enables BISON to predict the effect of restructuring on the fission gas release. Finally, the phase-field fracture simulation with dark zone specific microstructures were presented to provide the fragmentation criteria for the dark zone. This work introduces a first-of-its-kind restructuring model for high burnup fuel, significantly enhancing BISON's predictive capabilities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Impact of Using Cross Sections Versus Recipes in Transition Fuel Cycle Analysis using ORION

ORION, a nuclear fuel cycle simulator developed at the UK National Nuclear Laboratory (NNL), is designed to simulate the operations of general nuclear fuel cycles, including once-through, limited recycle, and continuous recycle fuel cycle options. ORION can simulate the full range of nuclear-related facilities e.g. interim and long-term storage locations, fabrication and enrichment plants, reprocessing facilities, and reactors. It can track over 2,500 nuclides and can model decay and in-reactor irradiation. Based on ORION's ability to track the full isotopic mass flow between fuel cycle facilities (with or without decay), it is possible to model all of the necessary nuclide production and destruction routes. This is essential when calculating accurate spent nuclear fuel (SNF) compositions and associated metrics (e.g., decay heat and radioactivity). When modeling the transition from the current once through US fuel cycle to an alternative fuel cycle, it is important to accurately model the US pressurized water reactor (PWR) discharged SNF composition while minimizing the computational time. This paper discusses two rapid methods used to model the US PWR discharged SNF-fixed recipes versus cross sections-and it discusses the differences between the results from these two methods in relation to fuel cycle transition analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

USABILITY EXTENSION OF THE ŠKODA VPVR/M CASK FOR TRANSPORTING IRRADIATED FUEL ASSEMBLIES

New abstract from the final version being submitted now: This paper discusses the evolution and enhanced usability of the ŠKODA VPVR/M cask for the transport of irradiated fuel assemblies, particularly within the context of the demand for the delivery of newly appearing irradiated HEU fuel types for which the cask did not yet have a license to transport. Over time, the cask’s internal basket construction has demonstrated notable adaptability to accommodate various exotic HEU fuel types from research reactors of differing origins. The paper outlines sever-al custom internal baskets developed for specific fuel types, including those from Belarus and Serbia, as well as from Georgia, Uzbekistan, and MNSR cores, as well as a recently designed basket for MTR and TRIGA assemblies. The findings high-light the high flexibility and adaptability of the cask, supported by successful rede-signs and licensing efforts, underscoring its value for the safe and secure transport of nuclear material. Old abstract from the draft version that was already approved: The Russian Research Reactor Fuel Return (RRRFR) program, since its inception, has continuously used the ŠKODA VPVR/M Cask fleet designed for the repatriation of irradiated highly enriched uranium (HEU) fuel. As the program progressed (from shut-down and a quasi-abandoned reactor, and/or as it began to include fuels of Chinese and US origin), new challenges emerged for the transport Cask. These were fuel types that had not yet been licensed for the Cask. Although these requirements did not arise during the design of the basic ŠKODA VPVR/M Cask, as revealed by the retrospective analyses, the Internal Basket of the ŠKODA VPVR/M Cask gives a high degree of flexibility to accommodate additional fuel types. This paper provides a brief overview of the ŠKODA VPVR/M Cask, which holds a B(U) type license, and introduces the different types of Internal Baskets that have already been licensed to transport so-called exotic irradiated HEU fuel types, in addition to the original license. The paper presents a new Internal Basket design for accommodating MTR-type and TRIGA-type irradiated HEU fuel assemblies. This includes a detailed presentation of the design basis and the new MTR-TRIGA Internal Basket, as well as the licensing matters of the package under the name ŠKODA MTR-TRIGA Cask, and the conformity test (dry- and wet-run) operations made to verify compliance with the new Internal Basket. Then, as a summary, the usage record for the Cask fleet is presented, and finally, the paper concludes with the consolidated experiences gained during the utilization of the ŠKODA VPVR/M Cask fleet, emphasizing the high degree of Cask flexibility ensured by the Internal Basket’s construction.

