The mechanical degradation of the Al matrix in U-7Mo dispersion fuels was detected after irradiation in previous studies, but more details of the degradation mechanism are not clear yet to clarify the root cause. In this study, microhardness mapping in fresh and irradiated U-7Mo dispersion fuels coated with ZrN, with burn-up of 3.35 & times; 10(21) and 6.28 & times; 10(21) f/cm(3), are measured to reveal the hardening of the Al matrix as a function of distance to fuel particles. Micro cubes were pulled by plasma focused ion beam scanning electron microscopy and tested by in situ nanoindenter with Berkovitch tip. Size effects are measured in medium and high burn-up samples. The higher burn-up specimen exhibits a more pronounced size effect when indent load is lower than 20 mN. Size effect correction models are also calculated by plots fitting, which predicts the true microhardness without size effect. Results shows that the hardness of the Al matrix near the ZrN coating has the highest value, then decreases moving away from the coating and become stable when the distance reaches similar to 10 & micro;m. The hardness starts increasing again getting closer to the next fuel particle, which makes the hardness distribution between two fuel particles to be "U" shape.

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.

FeCrAl alloys are considered as a candidate material for fuel cladding in commercial light water reactors (LWRs) replacing Zircaloy owing to their high temperature strength and corrosion resistance in steam environments (around 1000oC) [1]. However, FeCrAl alloys also suffer from embrittlement after aging at 475 oC and lower (the operating temperature of LWRs falls into this range), due to α’ Cr precipitation that results in an increase in the DBTT [2, 3]. This phenomenon is detrimental to the structural performance and also to the corrosion resistance of KD. Therefore, it is important to deeply study and understand α’ Cr precipitation in high Cr ferritic alloys. There exists no systematic study that looks into the influence of grain size on α’ Cr precipitation during isothermal aging. This study will be the first to systematically study the precipitation of nanoscale α’ Cr (5-10 nm) as a function of grain size [coarse-grained (CG, grain diameter >1 μm; ultrafine grained (UFG, 100 nm < grain diameter < 1 μm); nanocrystalline (NC, grain diameter < 100 nm)] during isothermal aging at or below 475 oC. Kanthal-D was manufactured and samples using two forms of severe plastic deformation (SPD), equal channel angular pressing (ECAP) and high-pressure torsion (HPT), respectively. had a non-homogenous microstructure with multi-modal grain size distribution revealed using electron-back scatter diffraction. Average

The United States High Performance Research Reactor project is tasked with fuel development and qualification leading to conversion of higher power research and test reactors in the US from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuels. This manuscript identifies the functional and operational design requirements of the first miniature test plate (mini-plate [MP]) irradiation campaign (MP-1) of commercially fabricated LEU U-10Mo monolithic plate-type fuel and is the precursor to a large parametric mini-plate test (MP-2) aimed at producing the data to support regulatory qualification of the LEU U-10Mo monolithic fuel. The manuscript (1) provides a general description of the selected U-10Mo LEU fuel and (2) defines the overall experiment design and functional requirements to accomplish the specific test objective of MP-1, which is to confirm that the commercially manufactured LEU U-10Mo monolithic fuel meets the established requirements of geometric stability, mechanical integrity and stable and predictable behavior. The fuel testing parameters are established by the need to bound performance behavior within the operational envelope of the reactors being converted.

