High Performance Research Reactors Research Articles

Taillefer—A Tool for Sensitivity Analysis and Uncertainty Propagation Studies for Steady-State Thermal-Hydraulic Simulations of Involute Fuel Element Research Reactors

Taillefer is a versatile Python tool for carrying out Sensitivity Analysis (SA) and uncertainty propagation (UP) studies based on Monte Carlo sampling. Developed with the primary goal of investigating sensitivities and uncertainties of steady-state thermal-hydraulic (SSTH) safety parameters of the high-performance research reactors Forschungs Neutronenquelle Heinz Maier-Leibnitz (FRM II) in Garching, Germany, and the Réacteur à Haut Flux (RHF) in Grenoble, France, it can also be used for a large variety of other modeling problems. The work presented here aims to explain the underlying mathematical background of SA and UP studies with Taillefer and to show some steps to verify these routines. Furthermore, a real-life application example is provided that demonstrates Taillefer’s use in SSTH analysis of the RHF. For this purpose, Taillefer is coupled to the external thermal-hydraulic software PLTEMP/ANL, which is one of the codes used at FRM II and RHF to access SSTH performance and safety parameters. Determining these crucial quantities is part of identifying possible low-enriched uranium (LEU) core designs that are suitable to replace the currently used highly enriched uranium fuels of the two reactors, supporting global nonproliferation efforts. Taillefer is a powerful tool in these conversion studies, as it increases the reliability of the LEU safety parameters by providing information about sensitivities and uncertainties in addition to the nominal values predicted by the thermal-hydraulic software.

Calculation of grain boundary diffusion coefficients in γU-Mo using atomistic simulations

The γU-Mo alloy has been selected for the conversion of U.S. High-Performance Research Reactor (HPRR) fuel from highly enriched uranium to low enriched uranium as a part of the effort to reduce nuclear proliferation risks. Although γU-Mo alloys have the advantage of high uranium density and good overall irradiation performance, the irradiation-induced swelling and creep in the metal remain important design parameters since they influence the mechanical and thermal integrity of the fuel. To account for these design criteria, engineering scale models need fundamental properties, such as diffusion coefficients, as input. In this study, the diffusion of related species along grain boundaries in γU-Mo fuel is quantified considering that grain boundaries act as sinks for point defects, nucleation sites for gas bubbles, avenues for Coble creep, etc. The diffusivities of U and Mo in selected grain boundaries of γU-Mo alloys (γU-7Mo, γU-10Mo, and γU-12Mo) are obtained utilizing molecular dynamics simulations for a temperature range of 600 K - 1200 K with an interval of 100 K. The structures analyzed include symmetric tilt, asymmetric tilt, and twist grain boundaries. The grain boundary diffusion coefficients of U and Mo in the examined γU-Mo alloys are on the order of 10−14 to 10−11 m2s−1. It is observed that the U diffusivity in the grain boundary is higher than the Mo diffusivity in all cases and that the increase in Mo content of the alloy correlates to a decrease in the grain boundary diffusion. Xe diffusion along γU-10Mo grain boundaries is also calculated in this work, and the diffusivity of Xe in the γU-10Mo grain boundaries is found to be 8 to 15 orders of magnitude higher than the intrinsic Xe diffusivity in γU-10Mo depending on the temperature. The information gathered in this work can inform fuel swelling and creep models and help understand various other phenomena related to γU-Mo fuel performance.

Micro X-ray computed tomography examination of mini plate fuel with hot isostatic pressed aluminum cladding

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.

Fluid structural interaction simulation of the MP-1 plate performance irradiated in the advanced test reactor