42 - ENGINEERING↗
DOE OSTI2016

Diffusivities of Ag, Cs, Sr, and Kr in TRISO fuel particles and graphite

Tri-structural isotropic (TRISO) coated particles have been developed and studied since the late 1950s when the concept of coated particles was invented by Roy Huddle of the United Kingdom Atomic Energy Authority. Several decades of work by half a dozen countries on fission product transport in TRISO fuel through numerous irradiation and heating experiments have led to several recommendations of transport data and to the adoption of various sets of diffusion coefficients. In 1997, the International Atomic Energy Agency (IAEA) gathered all these historical results and issued a technical document (TECDOC-978 [IAEA]) that summarizes these sets of recommended diffusion coefficients. Table 1 shows the reference literature articles for the diffusivities that have historically been recommended by the American and German TRISO fuel development programs and that are summarized in the IAEA report (see section 7 for full references of these articles).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of the Licensing Procedure and Regulatory Framework for the Spent Fuel Storage facility in Korea - 16423

The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the future national policy for spent fuel management of Korea has been established, taking into consideration the national/international trends on policy and technology development. Regardless of the recommendation report of PECOS, Spent fuel storage casks are necessary in Korea. Therefore, various types of spent fuel storage casks are now being developed by Korea vendors. For example, Concrete storage casks and dual purpose metal casks are now under development by the KORAD (Korea Radioactive waste agency) for future use in the interim storage facility. Unfortunately, there is no independent licensing procedure for spent fuel storage casks in Nuclear Safety Act (NSA) because is the casks are considered an integral part of the interim storage facility safety equipment. Furthermore, the definition of radioactive waste in the NSA was changed recently. Now, spent fuel is separated from the definition of 'radioactive waste'. Spent fuel can only become radioactive waste by the decision through the meeting of the relative ministers of Korea government. For that reason, the licensing procedure for interim spent fuel storage facilities was separated from the current 'disposal facility, etc.' in accordance with the article 63 of NSA. The aim of this study is to develop the licensing procedure for the storage cask, and also to develop a revised draft of interim storage facility procedure within the NSA. Once this independent licensing procedure on the spent fuel storage cask is introduced to the NSA, it is expected that developers or operators can develop and commercialize the storage cask in Korea. In this study, an independent licensing procedure for the spent fuel storage facility was developed including the storage cask licensing procedure for the NSA. The procedure was composed of the permit for construction and operation of interim storage facility, criteria for permit, inspection criteria, periodic safety review, and decommissioning criteria, etc. The detailed contents of these requirements were also developed to introduce the regulatory framework for the interim storage facility. Particularly, periodic safety review and decommissioning criteria were developed in accordance with the recommendation of IRRS (International Regulatory Review Service) of IAEA. The developed licensing procedure for spent fuel storage facility will be introduced in NSA in response to the revised definition of NSA and the recommendation report of PECOS. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Development of low-Cr ODS FeCrAl alloys for accident-tolerant fuel cladding

Low-Cr oxide dispersion strengthened (ODS) FeCrAl alloys were developed as accident tolerant fuel cladding because of their excellent oxidation resistance at very high temperature, high strength and improved radiation tolerance. Fe-12Cr-5Al wt.% gas atomized powder was ball milled with Y 2 O 3 +FeO, Y 2 O 3 +ZrO 2 or Y 2 O 3 +TiO 2 , and the resulting powders were extruded at 950°C. The resulting fine grain structure, particularly for the Ti and Zr containing alloys, led to very high strength but limited ductility. Comparison with variants of commercial PM2000 (Fe-20Cr-5Al) highlighted the significant impact of the powder consolidation step on the alloy grain size and, therefore, on the alloy mechanical properties at T < 500 °C. These low-Cr compositions exhibited good oxidation resistance at 1400 °C in air and steam for 4 h but could not form a protective alumina scale at 1450 °C, similar to observations for fine grained PM2000 alloys. As a result, the effect of alloy grain size, Zr and Ti additions, and impurities on the alloy mechanical and oxidation behaviors are discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Scientific Basis For Ultra-High Burn-Up Nuclear Fuels