Using scanning electron microscopy (SEM), this study evaluates the microstructure evolution of uraniummolybdenum (U-Mo) fuel foils made with and without heat treatment at medium burn-up (of approximately 5 x 1021 f/cm3). The impact of annealing treatments on critical microstructural properties of the U-Mo fuel foils, including porosity, grain structure, Mo homogeneity, and fuel interaction with the Zr interlayer, was examined using large area lift outs (LALOs). The heat-treated specimens presented less grain refining at these burnups when compared to the non-heat-treated specimens. Grain refinement was associated with porosities and fission products precipitation. These observations may indicate that heat treatment can influence fuel swelling during irradiation. Chemical inhomogeneity (Mo banding) was found to persist in the non-heat-treated samples but was not present in the heat-treated samples. Thus, heat-treated U-Mo foils allow for more predictable fuel behavior under irradiation with respect to non-heat-treated foils. The U-Mo and Zr interaction layer appears to be thicker and more continuous in the heat-treated sample which has been associated with stronger interface integrity during irradiation, as observed in previous studies. These observations may indicate an overall improved performance of heat-treated fuel foil in a reactor, which needs to be confirmed with higher burnup samples. The effect of local burn-up on grain size/refinement and porosities in each LALO specimen, sampled from different positions in the fuel foil, was difficult to analyze due to the large standard deviation of these parameters. Finally, evidence of grain refinement by polygonization may be present in these specimens.

Performance of two distinct fuel systems, U-7Mo fuel in Zircaloy (Zry-4) cladding and U-10Mo fuel in aluminum (Al6061-O) cladding was studied. First, a mini plate with Zry-4 cladding from a previous irradiation experiment was evaluated via finite element analysis (FEA). By using the same plate geometry and irradiation conditions, another plate consisting of U-10Mo fuel and Al6061-O cladding was simulated. The results were then comparatively evaluated to explore the feasibility of employing Zry-4 as an alternative cladding. Simulations indicated the Zircaloy cladding plate would operate roughly 50 degrees C hotter as compared with the Al alloy cladding plate. Larger deformations in the thickness direction for the plate with Zry cladding were noted. It was observed that the postfabrication stresses in the fuel would be relieved quickly in the reactor, regardless of cladding type. Although the fuel stresses would still develop at reactor shutdown, the fuel would be stress-free during the entire irradiation period for both cladding types. At shutdown, the plate with Zry cladding would have higher stresses due to higher operating temperatures. Similarly, the stresses after shutdown are higher in the foil core for the plates with Zry cladding. The Al cladding plate would have higher plastic strains as compared with the Zry cladding plate. The Zry cladding plate is significantly stiffer, causing higher stresses in the fuel zone and at the interface. Overall, employing Zry as an alternate cladding is not expected to produce a more favorable thermomechanical performance as compared to the performance of an Al alloy cladding plate.

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.

In this study we compared the microstructure evolution of U-Mo fuel foils produced with and without heat treatment at low burn-up via advanced post-irradiation examination. The aim of this study is to observe after irradiation the ways in which the fabrication processes have influenced fuel behavior at early-stage irradiation, as for very low burn up microstructural studies are lacking. In this work it was observed that the larger grain size detected in the heat-treated samples before irradiation led to decreased grain refinement after irradiation. Grain refinement was associated with the presence of small nano-size bubbles and precipitates. This phenomenon is hypothesized to influence early fuel swelling during reactor irradiation. Grain refinement was also observed to increase in regions where & gamma;-U decomposition was present. Thus, it was enhanced in the samples fabricated without heat treatment. The heat treatment also increased the thickness of the U-Mo/Zr interface, namely of the UZr2 layer. However, the influence of this layer on fuel performance needs further investigation. On one side, it may contribute to better mechanical bonding; on the other, it may influence swelling and blistering in the interaction layer as porosity increases when this layer is increased. This was observed especially in the presence of increased area containing low Mo concentration, and thus containing a higher fraction of the & alpha;-U phase, which is highly susceptible to irradiation induced swelling. Strong evidence of reverse transformation under irradiation (& alpha;-U + & gamma;& PRIME;-U2Mo -+ bcc & gamma;-U) was observed in these samples. While the precipitates (carbides and oxide) seem to be unaffected by the irradiation at these low burnups. However, further analyses are necessary at higher burn-up to assess the exact impact different heat treatments have on fuel performance.

