Reactor physics depletion benchmarks for high-assay low-enriched uranium (HALEU) fuel are limited in number. In particular, there is limited data for HALEU benchmarks for U-10Mo (uranium-10% molybdenum) plate fuel that is being developed for use in the United States’ high performance research reactors including the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), Massachusetts Institute of Technology Reactor (MITR), University of Missouri Research Reactor (MURR), National Bureau of Standards Reactor (NBSR). These six reactors currently operate with highly enriched uranium dispersed fuel in an aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiating fuel specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This paper documents the development of a benchmark model of the irradiation of the ATR Full-size plate In center flux trap Position 7 (AFIP-7) experiment using the depletion codes MC21 and Advanced Dimensional Depletion for Engineering of Reactors (ADDER).

This study presents the first measurements of the individual thermal conductivities of U-7 wt.%Mo fuel particles and U-Mo/Al interaction layers (ILs) from in-pile-irradiated dispersion fuel plates using the suspended-bridge method. Nanorods of U-7 wt.%Mo fuel and U-Mo/Al ILs were extracted by focused ion beam (FIB), and their microstructures were characterized with transmission electron microscopy (TEM). TEM revealed finely distributed nanobubbles in the U-7 wt.%Mo matrix, along with an amorphous structure in the ILs. The thermal conductivity of in-pile-irradiated U-7 wt.%Mo was approximately 30 % lower than that of the unirradiated material, ranging from 6.7 W/m center dot K at 300 K to 8.5 W/m center dot K at 380 K. The ILs exhibited even lower thermal conductivity, from 2.1 W/m center dot K at 300 K to 2.7 W/m center dot K at 380 K. These reductions, attributed to nanobubbles, fission products, and irradiation-induced point defects, were analyzed through a combination of microstructural characterization and literature-based transport models, which successfully reproduced the observed degradation trends.

Post-irradiation microstructure characterization plays an important role in qualifying the low-enriched uranium (LEU) monolithic U-10 wt%Mo plate-type fuel for United States high-performance research reactors (USHPRRs) program. Inhomogeneous features resulting from manufacturing and irradiation processes, including carbides, second phase stringers, and extensive void spaces caused by the combining of small porosities, may increase the risk of heat concentration in local regions of the fuel plate over the operating conditions. In this study, characteristics of carbides, stringers, and porosity after multiple levels of irradiation at varying fission densities were studied by electron microscopes to decipher the morphology of pores and the porosity evolution in U-10 wt%Mo. For carbides, the result shows that porosities start forming on UMo grain boundaries, then on UMo/carbides interfaces as the burn-up going higher. However, the porosities surrounding carbides grow larger than the ones on UMo grain boundaries. The porosities around the uranium carbides could interconnect to form larger void space. The study revealed that the void spaces larger than 5 & micro;m were found around uranium carbides after high burnup, while no evidence was observed to support the similar voids formed near second phase stringers even though the size of the stringers (> 50 & micro;m) was much larger than uranium carbides (< 20 & micro;m). The evolution of porosities suggests that the formation of second phase stringers may not create more significant porosities compared to regular uranium carbides regions during fuel operating conditions.

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. 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.

Advanced microstructural characterization techniques, such as scanning electron microscopy (SEM) and scanning transmission electron microscopy -energy dispersive x-ray spectroscopy (STEM-EDS), were used to interpret the fuel microstructure evolution and fission products behavior in U-Mo dispersion fuel irradiated in the Advanced Test Reactor (ATR) as part of the European Mini-Plate Irradiation Experiment (EMPIrE) test. The larger as-fabricated fuel grain size achieved by heat-treating the U-Mo powder resulted in slower high burnup structure (HBS) development and reduced fission gas porosity. Slower HBS kinetics was observed at the fuel kernels' periphery, which contained smaller and less fission gas bubbles at all fission densities (FDs) investigated and was attributed to a locally reduced damage density and fission products concentration, as corroborated with Monte Carlo simulations. The non-refined grains at the fuel kernel periphery hosted a perfectly ordered fission Gas Bubble Superlattice (GBS) up to 6.3 x 1021 fissions/cm3. Nano-scale STEM-EDS analysis presented in this study provided useful information on the GBS characteristic morphology and evolution in U-Mo fuel. The concentration of fission gas in the GBS progressively increased with FD, pointing to an evolution of the nanobubble pressure status with irradiation. A possible connection between the GBS collapse and HBS onset is proposed for which there exists a threshold in the misorientation of the refined sub-grains above which the GBS stability during irradiation is no longer preserved, resulting in the GBS collapse.

