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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

XRD and SEM/EDS characterization of two quaternary fuel alloys (U-2.5Mo-2.5Ti-5.0Zr and U-1.5Mo-1.5Ti-7.0Zr in wt. %) for fast reactors

The current study focuses on characterization of two potential fuel alloys for sodium-cooled fast reactors: U-2.5Mo-2.5Ti-5.0Zr (U-MT5Z) and U-1.5Mo-1.5Ti-7.0Zr (U-MT7Z) in wt. %. As-cast alloys and annealed (600°C, 500h) alloys are studied by X-ray powder diffraction (XRD) for their bulk structures, and scanning electron microscopy / energy dispersive X-Ray spectroscopy (SEM/EDS) for their microstructures and phase constituents. For as-cast alloys, the XRD results show that U-MT5Z mainly contains γ, while U-MT7Z alloy contains a. In both alloys, there are three phases indicated by EDS, a U-rich phase, secondary phase, and Zr-rich phase. Further, the elemental compositions varied in different phases and in different alloys. After annealing, both alloys exhibit the structures of α-U and U 2 Ti according to XRD. The EDS results suggest there are five phases found in both alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Powder Characterization Inter-Comparison

We performed qualitative and quantitative image analysis on SEM images for 5 uranium samples. Qualitative assessment was completed using the lexicon of Tamasi et al. 2017 on a subset of images from each sample to provide an overall morphological profile of each material. Quantitative analysis of the particles was done using the Morphological Analysis for Materials Attribution, or MAMA, software. We performed particle analysis primarily on samples labeled U Mo, U Si, and UO 2 . Samples labeled ADU and DU Ox were not prioritized for quantitative analysis due to staffing and time it took to segment these images. Two lab analysts worked on this effort, one focusing on the qualitative assessment and the other focusing on the quantitative assessment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

The role of ternary alloying elements in eutectoid transformation of U-10Mo alloy part II. In and Ex-situ Neutron diffraction-based assessment of eutectoid phase transformation kinetics in U-9.8Mo-0.2X alloy (X = Cr, Ni or Co)

HExploring the effects of minor ternary alloying additions, typically impurity elements, on the phase stability of U-10Mo is important for preventing undesirable phase decomposition during processing or during service. This work examines the influence small ternary additions of Cr, Ni, and Co. Both in-situ and ex-situ neutron diffraction measurements made during and after high temperature (450 – 525°C) exposures were used to better define the influence of these elements on the time-temperature-transformation (TTT) behavior of U-10Mo, providing information which is complementary to electron microscopy investigations of the same alloy systems performed in the first part of this work. Minor additions of Ni and Co decrease the ?-phase stability at all temperatures investigated. Signatures of U6X (X = Ni or Co) compounds were shown to be present in amounts of up to 6 wt% in the heat treated alloys, suggesting that the initial precipitation of this phase may catalyze further ?-phase decomposition. On the other hand, the Cr containing alloys were observed to have nearly the same, and in some cases slower, phase transformation kinetics when compared to the binary U-10Mo control samples. The results of the study have enabled preliminary estimates of the TTT curves for the ternary alloys.

phase transformation, kinetics, discontinuous prec↗
DOE OSTI2021

The effects of Ti content and quenching on phase transformations, microstructures, and mechanical properties in uranium-titanium alloys

In this work, the effect of Ti content on phase transformations, microstructures, and mechanical properties of U-Ti alloys are described for alloys containing 0.3 wt.% to 2.0 wt.%Ti. Rapid cooling is required to overcome diffusional decomposition of γ-phase and facilitate diffusionless transformation to supersaturated variants of α-phase. Critical cooling rate increases with increasing Ti content, opposite to the trend observed in U-Mo and U-Nb alloys. This difference occurs because the martensite transformation temperatures in these relatively dilute U-Ti alloys are above the knee of the C-curve for diffusional decomposition, unlike those in the more concentrated U-Mo and U-Nb alloys. In these U-Ti alloys critical cooling rate depends on the amount of undercooling required to reach M s , which increases with increasing Ti content, and the time for diffusional decomposition to occur just above M s , which decreases with increasing Ti content. The net result is that higher cooling rates are required as Ti content increases. Full quenching results in diffusionless transformation of γ-phase to supersaturated variants of α-phase. Very dilute alloys transform via a γ → β → α m sequence of massive transformations. Martensitic γ → α' a transformation begins at ~0.4%Ti, and 100% α' a microstructures are obtained from ~0.65% to ~1.4% Ti. A transition to banded α'b martensite occurs at ~1.5%Ti. Evidence suggests that the α' a to α' b transition may occur when the cubic γ-phase first transforms to tetragonal γ°, which in turn transforms to orthorhombic α' b via the sequence γ → γ° → α' b . Fully quenched alloys exhibit moderate strengths and ductilities, and their supersaturation with Ti makes them amenable to subsequent age hardening. Subcritical quenching typically results in two-phase microstructures with lower ductilities and near-zero Ti-supersaturation, eliminating the possibility of subsequent age hardening.