Due to the high neutron fluxes they generate and increased heat transfer performance, plate type fuels are used in the U.S. high-performance research reactors. During irradiation, a significant amount of fission energy (i.e., ∼200 MeV per atom) is released by the U-235 chain reaction that is carried away by the coolant. Through thermal hydraulic analysis, the coolant’s heat transfer capability is investigated to ensure that the desired fuel temperature can be maintained. At high temperatures, the plate undergoes elastic/plastic deformation, creep, and swelling as a result of both the temperature gradients and the fission gas production within the fuel. These effects are studied via fuel performance analysis. When the plate deformation is small enough that the coolant flow remains relatively unchanged, conducting these two types of analyses independently will suffice. But at high fuel burnups, the swollen plates may encroach into the coolant channels that separate the fuel plates from each other and cause these channels to narrow. Large reductions in channel gap size imperil cooling performance, causing fuel temperatures to rise. If the plate deformation is asymmetric, the fuel centerline will shift toward one side of the channel, causing an uneven reduction in coolability. In addition, a boehmite (oxide) layer will, over time, grow on the plate surface, further obstructing the heat transfer process. To precisely predict fuel temperatures/deformation, a complete coupled analysis that considers coolant flow, heat transfer, oxide growth, elastic/plastic deformation, creep, and swelling is needed; however, this type of analysis method is not available in the literature. To fill the gap, this research developed a fluid structure interaction (FSI) approach to the fuel plate analysis, then successfully applied it to the Mini-Plate (MP)-1 experiment, which was irradiated in the Advanced Test Reactor (ATR) for both one and two cycles. The complete analysis coupled STAR-CCM+, a computational fluid dynamic (CFD) software for calculating flow, with Abaqus, a finite element analysis code for calculating plate deformation. Improvements in the results were found when comparing the fully coupled analysis to the independently conducted analyses but they were not significant due to the miniature size of the plates and the relatively short irradiation time. In the future, the fully coupled approach presented herein will be applied to full-size fuel plates with longer irradiation cycles once additional experiments become available.

The reconciliation and validation of a combined interatomic potential for the description of Xe in γU-Mo

A U-Mo alloy has been selected as the fuel design for the conversion of high-performance research reactors in the United States. Efforts are ongoing to describe the fuel evolution as a function of time, for a variety of different reactor conditions. The accurate prediction of fuel evolution under irradiation requires the implementation of correct thermodynamic properties into mesoscale and continuum-level fuel performance modeling codes. Molecular dynamics has proven to be a valuable tool to parameterize or inform these higher-length scale models. However, there are currently inaccuracies in the only available U-Mo-Xe potential, which limits the predictive capabilities of molecular dynamics to inform critical phenomena in these fuel systems such as fission gas swelling. This work provides an updated U-Mo-Xe ternary interatomic potential which combines existing potentials in a reconciled format. The validation of the interatomic potential is performed by analyzing the phase stability and vacancy formation energies. Subsequently, Xe solution energies and an equation of state to describe Xe bubbles in U-Mo are calculated, providing 1) evidence of the significant differences between the prior ternary potential and the currently presented potential, and 2) updated data/tools for implementation into mesoscale simulation methodologies to study fission gas bubble evolution.

Innovation and qualification of LEU research reactor fuels and materials

Two projects within the Euratom Research and Training Programmes 2014–2018 and 2019–2020 are focused on the innovation and qualification of novel nuclear fuels for conversion from highly-enriched uranium to low-enriched uranium (LEU) and for securing the supply chain of EU research reactors into the future. The LEU-FOREvER project is drawing to a close and has made significant progress in developing and demonstrating the uranium-molybdenum fuel system, demonstrating the viability of a high-density uranium-silicide fuel for EU high-performance research reactors (BR2, RHF, FRM-II, JHR). This project has significantly increased the fabrication know-how and fuel performance understanding of the uranium-molybdenum and high-density uranium-silicide dispersion fuel systems. Further, a new, innovative and increased performance design for the LVR-15 research reactor fuel assembly has been engineered and a demonstration is planned in 2022. In the EU-QUALIFY project, which began in 2020, the planning of four demonstration irradiation tests has been nearly completed and fabrication development of the various fuel systems is ongoing, including the establishment of an EU monolithic uranium-molybdenum fabrication capability. It is expected that the results of this project will begin or complete the data gathering necessary for generic fuel qualification of the LEU uranium-molybdenum dispersion and monolithic fuel systems, and the LEU high-density uranium-silicide fuel system.