In order to minimize proliferation risks and improve security of nuclear material, the United States high performance research and test reactors (USHPRR) program is tasked with converting nuclear reactors that are fueled with highly enriched uranium (HEU) fuels to operate with low-enriched uranium (LEU) nuclear fuels. One favorable LEU fuel configuration is plate fuel with a metallic uranium-molybdenum foil clad within an aluminum alloy (AA 6061). In this fuel, the aluminum cladding is bonded with a hot isostatic pressing (HIP) method to seal the cladding around the fuel meat. However, the HIP process parameters influence the cladding performance, as a defective or incomplete bond can cause a pathway for corrosion. Micro X-ray computed tomography (XCT), a nondestructive technique that provides volumetric imaging, can be applied to inspect fuel plate cladding at the engineering scale. In this work, XCT methodology was developed and successfully utilized to not only observe the bond line of unirradiated mini fuel plates, but to also identify subsurface abnormalities in the plates’ cladding. In future work, this technique could be applied to fuel plates, pre- and post-irradiation, to quantify irradiation effects on cladding defects and bond line integrity.

Hot isostatically pressed AA6061 cladding is an important structural component of the high performance, Zrlaminated U-10Mo monolithic fuel system for the application in research and test reactors. In this study, the mechanical behavior of two diffusion bonded aluminum alloy, AA6061, was examined using tensile testing. Solid-to-solid diffusion bonding between two pieces of AA6061 was performed by isothermal annealing at 560 degrees C for 1.5 h, and diffusion couples were subsequently cooled via three different cooling methods: furnace cooling, air cooling, and water quenching. Dog-bone shaped tensile specimens, with 10 mm in gauge length (with diffusion bonded interface in the middle), and 1.5 x 1.5 mm2 gauge cross-sections, were fabricated from the diffusion bonded AA6061 by electro-discharge machining. Yield strength (% EL at failure) of furnace cooled, air cooled and water quenched tensile specimens determined was 82-89 MPa (10-30%), 112-116 MPa (10-14%), and 149-164 MPa (10-17%), respectively. This variation in mechanical behavior was examined with coolingrate dependent, concentrated precipitation of Mg2Si at the diffusion bonded interface, with due respect for mechanical properties of the AA6061 alloy that inherently vary as a function of cooling rate from 560 degrees C. Finite element analysis using ABAQUS was employed to augment experimental findings with the appropriate microstructural constituents and alloy properties. Results suggest that the strength is dominated by matrix/bulk properties of AA6061, while ductility is strongly influenced by the cooling method dependent presence of Mg2Si precipitates at the interface.

Considerable effort is being devoted to development and regulatory qualification of low enriched fuels for research and test reactors by many agencies worldwide. One promising fuel configuration being examined for United States higher power research and test reactors (USHPRRs) are plate-type fuels composed of a metallic uranium-molybdenum foil clad in an aluminum alloy. The two pieces of aluminum alloy cladding are bonded using a hot isostatic pressing method. The mechanical properties of the resulting bond line in the aluminum alloy cladding will vary by the HIP'ing parameters, requiring a need to characterize the bond line. Small scale mechanical testing can provide a path for evaluating the mechanical properties and deformation behavior of the bond line both prior to and following irradiation. In this research, room temperature micro-tensile specimens of non-irradiated and irradiated samples containing an Al alloy (AA 6061) bond line were tested to evaluate its strength and deformation behavior. Observations indicated that the strain rate did not affect the deformation behavior or strength and most of the micro-tensile specimens failed in a ductile mode in grains around the bond line. There was no indication that the microstructural features from the bond line affected the mechanical properties of the micro-tensile specimens. An initial examination was performed on irradiated material but further systematic studies of the effects of irradiation can be performed in the future.

