Control of Grain Properties and Slug Geometry for High-Performance Metallic Fuel
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11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Mechanical Properties of Irradiated U-10 wt. %Mo Alloy Degraded by Porosity Development
In this study, a plate-type nuclear fuel consisting of a solid monolithic foil of U-10 wt. %Mo is under development for use in the United States' high-performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate postirradiation mechanical properties of the commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo samples to generate the fuel material properties, including the modulus of elasticity and the bending strength. Although the material behaves in a brittle manner due to the accumulated porosity, a general trend of strength and modulus reduction was found as fission density increases. The data produced was evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development
A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States’ high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo fuel and the data produced includes bending strength and Young’s modulus. Although the material behaves in a brittle manner due to the developed porosity, a general trend of strength and modulus reduction are found as fission density increases. The data produced is evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Microstructural and Micro-Chemical Evolutions in the Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles
AGR-1 and AGR-2 TRISO fuel particles were fabricated with slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. To correlate those differences with the fuel kernel responses to neutron irradiations in terms of irradiated fuel microstructure, fission products chemical and physical states, and fission gas bubble evolutions, extensive microstructural and analytical characterizations were conducted. The studies used a state of art transmission electron microscopy (TEM) equipped with Energy-dispersive X-ray spectroscopy (EDS) of four silicon solid-state detectors which have super sensitivity and fast speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen to represent a representative irradiation conditions with a fuel burnup within the range from 10.8 to 18.6% FIMA, and the time-average volume-average temperatures vary from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% for the AGR1 fuel kernels and 14.03 wt.% for the AGR-2 fuel kernels. The TEM results show that there were significant microstructural reconstructions in the irradiated fuel kernels for both the AGR-1 and AGR-2 fuels. There are four major phases including fuel matrix of UO2 and UC, U2RuC2, and UMoC2 in the irradiated AGR2 fuel kernel. Zr and Nb form solid solution in the UC phase. UMoC2 phase often shows a detectable concentration of Tc. Pd was found to mainly locate in the buffer layer or to be associated with fission gas bubble within the UMoC2 phase. The EDS maps qualitatively show that the rare-earth fission products (Nb, et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there is a high number of rod-shape precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The difference of microstructural and micro-chemical evolutions in irradiated fuel kernels between the AGR-1 and AGR-2 TRISO fuel particle may result from a combined factor of irradiation temperature, fuel geometry and chemical composition. However, the irradiation temperature probably play a more deterministic role. Limited electron energy loss spectroscopy (EELS) characterizations on the AGR2 fuel kernel show that there is nearly no carbon in the UO2 phase while a small fraction of oxygen was detected in the UC/UMoC2 phase.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Atmospheric plasma spraying to fabricate metal‐supported solid oxide fuel cells with open‐channel porous metal support
Abstract Metal‐supported solid oxide fuel cells (MS‐SOFCs) have been fabricated by applying phase‐inversion tape‐casting and atmospheric plasma spraying (APS). The effect of the binder amount of the phase‐inversion slurries on the microstructure development of the 430L stainless steel metal support was investigated. The pore structures, the viscosity of the slurry, porosity and permeability of the as‐prepared metal supports are significantly influenced by the amount of the binder. NiO–scandia‐stabilized zirconia (ScSZ) anode, ScSZ electrolyte and La 0.6 Sr 0.4 Co 0.2 Fe 0.8 O 3−δ (LSCF) cathode layers were consecutively deposited on the metal support with an ideal microstructure by APS process. The effect of plasma power of the APS on the microstructure of the electrolyte and cathode was investigated. A dense electrolyte layer and a porous cathode layer were successfully obtained at 40 and 6 kW of the APS plasma power, respectively. MS‐SOFCs, with a cell configuration of 430L/Ni‐ScSZ/ScSZ/LSCF, achieved a maximum cell power density of 1079 mW cm −2 at 700°C using humidified H 2 as fuel and ambient air as oxidant. The corresponding ohmic resistance and total resistance of MS‐SOFCs was 0.14 and 0.32 Ω cm 2 , respectively. This work demonstrates the feasibility of fabricating high‐performance MS‐SOFCs with economical and scalable techniques.