The Korea Atomic Energy Research Institute (KAERI) is leading the Ki-Jang Research Reactor (KJRR) project with the intent to develop a new reactor for medical isotope production and other nuclear research purposes. The KJRR core is designed to use high density fuel system where uranium alloyed with 7 wt% molybdenum (U-7Mo) particles are dispersed in a matrix of aluminum alloyed with 5 wt% silicon (Al-5Si) and clad in aluminum alloy 6061 (Al-6061) to form fuel plates. KAERI developed a fabrication facility to construct KJRR fuel assemblies and partnered with the Idaho National Laboratory (INL) to irradiate a full-size fuel assembly, with 21 total fuel plates, in the Advanced Test Reactor (ATR). Irradiation testing and subsequent Post Irradiation Exam (PIE) campaigns were performed successfully over a multi-year project. Monte Carlo neutronic calculations, coupling with a depletion code, were performed based on ATR's as-run power history which showed that the highest power plate (plate 20) reached 83.1 % end-of-life (EOL) local burnup based on initial 235U content. Finite element thermal modeling was performed based as-run power history which showed a beginning-of-life (BOL) peak local heat flux of 184 W/cm2. No anomalous fuel performance was observed during the irradiation and target test conditions were achieved. PIE showed favorable performance of the fuel assembly regarding all important phenomena. This paper describes the KJRR fuel assembly irradiation conditions and PIE data to support the conclusion that it performed well, without evidence of unexpected or problematic fuel performance, within an irradiation test designed to bound the KJRR design environment.

In this study, the effect of various scanning electron microscope (SEM) imaging parameters, including voltage, beam current, and magnification, on the porosity analysis of U-Mo fuel was investigated by varying one parameter at a time during back-scattered electron (BSE) micrograph acquisition. Among the parameters assessed, porosity analysis is most sensitive to the SEM voltage. In the range from 5 kV to 30 kV, we reported relative differences as high as 30%, 10%, and 20% in the total porosity, average diameter, and pore density, respectively. Monte Carlo simulations were also performed to determine the influence of SEM voltage on the probing depths of back-scattered electrons. Increasing the voltage from 5 kV to 30 kV resulted in a change in probing depth from few tens of nanometers to several hundreds of nanometers. Portions of pores or entire pores residing below the polished plane that could not be seen at low voltages, and should not be counted for in the estimation of the open porosity, became visible at high voltages. Interestingly, in addition to allowing for a better estimation of the open porosity, the higher sensitivity to surface features at low voltages seems to reduce the number of pixels with intermediate gray levels in the final BSE image. This, in turn, could limit the ground for different interpretation by different analysts, and, as such, has the potential to facilitate a more consistent and uniform porosity analysis across different research laboratories.

This study investigated possible failure mode sequences in U-10Mo monolithic fuel irradiated to very-high-burnup. The U-Mo fuel plate used in this study was characterized using scanning electron microscopy (SEM), wavelength dispersive spectroscopy (WDS), and image analysis techniques to investigate how microstructural features may initiate crack formation and propagation through the microstructure, potentially leading to blistering. Distinctly large fission gas pores (FGPs) were observed to preferentially align 5-15µm away from the U-10Mo/Zr interface, parallel to the Zr diffusion barrier. Preferential growth, alignment, and interconnection of FGP could initiate blistering in the fuel plate by creating a fission gas channel, parallel to the U-Mo/Zr interface. FGP alignment in the fuel phase near the U-Mo/Zr interface is believed to be one of the precursors to crack and Type 2 blister formation in monolithic U-Mo fuels. The chemical maps revealed differences in fission product behavior near the U-Mo/Zr interface such that Nd remains in the fuel matrix at the FGP sites. On the other hand, other fission products like Xe and Cs can get trapped in FGPs by Nd in the fuel phase and/or diffuse toward the Zr diffusion barrier.

