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Interdiffusion and reaction between Al and Zr was investigated as functions of Zr purity, temperature, and time, using Al versus Zr solid-to-solid diffusion couple annealed in the temperature range from 425 °C to 475 °C. All diffusion couples were observed to develop two intermetallic layers, i.e., Al 3 Zr and Al 2 Zr. The Al 3 Zr phase grew with planar morphology, while the Al 2 Zr phase developed a non-planar interfacial morphology. Growth rate and integrated interdiffusion coefficients were determined using Wagner’s approach for each phase. Purity of Zr had a significant effect on the development of Al 3 Zr and Al 2 Zr phases. Diffusion couples with low-purity Zr (i.e., 99.2%) exhibited a higher growth rate for the Al 3 Zr, at the expense of Al 2 Zr growth. Couples with low purity Zr also resulted in a higher degree of non-planarity for Al 2 Zr phase. In general, degree of non-planarity increased with an increase in anneal time, while it decreased with an increase in temperature. Non-planar morphology was simulated based on 3-D anisotropic diffusion using diffusivity tensor (quadric), and needle-like growth, similar to that observed by experiments were simulated.

A low-enriched uranium U-10Mo monolithic nuclear fuel is being developed by the Material Management and Minimization Program, earlier known as the Reduced Enrichment for Research and Test Reactors Program, for utilization in research and test reactors around the world that currently use high-enriched uranium fuels. As part of this program, reactor experiments are being performed in the Advanced Test Reactor. It must be demonstrated that this fuel type exhibits mechanical integrity, geometric stability, and predictable behavior to high powers and high fission densities in order for it to be a viable fuel for qualification. This paper provides an overview of the microstructures observed at different regions of interest in fuel plates before and after irradiation for fuel samples that have been tested. These fuel plates were fabricated using laboratory-scale fabrication methods. Observations regarding how microstructural changes during irradiation may impact fuel performance are discussed.

Monolithic fuel plates have been developed utilizing low enriched U alloyed with 10 wt.% Mo to replace highly enriched fuels in research and test reactors, in accordance with the goals of the Materials Management and Minimization Reactor Conversion Program. The fuel plates consist of U10Mo fuel, Zr diffusion barrier, and AA6061 cladding. They are fabricated by co-rolling the U10Mo and Zr, which are then encapsulated via hot isostatic pressing of the entire U10Mo/Zr/AA6061 assembly. During fabrication, the metal constituents of the fuel plates undergo phase transformations as well as interdiffusion and reactions at interfaces. The areas of interest are the U10Mo fuel, U10Mo/Zr interface, U10Mo/AA6061 interface, Zr/AA6061 interface, and AA6061-AA6061 bond line. Knowledge of the transformations and growth in the plates is necessary to optimize fabrication parameters and predict behavior as they relate to irradiation performance. Numerous studies have been conducted to analyze these reactions in monolithic fuel plates, and a summary of their observations is provided in this paper.

U-Mo alloys are being developed as low enrichment uranium fuels under the Reduced Enrichment for Research and Test Reactor (RERTR) Program. In order to understand the fundamental diffusion behavior of this system, solid-to-solid pure U vs Mo diffusion couples were assembled and annealed at 923 K, 973 K, 1073 K, 1173 K, and 1273 K (650 °C, 700 °C, 800 °C, 900 °C, and 1000 °C) for various times. The interdiffusion microstructures and concentration profiles were examined via scanning electron microscopy and electron probe microanalysis, respectively. As the Mo concentration increased from 2 to 26 at. pct, the interdiffusion coefficient decreased, while the activation energy increased. A Kirkendall marker plane was clearly identified in each diffusion couple and utilized to determine intrinsic diffusion coefficients. Uranium intrinsically diffused 5-10 times faster than Mo. Molar excess Gibbs free energy of U-Mo alloy was applied to calculate the thermodynamic factor using ideal, regular, and subregular solution models. Based on the intrinsic diffusion coefficients and thermodynamic factors, Manning’s formalism was used to calculate the tracer diffusion coefficients, atomic mobilities, and vacancy wind parameters of U and Mo at the marker composition. The tracer diffusion coefficients and atomic mobilities of U were about five times larger than those of Mo, and the vacancy wind effect increased the intrinsic flux of U by approximately 30 pct.

Microstructural characterization using scanning electron microscopy and scanning transmission electron microscopy (TEM/STEM) was carried out near the interface between the two AA6061 alloys that were hot isostatically pressed (HIP) to clad Zr laminated U-10 wt.% Mo metallic nuclear fuel. The HIP-bonded AA6061–AA6061 interface consisted of discontinuous layer of Mg 2 Si along with traces of fine MgO dispersoids and small precipitates of Al 19 (Fe, Cr, Cu) 4 MnSi 2 . To examine the presence of statistical variation, quantitative microscopy was also conducted, using several HIP’ed samples, to measure the relative linear density of the Mg 2 Si precipitates at the HIP bonded AA6061–AA6061 interface. In order to better understand the formation of discontinuous Mg 2 Si layer, solid-to-solid diffusion couple experiments were carried out using temperature and time relevant to HIP. The discontinuous Mg 2 Si layer was not observed in diffusion couples that were rapidly water-quenched, but those slowly cooled in air and in furnace developed the discontinuous Mg 2 Si. Presence of oxygen, confirmed by electron energy loss spectroscopy via STEM, at the interface would be the potential driving force for the migration of Mg and Si atoms, where Mg would preferentially react with oxygen to form MgO, and excess Mg would react with Si to form Mg 2 Si during cooling. Faster cooling after HIP may minimize the formation of excessive Mg 2 Si.