Computational determination of a primary diffusion mode in [formula omitted]U-10Mo under irradiation

Low enriched uranium (< 20 %235U)-molybdenum (U-Mo) monolithic fuel is the primary candidate for high-performance research and test reactors and is in the process of being qualified to replace highly-enriched uranium (≥ 20 %235U) fuel. As part of the qualification process, it is critical to understand and predict the behavior of fission gas bubbles under irradiation, which affects fuel swelling and fuel failure. Mechanistic fuel models are being developed that can both reproduce the existing experimental data for fuel swelling, and be further applied to irradiation conditions beyond the experimental scope. Diffusion of species under irradiation conditions is an important parameter in the mechanistic fuel models; however, no temperature-relevant experimental diffusion data exists. In the present work, radiation-enhanced diffusion coefficients of U, Mo, and Xe in γU-10wt.%Mo were calculated in the temperature range between 300 K and 1400 K via rate-theory models and molecular dynamics simulations with an embedded-atom method interatomic potential for the U-Mo-Xe system. Accordingly, total diffusion coefficients under relevant irradiation conditions are determined using previously obtained intrinsic thermal diffusion and radiation-driven diffusion coefficients, as well as the newly calculated radiation-enhanced diffusion coefficients presented herein. Radiation-enhanced diffusion of U and Mo was dominant in the intermediate temperature range, whereas radiation-enhanced diffusion of Xe did not significantly contribute to total diffusion of Xe at the relevant fission rate densities. Radiation-enhanced diffusion of Xe became faster than both intrinsic thermal diffusion and radiation-driven diffusion at a fission rate density of 5×1022fiss/m3/s, which is higher than the typical fission rate density range in research reactors. The temperature regime where radiation-enhanced diffusion of each element dominated was dependent on the fission rate density. The total diffusion coefficients of U, Mo, and Xe, updated in this work, will be utilized as parameters in the mechanistic fuel models to help predict the behavior of fission gas bubbles under irradiation more accurately.

Multiphase Alloy Nuclear Waste Forms Developed for Pyrochemical U-10Mo Scrap Recovery Waste Streams

The US High Performance Research Reactor (USHPRR) Conversion program is developing an Al-clad Zr-bonded U-10Mo nuclear fuel foil to enable the conversion of research reactors to high-assay low-enriched uranium (HALEU) alloy fuel. Fabrication scrap from the fuel manufacturing will be electrorefined to recover the HALEU, which will generate waste streams consisting of Zr, Mo, and residual U retained in the stainless steel anode basket used in the electrorefiner. Four alloys were formulated with different relative amounts of Zr, Mo, and 316L-SS to represent the expected range of waste compositions. Laboratory-scale ingots were made and metallurgically characterized to assess differences in the microstructures and distributions of the surrogate waste metals. Different amounts of the same dominant phases with similar compositions were formed in most materials: a γ-austenite based matrix, two ZrFe2 intermetallic phases that could host residual uranium, an FeCrMo sigma (σ) intermetallic phase, a chi (χ) phase, and a secondary austenite (γ2) phase. Electrochemical tests were conducted to compare the corrosion behaviors in acidic and alkaline brine solutions for a range of simulated environmental redox conditions. Corrosion of the Mo-rich γ and γ2 solid solutions and the σ and χ intermetallic phases occurred. The multiphase alloy waste form formulated with the lowest molybdenum and highest zirconium contents showed the highest corrosion resistance and has sufficient amounts of ZrFe2 to immobilize trace amounts of residual uranium in the waste stream. The alloy formulation strategy and testing approach will be presented with key results.Work conducted at Argonne National Laboratory is operated for the U.S. Department of Energy, Office of Science under contact DE-AC02-06CH11357.