This study investigated the microstructural behavior of both full-size and mini-size monolithic U-10Mo fuel plates irradiated to high burnup with a focus on the evolution of the second phase impurities in monolithic U-Mo using scanning electron microscopy (SEM), energy dispersive spectroscopy (EDS), and wavelength dispersive spectroscopy (WDS). Key indicators of possible mechanical and thermal compro-mise include cracks, large fission gas porosity, and interconnection of fission gas pores. For the fission densities evaluated in this work (3.5 x 10 21 fissions/cm3-5.1 x 10 21 fissions/cm3), fine porosity can de-velop along the UC phase boundary; however, the size of the fission gas pores is no more than those observed in the U-Mo fuel phase. Other inclusions such as Si-rich second-phase impurities found in the as-fabricated microstructure were difficult to resolve post-irradiation because they can become overshad-owed by porosity development in the fuel phase. Additionally, the presence of a Fe-rich sublayer formed in the Zr diffusion layer during fabrication remained enriched in the Zr layer in the irradiated U-10Mo microstructures near the U-Mo/Zr interface; however, based on the burnup assessed in this study the identified impurities did not appear to contribute to notable microstructural degradation.(c) 2023 Published by Elsevier B.V.

For years, researchers have reported competing ideas on the formation of fission gas bubble superlattices in metallic fuels, which was postulated to be comprised of elemental xenon (Xe) atoms. However, the chemical and physical arrangement of these elements within this gas bubble superlattice had not been verified. In this contribution, Xe was chemically profiled using atomic resolution scanning transmission electron microscopy (STEM) and atom probe tomography (APT) techniques in irradiated uranium-molybdenum (U-Mo) fuels. These complementary techniques provide conclusive evidence of Xe presence in the bubble superlattice up to fission densities of 4.5 x 1021 fissions/cm3 and give a quantitative assessment of Xe content within the bubble super lattice. Based on the results of simultaneous imaging and spectroscopy using STEM, the chemical composition of the Xe bubble superlattice was measured and found to be up to 8 +/- 1.3 atomic %. APT data complemented STEM findings on Xe distribution in irradiated U-Mo fuel sample. This study provides conclusive evidence that Xe is not only present within an atomically arranged bubble superlattice but provides fundamental insight into the physical state of Xe in low enriched U-Mo monolithic fuel.

The use of burnable absorbers can be beneficial for nuclear reactors by extending the fuel's operational cycle, providing additional criticality control, and flattening the power profile. In this work, three burnable absorber materials (boron carbide, aluminum boride, and zirconium boride) embedded in aluminum have been fabricated into foils and clad in AA-6061 for potential use in high performance research reactors. The as-fabricated boron-containing phases were determined using transmission electron microscopy to be AlB2, B4C, and ZrB2. TEM also revealed incomplete bonding at the B4C-matrix interface. SEM showed a relatively uniform spatial distribution of boron-containing phases for all the candidate materials. Higher porosity was observed in the foil containing ZrB2 in its as-rolled condition. The porosity in the ZrB2 foil was reduced by hot isostatic pressing. The size and shape distributions of the boron-containing phases were analyzed on the criteria of cross-sectional area, perimeter, roundness, circularity, and aspect ratio. A method of converting the 2D burnable absorber dispersoids seen in cross-sectional microscopy images into 3D volumes was derived using both spherical and ellipsoidal geometry models. The difference in calculated burnable absorber dispersoid average volume between the two models ranges from 20% to 100%, which could impact burnable absorber burnout rates due to differences in neutron self-shielding.

The European Mini-Plate Irradiation Experiment (EMPIrE) was designed to support the development and testing of a coated uranium-molybdenum (U-Mo) dispersion fuel for the conversion of select high-performance research reactors (HPRRs) to utilize low-enriched uranium (LEU). To aid in the development of the coated fuel form, the EMPIrE test included several plate designs and irradiated them in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) at a high meat power density (similar to 21 kW/cm(3)) and to high fuel particle fission densities (similar to 6.4 x 10(21) fissions/cm(3)). These conditions mimic the bounding conditions of the BR-2 reactor in Belgium, where a concurrent irradiation experiment was performed, and exceed those previously explored in dispersion U-Mo fuel plates. A local fuel swelling analysis, as determined through high-fidelity, post-irradiation mini-plate profilometry, was used along with statistical methods to non-destructively evaluate the overall performance and separate the effects of convoluted fabrication variables. While some effects observed with this non-destructive analysis were subtle, others had more significant, and possibly competing, effects on the fuel swelling behavior. These observations will be examined further with destructive examinations to more fully assess them as the fuel design is developed and qualified. (C) 2022 Elsevier B.V. All rights reserved.