Decomposition of metastable, isotropic, body-centered cubic γ-phase in the U-10 wt pct Mo alloys with varying Zr contents was investigated as function of heat treatment parameters (i.e., temperature and time) relevant to the fabrication of monolithic fuel plate in development for research and test reactors. Phase constituents and microstructural characterization was performed using X-ray diffraction via Rietveld refinement, scanning electron microscopy with quantitative image analysis, and analytical transmission electron microscopy. Results were compared with the amount of equilibrium phases calculated using the lever rule from the available ternary phase diagrams. After sequential three-step heat treatment (900 °C for 168 hours, 650 °C for 3 hours, and 560 °C for 1.5 hours) decomposition of γ-phase was observed in alloys containing 2 wt pct or higher Zr. Decomposition of γ-phase occurred by the depletion of Mo in γ-phase due to the formation of Mo2Zr. Formation of Mo2Zr at 900 °C and 650 °C produced γ-phase with low-Mo content which in turn promoted a faster onset of eutectoid decomposition at 560 °C. The U-10Mo-4Zr and U-10Mo-10Zr alloys had the highest amount of Mo2Zr, and exhibited the largest amount of γ-phase decomposition after the heat treatment at 560 °C. Cooling rate from the heat treatment at 560 °C did not influence the phase constituents of the U-10Mo-xZr alloys, but the slower cooling promoted the better-defined lamellae microstructure associated with eutectoid decomposition. Trace amount of δ-phase was also observed in the microstructure after the heat treatment at 560 °C, presumably due to local inhomogeneity in alloy compositions.

Monolithic fuel system with U – 10 wt.% Mo (U10Mo) fuel alloy has been developed for the Materials Management and Minimization reactor conversion program to replace highly-enriched fuels in research and test reactors with low-enriched fuels. Interdiffusion and phase transformations in the system constituents, i.e., U10Mo fuel, AA6061 cladding, and Zr diffusion barrier, have been investigated using fuel plates fabricated via rolling and hot-isostatic pressing. Diffusion couples, utilizing the constituents of the fuel system were also carried out to help understand the findings from fuel plates based on phase equilibria and diffusion kinetics. Findings from both fuel plates and diffusion couples can provide a comprehensive knowledge to assess, model, and predict the performance of monolithic low-enriched fuel system from fabrication to irradiation. This paper summarizes the experimental results reported from characterization of the fuel plates and diffusion couples with emphasis on interactions at the fuel-cladding, fuel-diffusion barrier, cladding-diffusion barrier, and cladding-cladding interfaces. Constituent phases and relevant diffusion kinetics are compared and contrasted, taking into account differences in thermodynamics and kinetics variables such as pressure, temperature, and cooling rate.

Nuclear energy, through the use of light water reactors (LWRs), provides a significant percentage of carbon-free electricity in the United States. In the short term, it is the standard boiling water reactor or pressurized water reactor that will carry the load for nuclear-generated electricity in the United States. In the longer term, it will be more advanced reactors that will contribute significantly to worldwide electricity generation. These new advanced reactors will take advantage of the development of more robust fuels and materials that can be employed in more aggressive environments, safely and efficiently. This collection of JOM articles, sponsored by the TMS Nuclear Materials Committee, touches on some recent developments in the areas of nuclear fuels and materials development, processing, characterization, testing, and performance in the context of advancing the use of nuclear energy.

U-Mo dispersions in Al-alloy matrix and monolithic fuels encased in Al-alloy are under development to fulfill the requirements for research and test reactors to use low-enriched molybdenum stabilized uranium alloy fuels. Significant interaction takes place between the U-Mo fuel and Al during manufacturing and in-reactor irradiation. The interaction products are Al-rich phases with physical and thermal characteristics that adversely affect fuel performance and result in premature failure. Detailed analysis of the interdiffusion and microstructural development of this system was carried through diffusion couples consisting of U-7 wt pct Mo, U-10 wt pct Mo and U-12 wt pct Mo in contact with pure Al, Al-2 wt pct Si, and Al-5 wt pct Si, annealed at 823 K (550 °C) for 1, 5 and 20 hours. Scanning electron microscopy and transmission electron microscopy were employed for the analysis. Diffusion couples consisting of U-Mo in contact with pure Al contained UAl 3 , UAl 4 , U 6 Mo 4 Al 43 , and UMo 2 Al 20 phases. Additions of Si to the Al significantly reduced the thickness of the interdiffusion zone. The interdiffusion zones developed Al- and Si-enriched regions, whose locations and size depended on the Si and Mo concentrations in the terminal alloys. In these couples, the (U,Mo)(Al,Si) 3 phase was observed throughout the interdiffusion zone, and the U 6 Mo 4 Al 43 and UMo 2 Al 20 phases were observed only where the Si concentrations were low.

Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.