36 MATERIALS SCIENCE↗
Robust measurement of microbial reduction of graphene oxide nanoparticles using image analysis
ABSTRACT Shewanella oneidensis ( S. oneidensis ) has the capacity to reduce electron acceptors within a medium and is thus used frequently in microbial fuel generation, pollutant breakdown, and nanoparticle fabrication. Microbial fuel setups, however, often require costly or labor-intensive components, thus making optimization of their performance onerous. For rapid optimization of setup conditions, a model reduction assay can be employed to allow simultaneous, large-scale experiments at lower cost and effort. Since S. oneidensis uses different extracellular electron transfer pathways depending on the electron acceptor, it is essential to use a reduction assay that mirrors the pathways employed in the microbial fuel system. For microbial fuel setups that use nanoparticles to stimulate electron transfer, reduction of graphene oxide provides a more accurate model than other commonly used assays as it is a bulk material that forms flocculates in solutions with a large ionic component. However, graphene oxide flocculates can interfere with traditional absorbance-based measurement techniques. This study introduces a novel image analysis method for quantifying graphene oxide reduction, showing improved performance and statistical accuracy over traditional methods. A comparative analysis shows that the image analysis method produces smaller errors between replicates and reveals more statistically significant differences between samples than traditional plate reader measurements under conditions causing graphene oxide flocculation. Image analysis can also detect reduction activity at earlier time points due to its use of larger solution volumes, enhancing color detection. These improvements in accuracy make image analysis a promising method for optimizing microbial fuel cells that use nanoparticles or bulk substrates. IMPORTANCE Shewanella oneidensis ( S. oneidensis ) is widely used in reduction processes such as microbial fuel generation due to its capacity to reduce electron acceptors. Often, these setups are labor-intensive to operate and require days to produce results, so use of a model assay would reduce the time and expenses needed for optimization. Our research developed a novel digital analysis method for analysis of graphene oxide flocculates that may be utilized as a model assay for reduction platforms featuring nanoparticles. Use of this model reduction assay will enable rapid optimization and drive improvements in the microbial fuel generation sector.
Bennett, Danielle T. (ORCID:0009000188748827)↗
Irradiation of MiniFuel Targets Bearing TRISO Fuel Compacts (Status Report)
Irradiation testing of MiniFuel compacts bearing tristructural isotropic (TRISO) fuel particles was performed at Oak Ridge National Laboratory (ORNL) to support the development of Kairos Power’s (KP’s) fluoride salt–cooled high-temperature reactor concept. The fuel compacts were fabricated with TRISO fuel particles of different types—including low-enriched uranium oxide, uranium carbide (LEUCO), natural uranium oxide, uranium carbide (NUCO), and low-enriched uranium dioxide (LEUO 2 )—and inserted into MiniFuel irradiation targets. Five targets were assembled and inserted in the High Flux Isotope Reactor (HFIR) for four cycles. The data collected post-irradiation will provide experimental input to validate TRISO fuel performance models for high particle power operations. This report summarizes the completion of the HFIR irradiation, the as-irradiated numerical analysis, and the post-irradiation work performed to date. This work was performed under the Nuclear Science User Facility program.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
ECAR-3873 Acceptance of Nonconforming Fuel for the AGR-5/6/7 Irradiation Experiments
Low-enriched fuel compacts fabricated for the AGR-5/6/7 irradiation experiments did not conform to all fuel specifications. Reasons why the fuel is acceptable for use in the AGR-5/6/7 irradiation are discussed. Revision 1 alters the characterization data for the dispersed uranium and exposed kernel defect fractions. Additional characterization data was obtained for fuel compacts analyzed by Oak Ridge National Laboratory. This enabled statistical outliers to be rejected and to confirm that results obtained from BWX Technologies Nuclear Operations Group were statistically independent of the Oak Ridge results and should be excluded from the analyses. Revision 2 alters the characterization data for the TRISO lot, J52R-16-98005, used in the fabrication of the AGR-5/6/7 fuel. Data from the BWXT J52R-16-98005 Certification package is used to calculate the reported values.