The objective of this paper is to analyze the effect of different heat treatments on the fuel microstructure and texture in laboratory scale monolithic U-Mo fuel plates. Such analyses are relevant as they could possibly inform future fabrication methods for the optimization of microstructure and of material properties to ultimately improve fuel performance during irradiation. For this reason, detail characterization including Energy x-ray dispersive spectroscopy (EDS) and electron backscattered diffraction (EBSD) techniques were applied on U-Mo plates fabricated with different heat treatments to understand the influence of fabrication on fuel microstructure. U-Mo fresh fuel specimens with or without annealing treatments (homogenization and stress release annealing) after rolling were analyzed in this work. In this study for the first time EBSD was systematically used to analyze the heat treatments influence on texture present in fresh fuel monolithic U-Mo fuel plates. Such changes were connected to other known microstructural and chemical changes. In this work minimization of molybdenum concentration variation and gamma phase decomposition of the U-Mo fuel core after homogenization was observed. Also, an increase in the interaction layer of the U-Mo fuel core with the Zr interlayer diffusion barrier was observed with the UZr 2 growing up to 1 mu m in thickness. Such microstructural changes were aligned to the changes in texture and grain structure. Indeed, common texture observed in body-centered cubic (BCC) metal after cold rolling (e.g., alpha, xi, gamma fibers) were minimized by the annealing/homogenization process. Moreover, grain structure was influenced by the fabrication route with elongated grains formed in the rolled sample and equiaxed grains found after annealing. Such features are relevant since the formation of microstructural features formed during fuel fabrication can impact fuel performance in reactor.

Background In nuclear fuel plates of low-enriched U-10Mo (LEU) clad with aluminum by hot isostatic pressing (HIP), post-irradiation stresses arising during reactor shutdown are a major concern for safe reactor operations. Measurement of those residual stresses has not previously been possible because the high radioactivity of the plates requires handling only by remote manipulation in a hot cell. Objective The incremental slitting method for measuring through-thickness stress profiles was modified, and a system for automated, remote operation was built and tested. Methods Experimental modifications consisted of replacing electric-discharge machining (EDM) with a small end mill and strain-gauge measurements with cantilever displacement measurements. The inverse method used to calculate stresses was the pulse-regularization method modified to allow discontinuities across material interfaces. The new system was validated by comparing with conventional slitting on a depleted U-10Mo (DU) fuel plate. Results The new system was applied to two measurements each on six as-fabricated (pre-irradiation) LEU miniature fuel plates. Variations between the measurements at two locations in the same plate were strongly correlated with measured geometrical heterogeneity in the plate—a tilt in the fuel foil. Compressive stresses in the U-10Mo were shown to increase from 20 to 250 MPa as the ratio of aluminum thickness to U-10Mo thickness increased causing increased constraint during cooling. Faster cooling rates during processing also increased stress magnitudes. Conclusions The measurements trends agreed with data in the literature from similar plates made with DU, which further validates the method. Because other methods are impractical in a hot cell, the modified slitting method is now poised for the first measurements of post-irradiation stresses.

Low-enriched ( similar to 19.75% 235 U) U-10Mo(wt%) foils were evaluated for both mechanical and thermophysical properties to establish baseline characteristics prior to irradiation testing. Cast U-10Mo ingots were sequentially hot and cold rolled to two different final foil thicknesses, resulting in different levels of cold work being present in the finished foils. The mechanical response of the U-10Mo foils was evaluated in the temperature range of 293-623 K. Increasing test temperatures resulted in decreases of both yield and ultimate tensile strength. Comparison of results to earlier work performed on depleted U-10Mo rolled foils indicates that the material in this study has lower strength and increased ductility. However, a comparison of the two different foil thicknesses in this study found that they were nearly identical in yield and ultimate tensile strength, differing by only 10-20 MPa ( < 2%). Thermal conductivity was calculated from the measurements of specific heat, linear thermal expansion, and thermal diffusivity from 323 to 1273 K on both thicknesses of the foils. These data were approximately 5-10% lower than previously published data above 573 K and were similar to data for temperature ranges below 573 K. Differences are likely related to differences in grain size and/or impurity content and variation in fabrication history. (c) 2022 Elsevier B.V. All rights reserved.

This article presents the essential aspects of the design and fabrication of nuclear fuel systems utilized in research and test reactors. Fuel attributes of interest include mechanical, physical, and thermal properties as well as overall manufacturability. The article focuses primarily on plate-type fuels utilized in the highest power research reactors where design and function are critical to operational safety and successfully meeting each reactor's unique mission.