Understanding swelling behavior in monolithic, uranium-molybdenum, plate-type fuel is necessary to qualify the fuel for reactor use and for the conversion of high performance research reactors from highly enriched to low-enriched uranium. Multiple mechanisms influence plate dimensional stability, including solid and gaseous fission-product induced swelling, irradiation-assisted creep, fuel-phase transformation, and interaction-layer formation. Separating these phenomena remains a challenge, and current models do not appear to adequately predict experimental results at higher fission densities where it is most critical. To mitigate the mechanistic uncertainty, post-irradiation profilometry is used to increase the number of data points available over a range of in-reactor irradiation experiments conducted at the Advanced Test Reactor, and to provide a statistical precedent for a swelling-behavior model. This work establishes a predictive swelling correlation as a function of fission density, with associated confidence and prediction bounds, by analyzing more than 18,000 thickness data points collected on 74 irradiated U-10Mo monolithic fuel test plates over a range of irradiation conditions.

This work describes the microstructural evolution of prototypical monolithic U-Mo fuel plates analyzed via scanning electron microscopy and electron backscattered diffraction (EBSD). The understanding of the microstructural and textural evolution of nuclear fuel from as-fabricated to post-irradiation is important in assessing changes in material properties during irradiation. In our work it was observed that the typical fabrication techniques applied in U-10Mo monolithic fuel plates develop features associated with a cold-rolled body-centered cubic (bcc) texture and development of alpha and gamma fiber (parallel to the rolling and normal direction). After irradiation, a loss of the fabrication-induced preferred orientation was observed with an increased spread of grain-boundary misorientation as burnup increases. Grain subdivision was observed in the irradiated samples with the formation of submicron grains (200-500 nm). Evidence for polygonization as the mechanism leading to grain subdivision was detected. This has been observed for the first time for U-Mo monolithic fuel via EBSD and has been associated to formation of low-angle grain boundaries (<15 degrees) at the site of the submicron grains. Such analyses of the microstructural and textural evolution of fuel (from fabrication to after irradiation) have the potential to help develop and validate microstructural physics-based models and provide a key feedback loop to further understand the interplay between fabrication processes and fuel performance. (C) 2020 Elsevier B.V. All rights reserved.

Irradiation conditions such as fission power, fission rates, and temperature are important parameters for nuclear fuels because they influence the microstructural behavior and ultimate irradiation performance. Two U-7Mo/Al-3.5 wt.% Si dispersion fuel plates were irradiated in the Advanced Test Reactor edge-on to the core in the RERTR-9B experiment at different powers to study the effects of power on fuel phase swelling behavior. The results of non-destructive measurements have been reported in a previous paper [1], and the destructive examination results are being reported here. In particular, the microstructural evolution of the U-Mo fuel phase in irradiated fuel plates has been investigated by cutting transverse cross-sections, using generated scanning electron microscopy micrographs, and image processing methods to assess the changes in microstructural features, such as fission gas pore growth and non-recrystallized grain fraction. In this study, the focus was to examine the evolution of fission gas pores from both the constrained and unconstrained regions of U-Mo fuels irradiated at different powers. Based on an estimated linear gradient, the rate of change of fission gas pore size with locally calculated fission densities is more prominent in the higher power specimen. As such, the average porosity and pore area is approximately 11% and 2 xlarger in the fuel irradiated at higher power than that of the lower power specimen. The growth of fission gas pores can accelerate the pore interconnectivity and thus fission gas outside the fuel kernel (FGO), which are two precursors for swelling in U-Mo fuels. The influence of increased power, internal plate stresses, and fission-induced creep on fuel kernel deformation and lateral mass transfer is also highlighted. (C) 2020 Elsevier B.V. All rights reserved.