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Characterization Report for the MP-1 Experiment Fabrication Campaign
This report covers the results of MP 1 fuel characterization analysis to provide a detailed understanding of the as-fabricated fuel that would be irradiated in the MP 1 (Mini-Plate 1) experiment. The work was performed at Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and Los Alamos National Laboratory (LANL). Characterization of the as-fabricated fuel was performed in accordance with Characterization Plan for the Fabrication of U 10Mo for the MP 1 Experiment, hereinafter the MP 1 Characterization Plan (MAQP 2016). Under the MP 1 Characterization Plan, fuel foils at different points in the foil fabrication process (e.g., hot rolled annealed, cold rolled annealed) were studied to understand the effect of various thermomechanical processes on the fuel microstructure. Samples were sent to each of the three organizations so that testing and analysis could be performed independently using similar equipment and a standardized set of measurement and analysis procedures. The MP 1 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare independently obtained results, and achieve a three-way validation. PNNL characterized 28 uranium–10 wt% molybdenum (U 10Mo) samples. Six pieces/specimens from each sample/foil were sectioned, in accordance with the MP 1 Characterization Plan. Similarly, INL and LANL also examined 24 and 17 U 10Mo specimens, respectively. These total 69 samples consist of six types of MP 1 characterization foils fabricated by Babcock and Wilcox Technologies (BWXT) from ingots produced by the Y-12 National Security Complex: • 0.047 in.-thick hot-rolled and annealed samples with and without Zr layers • 0.02 in.-thick cold-rolled and annealed samples with and without Zr layers • 0.01 in.-thick cold-rolled and annealed samples with and without Zr layers. Microstructure, chemical composition, carbide morphology, U 10Mo foil thickness, Zr thickness, mechanical properties (microhardness), and density were evaluated in both longitudinal and transverse directions for foils of the three different thicknesses.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Characterization Report for the MP-1 Experiment Fabrication Campaign (Rev.1)
This report covers the results of MP 1 fuel characterization analysis to provide a detailed understanding of the as-fabricated fuel that would be irradiated in the MP 1 (Mini-Plate 1) experiment. The work was performed at Pacific Northwest National Laboratory (PNNL), Idaho National Laboratory (INL), and Los Alamos National Laboratory (LANL). Characterization of the as-fabricated fuel was performed in accordance with Characterization Plan for the Fabrication of U 10Mo for the MP 1 Experiment, hereinafter the MP 1 Characterization Plan (MAQP 2016). Under the MP 1 Characterization Plan, fuel foils at different points in the foil fabrication process (e.g., hot rolled annealed, cold rolled annealed) were studied to understand the effect of various thermomechanical processes on the fuel microstructure. Samples were sent to each of the three organizations so that testing and analysis could be performed independently using similar equipment and a standardized set of measurement and analysis procedures. The MP 1 experiment will provide an opportunity to understand the effects of processing conditions on the final fuel microstructure, to compare independently obtained results, and achieve a three-way validation. PNNL characterized 28 uranium–10 wt% molybdenum (U 10Mo) samples. Six pieces/specimens from each sample/foil were sectioned, in accordance with the MP 1 Characterization Plan. Similarly, INL and LANL also examined 24 and 17 U 10Mo specimens, respectively. These total 69 samples consist of six types of MP 1 characterization foils fabricated by Babcock and Wilcox Technologies (BWXT) from ingots produced by the Y-12 National Security Complex: (a.) 0.047 in.-thick hot-rolled and annealed samples with and without Zr layers; (b.) 0.02 in.-thick cold-rolled and annealed samples with and without Zr layers; and (c.) 0.01 in.-thick cold-rolled and annealed samples with and without Zr layers. Microstructure, chemical composition, carbide morphology, U 10Mo foil thickness, Zr thickness, mechanical properties (microhardness), and density were evaluated in both longitudinal and transverse directions for foils of the three different thicknesses.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A Robust Processing Approach for Producing Highly Loaded Dispersion Fuels
A robust fabrication method, resulting in higher yields, to produce highly loaded U3Si2-Al dispersion fuels for converting research reactors from a high enriched to a low enriched uranium fuel needs to be developed. To reliably produce a highly loaded dispersion fuel, process changes need to be implemented where the traditional approach has experienced challenges and poor yields. The present work describes the key changes needed. Parts of the work were done with uranium silicide and parts were completed with a representative silicide surrogate. The major deviations from traditional fabrication methods are associated with a refinement in particle size distribution, method for compacting to achieve complex shapes, and welding of the aluminum picture frame used to encapsulate the fuel compact. Methods for rolling and arc melting are also discussed. By using the methods described within, fabricating a highly loaded dispersion fuel that can meet stringent fuel homogeneity and geometry requirements at higher yields and lower costs may be possible.