Minimizing thickness variation in monolithic U-10Mo fuel foil and Zr interlayer during hot rolling: A microstructure-based finite element method analysis

Low-enriched uranium alloyed with 10 wt. % molybdenum (U-10Mo) has been identified as a promising alternative to highly enriched uranium fuel for the United States’ high performance research reactors. The monolithic U-10Mo fuel plate consists of a metallic U-10Mo fuel foil with a 25 µm Zr interlayer and a relatively thick cladding of aluminum alloy 6061. The Zr interlayer is typically applied during the hot co-rolling process, and this process dictates the uniformity of the Zr interlayer. Thickness variation observed in the U-10Mo and Zr interlayer has been attributed to several sources: the initial grain size of the U-10Mo castings, can materials, rolling temperature, inhomogeneous molybdenum content, and porosity in the cast U-10Mo. This thickness variation limits the ability to meet the dimensional specification; thus, a better understanding of the factors causing the nonuniform thickness is needed. In this work, we used a novel, microstructure-based finite element method to model the hot rolling process to address these concerns. Grain microstructures in U-10Mo were tessellated and explicitly considered in the finite element model. Each grain was assigned a random material property to mimic the grain strength variations induced by different grain orientations. Simulations were performed using six steel can thicknesses, four grain sizes, and with or without a Zr interlayer to investigate the influences of those variables on the thickness nonuniformity. The simulation results showed that a thinner steel can and finer U-10Mo grain size reduce thickness variations in both the U-10Mo fuel foil and Zr interlayer. The direct findings from the simulations and analysis can be used to optimize the hot rolling schedule, reduce fabrication defects, and meet the dimensional specifications. The proposed microstructure-based finite element model can be also coupled with experimental microstructure characterization data, images, and models to simulate multi-pass hot rolling.

Non-destructive analysis of swelling in the EMPIrE fuel test

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 (∼21 kW/cm3) and to high fuel particle fission densities (∼6.4 × 1021 fissions/cm3). 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.

Morphological and microstructural characterizations of the fresh fuel plates for the SEMPER FIDELIS in-pile test

The UMo/Al dispersion fuel in plate form is considered for the conversion of high-performance research reactors in Europe. In the framework of the UMo fuel qualification program, the adequate margin of safety performance by considering several technological solutions associated with the fabrication parameters, such as heat-treatment of the UMo particles, coating with a diffusion barrier material and powder size distribution as some examples. All these parameters, along with the effect of the hot-rolling process were evaluated by means of image processing and detailed microstructural characterizations for fresh samples i.e. prior to irradiation tests. Principle macroscopic features of powder batches include the size and shape distributions and coating surface examinations. Microscale investigations explored both the coating and kernel microstructures as well as the interface layer between them. Finally, nanoscale analyses examined the UMo–coating interface. The extensive stresses associated with the hot-rolling process have a significant impact on the deformation of the UMo kernels and the degradation of the coated film. The UMo kernels mostly lost their spherical shape for faceted and elongated shapes whereas three types of film degradations were identified including cracks, chippings and delamination.

Microstructural characterization of as-fabricated monolithic plates with boron carbide, aluminum boride, and zirconium boride burnable absorbers

• Fabrication of aluminum-based research reactor plates containing boron-based burnable absorbers. • Scanning electron microscopy. • Transmission electron microscopy. • Burnable absorber size and shape distribution analysis. 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 AlB 2 , B 4 C, and ZrB 2 . TEM also revealed incomplete bonding at the B 4 C-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 ZrB 2 in its as-rolled condition. The porosity in the ZrB 2 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.

Reactor fuel characterization: Evaluation of multiple techniques for expedited analysis of 235U/238U isotopic ratios in U-10Mo

Uranium (U) enrichment measurements for the U.S. High Performance Research Reactor (USHPRR) conversion program are needed to certify the fuel is <20% 235U (tolerance range: 19.55%–19.95%). The main thrust of this work was to determine whether thermal ionization mass spectrometry (TIMS) and multi-collector inductively coupled plasma mass spectrometry (MC-ICP-MS) U isotopic measurements can be made without chemically separating U from molybdenum (Mo) and other trace metal impurities. Although the data TIMS produces is considered the most accurate and precise, it is not typically used for isotopic analysis of U alloyed with 10 wt% Mo (U-10Mo) materials because of the complex sample preparation required prior to analysis. Results indicate that relative uncertainties on measurements made on chemically separated and unseparated samples was less than 0.03% (2 standard deviations) for this sample set with <0.06% trace metal impurities with no observable biases stemming from the presence of 10% Mo. This methodology provides relief from the bottleneck of separations chemistry, which discourages the use of TIMS and MC-ICP-MS in the case of 235U/238U ratios for U-10Mo, and allows for more rapid, highly accurate and precise determination of the 235U content in final fuel foils by TIMS. Additionally, samples were isotopically characterized for laser ablation multi-collector inductively coupled mass spectrometry (LA-MC-ICP-MS) by both TIMS and MC-ICP-MS in order to utilize them as in-house matrix matched reference materials in the absence of a standard.