A key issue on the path to nuclear energy becoming an essential component of the U.S. clean energy strategy is complete burn of the nuclear fuel. We propose research that will enable high-burn-up fuels by establishing the basic science for development and qualification of advanced nuclear fuel that will couple modern computational materials modeling, fabrication, and characterization capabilities with targeted performance-testing experiments using ion-beam facilities. This work will establish the scientific foundation and guide selection of the optimum fuel type for advanced reactor concepts. We expect to experimentally quantify the phase stability and kinetics of phase transformations, inter-diffusion, microstructural evolution, micromechanical properties, and the influence of severe radiation environments on fuel performance and, by the end of the project, develop a validated model for advanced nuclear energy materials under extreme conditions of radiation, temperature, and evolving chemistry. We will provide a science-based path forward to an optimized inert matrix fuel, while contributing to the development of a validated nuclear fuel database. We expect this project will further establish LLNL’s credibility in the nuclear energy community.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Detailed Analysis of RUS Insertion Experiment and Scoping Studies for Performing Next Experiment using an Enriched Fuel Sample

Microstructure evolution due to irradiation in a nuclear reactor can have a dramatic effect on material properties. A better understanding of this evolution is necessary for developing improved nuclear fuels and materials. The ability to measure such changes in real time is extremely challenging due to high temperatures, high radiation fields, and limited access of the reactor environment. Through carefully designed experiments, measurement of elastic properties can be tied directly to microstructure. We present an instrument that has been developed to monitor in-pile changes in grain microstructure. The measurement approach involves optically exciting and detecting flexural waves in a thin cantilever beam. An instrument capsule based on this technique was fabricated and underwent an irradiation test in the TREAT reactor in May 2019. Analyses of the test results are presented in this report. Scoping studies on the expected impact of radiation on similar tests are also presented.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Characterization of MC&A for the Molten Salt Fuel Cycle

Advanced reactor developers are exploring diverse reactor designs, including molten salt reactors (MSRs). These advanced reactors are considered for wider applications and a range of deployment locations, including supporting the integration of renewable energy sources in the grid. There are three main types of MSRs: (1) reactors in which the fuel salt freely circulates within the core; (2) reactors with the fuel salt contained within vented fuel tubes; and (3) reactors that use molten salt solely as a coolant, with the fuel in a separate, solid form. In this document, the term MSR refers specifically to the first two types, which use fuel salt—special nuclear material (enriched uranium, plutonium, and 233 U) in chloride or fluoride form mixed with chloride- or fluoride-based carrier salt in a peritectic mixture—as the primary medium for fission. The composition of fuel salt, both at startup and for makeup or refueling, varies depending on the MSR design and the chosen fuel cycle approach, which can be either once-through or closed. For MSRs, a variety of fuel cycle approaches (e.g., U, U–Pu, U–Pu–TRU, U–Th, U–Pu–Th) are being considered. Fuel in MSRs is much different than traditional solid fuel, including its preparation. The uniqueness warrants investigation into characterizing fuel preparation processes, known as fuel salt synthesis . This effort characterized major fuel preparation and synthesis processes, identifying temperature, equipment, and environmental requirements for uranium-, plutonium-, and thorium-based fuel preparation and synthesis. Because MSR fuel salt synthesis facilities handle special nuclear material in loose, bulk form, a material control and accounting plan will be required for licensing from the US Nuclear Regulatory Commission or under the US Department of Energy authorization. This effort serves as a foundation to investigate material control and accounting approaches for synthesis facilities, including determining measurement points and techniques. Because several MSR developers are planning demonstration facilities in the coming years, this effort will support stakeholders with preparing or reviewing material control and accounting plans for providing assurance that all special nuclear material is accounted for at fuel salt synthesis facilities. This report was produced for Materials Protection, Accounting, and Control Technologies (MPACT) program under the US Department of Energy (DOE), Office of Nuclear Energy, Nuclear Fuel Cycle and Supply Chain.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