U3Si2, HFIR, dispersion fuel, Uranium, Research Re↗
Research and Test Reactor Fuels
PRO-RR is the research reactor focused program element of the broader Proliferation Resistance Optimization program (PRO-X) under the National Nuclear Safety Administration (NNSA) in the U.S. Department of Energy (DOE). PRO-X provides a framework for integrating proliferation resistance in nuclear system designs to minimize weapons usable nuclear materials (WUNM) production and diversion pathways while optimizing systems performance for peaceful use missions. PRO-RR applies the PRO-X mission objectives to research reactor system design. This document serves as one of the foundational documents for the PRO-RR-Fuel System Design technical team by documenting current research reactor fuels usage. The PRO-RR-Fuel System Design technical team consists of subject matter experts from Argonne National Laboratory (Argonne) and Savannah River National Laboratory (SRNL). In order to determine the preferred fuel of use in upcoming research and test reactors to optimize proliferation resistance, performance, and safety, it is useful to assess the fuels that have been used in the past, or are currently in use. This report reviews the historical and current fuels used in research and test reactors to inform future fuel selection. Chapter 2 discusses the low-enriched uranium (LEU) fuels currently in use in terms of thermal power level and utilization of the reactor. Chapter 3 summarizes the fabrication processes for common fuel types. Chapter 4 discusses in detail the fuel types in use in research and test reactors. A review of the cladding types in use is presented in Chapter 5, and a historical review of research and test reactor fuel fabricators is presented in Chapter 6. The data collection strategy used the International Atomic Energy Agency (IAEA) research reactor database [1] as a starting point. Information on the fuel used was gathered on research reactors (other than critical assemblies) that were listed as operational, planned, or in temporary shutdown in the IAEA database. Data on the fuel type, geometry, enrichment, uranium loading, cladding type, and fabricator were collected for each of the reactors available in the public domain. Sources of data included conference papers, journal articles, and facility and fabricator websites. Data on research reactors operating on LEU fuels are presented in Appendix A, while Appendix B presents data collected on all reactors at the time of publication of this report. Appendix C presents data collected on reactors that were part of the M3 research and test reactor conversion program.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Thermal Model for Horizontal In-Water and In-Air Mini-Plate Experiments in the Advanced Test Reactor
Mini-Plate (MP) irradiation tests are a fueled experiment designed for irradiation in multiple test locations in the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). MP experiments are considered a non-instrumented drop-in test where small aluminum-clad fuel plate samples are cooled directly by the ATR Primary Coolant System (PCS) water. These fueled experiments contain aluminum-clad fuel mini plates consisting of monolithic U-10Mo. Previous thermal analyses have been documented concerning oxide growth[1] and thermal safety margin[2] for the MP-2 experiment. A thermal analysis has been performed on the MP-2 experiment using a new technique for calculating the necessary time after the experiment has been irradiated and reactor shutdown necessary to not exceed the 95% lower bound blister threshold temperature for the fuel in a horizontal in air configuration. A horizontal in water model has also been analyzed and will be discussed. After the experiment is irradiated in the ATR, it is necessary to calculate the fuel meat temperatures at 12 hours after shutdown. This is done to assure that at the 12 hour mark, if this experiment is removed from the reactor and accidentally dropped onto the working tray and is in the most severe geometrical configuration of horizontal (minimum heat transfer) that fission products will not be released. The blister threshold temperature is considered this safety limit. The horizontal in air analysis is also performed considering that the experiment is horizontal in a drained canal and filled with air and natural convection heat transfer with air on the outside. The initial mini-plate testing campaign known as MP-1 [3] has been irradiated and its purpose was to form the base fuel specimens fabricated by various processes so that a single process can be down-selected for