Enhancing isotope mixing in U-10Mo downblend castings with electromagnetic stirring

Uranium alloyed with 10 wt percent molybdenum (U-10Mo) is the proposed fuel for use in the United States’ high-performance research reactors. The U-10Mo fuel is fabricated by a two-step casting process that downblends highly enriched uranium with depleted uranium (DU) and/or natural uranium (NU) and subsequently alloys the resulting high-assay low-enriched uranium (HALEU) with Mo. Currently, a two-step casting process is needed to meet the ingots’ U-235 enrichment homogeneity specifications.

An atomistic study of defect energetics and diffusion with respect to composition and temperature in γU and γU-Mo alloys

• Vacancy and self-interstitial formation energies were calculated in γ U and γ U-Mo. • Self-diffusion and interdiffusion coefficients in γ U- x Mo ( x = 7 , 10, 12 wt.%) were calculated. • Results can be used as input parameters in mesoscale and engineering scale fuel modeling. Uranium-molybdenum (U-Mo) alloys are promising candidates for high-performance research and test reactors, as well as fast reactors. The metastable γ phase, which shows acceptable irradiation performance, is retained by alloying U with Mo with specific quenching conditions. Point defects contribute to the atomic diffusion process, defect clustering, creep, irradiation hardening, and swelling of nuclear fuels, all of which play a role in fuel performance. In this work, properties of point defects in γ U and γ U- x Mo ( x = 7, 10, 12 wt. % ) were investigated. Vacancy and self-interstitial formation energies in γ U and γ U- x Mo were calculated with molecular dynamics (MD) simulations using an embedded atom method interatomic potential for the U-Mo system. Formation energies of point defects were calculated in the temperature range between 400 K and 1200 K. The vacancy formation energy was higher than the self-interstitial formation energy in both γ U and γ U- x Mo in the evaluated temperature range, which supports the previous results obtained via first-principles calculations and MD simulations. In γ U- x Mo, the vacancy formation energy decreased with increasing Mo content, whereas the self-interstitial formation energy increased with increasing Mo content in the temperature range of 400 K to 1200 K. The self-diffusion and interdiffusion coefficients were also determined in γ U- x Mo as a function of temperature. Diffusion of U and Mo atoms in γ U- x Mo were negligible below 800 K. The self-diffusion and interdiffusion coefficients decreased with increasing Mo concentration, which qualitatively agreed with the previous experimental observations. Point defect formation energies, self-diffusion coefficients, and interdiffusion coefficients in γ U- x Mo calculated in the present work can be used as input parameters in mesoscale and engineering scale fuel performance modeling.

Evaluating the microstructure and origin of nonmetallic inclusions in as-cast U-10Mo fuel

Low enriched uranium alloyed with 10 wt. % molybdenum (U-10Mo) has been identified as a promising alternative to highly enriched uranium oxide dispersion fuels for use in high performance research and test reactors. Manufacturing U-10Mo alloy fuel involves several complex thermomechanical processing steps and understanding of the microstructure and its evolution throughout the various fabrication steps is critical to enable the deployment of a reliable fuel production capability. Nonmetallic inclusions are often found in U castings and may affect subsequent fuel processing steps and microstructure evolution. Yet, the origin of these inclusions is not well established. To elucidate their origin and formation mechanisms, nonmetallic inclusions in U-10Mo castings were characterized on sub-nanometer to millimeter scale. Inclusion distribution, morphology, size, and composition were determined for the metallic fuel samples. Inclusions were estimated to comprise ~ 0.4 % of the fuel (by area), were identified at both grain boundaries and grain interiors, and were found to have varying morphologies (e.g., core-shell, elongated, blocky). All inclusions were either uranium carbides or oxides, or a combination of the two (i.e., dual-phase inclusions). Analysis of inclusions via atom probe tomography revealed that carbides and oxides were hypostoichiometric, with minor amounts of additional impurity elements present (e.g., Si, H). All analyzed inclusions were found to be enriched to ~20 at. % 235U, consistent with the surrounding γ-UMo matrix and target enrichment for the low enriched U fuel, indicating that the inclusions formed during the downblending of highly enriched U metal with depleted U via the melting and casting processes.