In order to investigate the irradiation performance of U3Si2/Al dispersion fuel at relatively aggressive reactor conditions, compared to how this fuel is typically used in a research or test reactor, post irradiation examination was performed on two fuel plates that were irradiated at similar to 270 W/cm(2) average surface heat flux and to a maximum local burnup greater than 8 x 10(21) fissions/cm(3) in the Advanced Test Reactor as part of the RERTR-8 experiment. As part of the non-destructive examinations that were performed on both fuel plates, detailed thickness measurements using a high-fidelity measurement bench were performed on one fuel plate and a segment of the second fuel plate. This paper provides the first reported results of such high-resolution thickness measurements for U3Si2/Al dispersion fuel plates, containing primarily the U3Si2 fuel phase, irradiated at such power levels to high burnup. For destructive examination using optical metallography, samples were generated from both irradiated plates. Localized regions of high swelling (characteristic of internal blistering) were observed on the surface of both fuel plates, suggesting that the performance limits had been reached for each plate. Destructive examinations showed that the fuel particles that are primarily U3Si2, which become amorphous during irradiation, in both plates contained relatively large fission gas bubbles with evidence of fission gas bubble interconnection. (C) 2019 Published by Elsevier B.V.

Post-irradiation examination of RERTR-12 miniplates showed that in-reactor pillowing occurred in at least 4 plates, rendering performance of these plates unacceptable. To address in-reactor failures, efforts are underway to define the mechanisms responsible for in-reactor pillowing, and to suggest improvements to the fuel plate design and operational conditions. To achieve these objectives, the mechanical response of monolithic fuel to fission and thermally induced stresses was modeled using a commercial finite element analysis code. Calculations of stresses and deformations in monolithic miniplates during irradiation and after the shutdown revealed that the tensile stress generated in the fuel increased from 2 MPa to 100 MPa at shutdown. The increase in tensile stress at shutdown possibly explains in-reactor pillowing of several RERTR-12 miniplates irradiated to the peak local burnup of up to 1.11x10(22) fissions/cm(3). This paper presents the modeling approach and calculation results, and compares results with post-irradiation examinations and mechanical testing of irradiated fuel. The implications for the safe use of the monolithic fuel in research reactors are discussed, including the influence of fuel burnup and power on the magnitude of the shutdown-induced tensile stress.

Electron Backscattered Diffraction (EBSD) is an effective technique for revealing many details about the microstructure of materials, e.g. crystallographic orientation, grain size, grain boundary properties, texture, intragranular misorientation, and subgrain formation. Since these details are of interest for improving the irradiation performance understanding of any irradiated nuclear fuel (e.g. swelling behavior), this technique has been applied successfully on U-7 wt% Mo before and after irradiation. This fuel is a high-density, low-enriched uranium fuel currently being developed for application in research and test reactors. Based on the results of this characterization, it was found that when as-fabricated U-7 wt% Mo is irradiated to around 5.3 x 10(21) fissions/cm(3) the original large grains (diameter similar to 4 mu m) are subdivided into much smaller grains (diameter similar to 0.3 mu m) and most of these subdivided grain boundaries are low angle boundaries. The EBSD analysis suggests that the grain subdivision in the irradiated U-7 wt% Mo was driven by polygonization, not recrystallization as defined by classic metallurgy as a result of heavy cold work followed by heat treatment. (C) 2018 Elsevier B.V. All rights reserved.

The Materials Management and Minimization program is developing fuel designs to replace highly enriched fuel with fuels of low enrichment. Swelling is an important irradiation behavior that needs to be well understood. Data from high resolution thickness measurements performed on U-7Mo dispersion fuel plates with Al-Si alloy matrices that were irradiated at high power is sparse. This paper reports the results of detailed thickness measurements performed on two dispersion fuel plates that were irradiated at relatively high power to high fission densities in the Advanced Test Reactor in the same RERTR-9B experiment. Both plates were irradiated to similar fission densities, but one was irradiated at a higher power than the other. The goal of this work is to identify any differences in the swelling behavior when fuel plates are irradiated at different powers to the same fission densities. Based on the results of detailed thickness measurments, more swelling occurs when a U-7Mo dispersion fuel with Al-3.5Si matrix is irradiated to a high fission density at high power compared to one irradiated at a lower power to high fission density.