qualification. MP-2 will then serve to populate a comprehensive fuel performance data set for the down-selected fabrication process. The MP-2 experiment will be irradiated in several locations of the ATR as shown in Figure 1. The fuel plate thickness and geometry is identical to the MP-1 experiment [3], with the only variable being the fuel meat thickness. This model of the B-12 position is treated as encompassing the I-21, I-24, and B-9 positions since it has more fuel and higher total heat than the other positions. Since the blister threshold temperature is a function of fission density, a series of fission densities is calculated for the B-12 position to cover the other three positions. The B 12 position specifically uses fuel meat that is 0.016 in thick by 3.25 in long and 0.75 in tall. The total plate thickness is 0.050 in. Aluminum cladding forms the barrier between the PCS and the fuel meat. The purpose of this paper is to demonstrate this new technique developed to calculate the necessary decay time after shutdown for a series of fission densities using the ABAQUS [4] finite element and heat transfer code.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
University of Missouri Research Reactor LEU Fuel Element Flow Test Conceptual Design—Hydraulic Reactor Design Parameters
The University of Missouri-Columbia Research Reactor (MURR®) is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that actively collaborates with the National Nuclear Security Administration (NNSA) Material Management and Minimization(M 3 ) Reactor Conversion Program to convert to the use of low-enriched uranium (LEU, < 20 wt.% U-235) fuel. A new type of LEU fuel with very high density, based on an alloy of uranium and 10 weight percent molybdenum (U-10Mo), is expected to allow the conversion to LEU of USHPRR that have been found unable to be converted with previously qualified uranium silicide-aluminum (U 3 Si 2 -Al) dispersion fuel. MURR has been working with the USHPRR Reactor Conversion (RC) Pillar at Argonne National Laboratory to perform fuel element design and fuel cycle performance analyses, steady-state thermal hydraulics safety analyses, and accident safety analyses in preparation for the conversion of MURR and to support a preliminary Safety Analysis Report (SAR) for conversion to LEU fuel. This work is performed in preparation for the flow test campaign that will be conducted by the USHPRR RC Pillar. The purpose of the hydraulic performance evaluation of the MURR LEU fuel element designed by the RC Pillar is to test a prototypic commercially fabricated LEU fuel element to determine whether any failure modes are observed or predicted in the fuel element, including significant deformations such as plate bending, twisting, or plate detachment from the side plate under selected safety-basis limits for reactor hydraulic conditions. To support the design of the flow test for MURR LEU fuel element hydraulic performance evaluation, design parameters for hydraulic testing of the LEU fuel element are laid out in this report.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Summary of Leach-Burn-Leach Round-Robin Test Results
Tristructural isotropic (TRISO) coated particle fuel is of interest to multiple member states associated with the Generation IV International Forum (GIF). Qualification and acceptance testing of TRISO fuel includes the determination of particle defect fractions and impurities in the as-fabricated TRISO fuel particle and/or composited fuel form, which are typically determined using burn-leach (BL) or leach-burn-leach (LBL) analysis methods. A round-robin test focusing on LBL analysis was undertaken by the GIF Very High Temperature Reactor System Fuel and Fuel Cycle Project Management Board. Participating in this LBL round-robin test were research staff from the Institute of Nuclear and New Energy Technology (INET) in the People’s Republic of China, the Korean Atomic Energy Research Institute (KAERI) in the Republic of Korea, and Oak Ridge National Laboratory (ORNL) in the United States of America. The motivation of this LBL benchmarking exercise was to explore consistency in results generated by LBL and to elucidate the potential influence of variations in the procedures used by the participating members on the accurate determination of the measured fuel properties.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Nuclear fuel irradiation testbed for nuclear security applications
The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).
98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates
Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