Thermal properties of U-Mo alloys irradiated under high fission power density

A variety of physical and thermal property measurements made on irradiated U-Mo alloy monolithic fuel samples with a Zr diffusion barrier and clad in Al alloy 6061 are reported. These properties form the basis for additional investigations in thermal behavior of U-Mo fuel under high performance research reactor irradiation conditions. Measurements were performed on samples harvested from fuel plates irradiated under high fission power density and surface heat flux and complement previous measurements on samples subjected to moderate fission power density and surface heat flux. Sample measurements reported in this work represented fission density ranging from 2.42 – 4.81 × 1027 fissions•m−3 (20.8 – 36.5 235U depletion) and average plate surface temperatures from 355 – 414 K. Specific heat capacity of irradiated monolithic U-Mo increases with increasing temperature and fission density, but carries large uncertainties associated with thickness measurements used to determine the mass fraction of the layer constituents for extraction of the U-Mo specific heat capacity, specifically the very thin and nonuniform Zr diffusion barrier. Density of irradiated monolithic U-Mo decreases with increasing fission density resulting from fission gas swelling and porosity formation. Thermal diffusivity of irradiated monolithic U-Mo increased with increasing temperature and decreased with increasing fission density. Thermal conductivity of irradiated monolithic U-Mo increased with increasing temperature, and in general, appeared to decrease and become less sensitive to temperature with increasing fission density. Thermal conductivity measurements compared to a semi-empirical model agreed within ±30%. It is suspected that the main driver in deviation between model calculations and measurements is the fission gas swelling determination, which does not account for fission power density or irradiation temperature.

Swelling of U-Mo Monolithic Fuel: Developing a Predictive Swelling Correlation under Research Reactor Conditions

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.

Radiation driven diffusion in γU-Mo

A monolithic fuel design based on a U-Mo alloy has been selected as the fuel type for conversion of the U. S. High-Performance Research Reactors. A critical phenomenon of interest regarding U-Mo monolithic fuel is the large amount of swelling that takes place during operation, particularly at high fission densities. The accurate prediction of fuel evolution under irradiation requires implementation of correct thermodynamic and kinetic properties into mesoscale and engineering fuel performance modeling codes. One such property where there exists incomplete data is the diffusion of relevant species under irradiation. Fuel performance swelling predictions rely on an accurate representation of diffusion in order to determine the rate of fission gas swelling and the local microstructural evolution. In this work, we present molecular dynamics simulations of the radiation driven diffusion of U, Mo, and Xe in U-Mo nuclear fuels. Diffusion coefficients for each species are determined over a range of temperatures and compositions. Updated diffusion coefficients are presented that are applicable under irradiation and incorporate both intrinsic and radiation driven diffusion.

Transition Core Analysis for HEU to LEU Fuel Conversion at the University of Missouri Research Reactor

A new type of low-enriched uranium (LEU) fuel based on an alloy of uranium and molybdenum is expected to allow the conversion of U.S. domestic high-performance research and test reactors requiring high density fuel from highly enriched uranium (HEU) to LEU. The University of Missouri Research Reactor (MURR®) has undergone design and performance calculations for conversion to this LEU fuel. Presented in this paper is the analysis of a crucial step in the conversion process: the sequence of MURR transition cores from all fresh to equilibrium burnup LEU operations. During the initial conversion from HEU to LEU fuel, MURR will operate atypically due to the lack of burned LEU elements. Given the constraints of MURR operation and experiments, a proposed transition scheme minimizes the time MURR operates atypically compared to the prototypic cycles currently run with HEU fuel and moves quickly to the same sort of equilibrium cycles for the LEU fuel.