36 MATERIALS SCIENCE↗
DOE OSTI2021

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 United States High-Performance Research Reactors (HPRRs). A critical phenomenon of interest with 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 continuum level 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. In this work, updated diffusion coefficients are presented that are applicable under irradiation that incorporate both intrinsic and radiation driven diffusion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Possible impacts of Mo chemical banding and second phase impurities on the irradiation behavior of monolithic U-10Mo fuels

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 compromise include cracks, large fission gas porosity, and interconnection of fission gas pores. For the fission densities evaluated in this work (3.5 × 10 21 fissions/cm 3 -5.1 × 10 21 fissions/cm 3 ), fine porosity can develop 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. Further, 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 overshadowed 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Formation of uranium nitride nanoparticles via mechanical alloying of uranium-molybdenum alloy fuels in gaseous nitrogen

Uranium-molybdenum (U-Mo) alloys show promise as a nuclear fuel system due to their high thermal conductivity and fuel loading capability. However, U-Mo systems are susceptible to irradiation induced swelling ultimately affecting the cladding via mechanical and chemical interactions. To address these shortcomings, this research investigated the formation of uranium mononitride (UN) nanoparticles within a 90 wt% U/10 wt% Mo (U-10Mo) matrix to act as a prospective defect sink for fission products at nanometric hetero-interfaces. To promote the formation of UN, U-10Mo powders were mechanically alloyed under a high purity nitrogen atmosphere. Variations of the milling process investigated included media size, duration of milling, and number of times the milling jar was re-aerated with nitrogen gas. Characterization of the fuel microstructure was completed using light element analysis, X-ray diffraction, scanning and transmission-electron microscopy, electron energy loss spectroscopy, and atom probe tomography. UN nanoparticles measuring 1–5 nm in radius were observed in the U-Mo matrix as early as 1 h into the mechanical alloying process. Milling time in excess of 10 h was found to lead to deleterious effects induced by the stainless-steel milling media.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

An investigation of the failure modes in U-10Mo monolithic fuel irradiated to high burnup

Here, 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Computational determination of a primary diffusion mode in γ U-10Mo under irradiation

Low enriched uranium (< 20 % 235 U)-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 % 23 5U) 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 λU-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 x 10 22 fissions/m 3 /s, which is higher than the typical fission rate density range in research reactors. The temperature regime that radiation-enhanced diffusion of each element dominated was dependent on the fission rate density. Finally, 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.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

First Principle based Thermal Conductivity Modeling of Metal Alloys with Applications on Uranium Alloys

Thermal conductivity is an important materials property related to heat transport, which is essential to many applications, ranging from thermoelectrics to nuclear reactor materials. High- quality thermal conductivity data is critical to these materials and their associated technologies, including the nuclear fuel materials like traditional oxide fuels and metallic uranium (U) fuels, e.g. U-Zr and U-Mo alloys. And thermal conductivity modeling is widely used to interpolate or extrapolate experimental data, to obtain high-quality thermal conductivity data over wide temperature and composition range, and to understand the impacts of different factors including defects and microstructures. However, a general thermal conductivity model for metal alloys which can work regrading different phase components is still missing. Also, for U alloys, a thermal conductivity model working with different phases is needed. Therefore, in this work, we developed thermal conductivity models for metal alloys, based on ab-initio calculations, semi-classical physics rules, and limited experimental data. The goals of these models are to help obtain high- quality thermal conductivity data within a little experimental input as possible, have models that are extendable to varying microstructures, including different types of irradiation effects, and provide mechanistic understanding of heat transfer in the modeled alloys. In this work, our model solves several challenges in the development. The DFT-BTE approach is applied to decrease the reliance on experimental data and make the model extendable to composition changes and defects. A practical and efficient way to combine the DFT inputs with physics rules is pointed out in this work too. A staged approach, which starts from simple cases of elemental metals, then extends to solid solutions, different compound phases, and multi-phase mixtures, is presented to work with the metal alloys in different phase components. Our models are demonstrated respectively on aU for elemental metal model, on U-Zr and U-Mo alloys in aU temperature range for multi-phase mixture model, on high-temperature U-Zr, U-Nb, and U-Mo alloys in the body-centered cubic phase for concentrated solid solution model, and on irradiated U-Mo alloys for irradiated metal alloy model. All models show great agreement with experimental data. In these demonstrations, our model shows its advantages compared to previous empirical fitting model. Our model requires fewer experimental data, due to the inputs from DFT. The quantitative insights into the different physical factors are provided in our model, as it incorporates the electron and phonon scattering mechanisms. Our model also can be extended to incorporate the contributions of point defects, grain boundaries, and noble gas bubbles, to integrate their effects on heat transfer. This model can serve as both a foundation for understanding the more complex thermal conductivity of realistic U alloy fuels and a useful tool to guide further modeling of thermal conductivity to aid materials and device design.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Mechanical Properties of Irradiated U-10wt%Mo Alloy Degraded by Porosity Development

A plate-type nuclear fuel consisting of a solid monolithic foil of U-10wt%Mo is under development for use in the United States’ high performance research reactors. In support of developing this fuel, the fuel has been fabricated for the first time by a commercial fuel vendor and subsequently irradiated in a test reactor. This provides an opportunity to evaluate post-irradiation mechanical properties of commercially fabricated fuel. Four-point bend testing was conducted on the irradiated U-10Mo fuel and the data produced includes bending strength and Young’s modulus. Although the material behaves in a brittle manner due to the developed porosity, a general trend of strength and modulus reduction are found as fission density increases. The data produced is evaluated using both Weibull statistics and a modulus degradation model with recommendations provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Uranium–Molybdenum Alloy Critical Experiments for the Design of the Health Physics Research Reactor

Clean critical experiments with a uranium-molybdenum alloy (average of 10.1616 wt. % Mo with a density of 17.08 g/cm 3 ) were performed at the Oak Ridge Critical Experiments Facility in 1961 to support the design of the Health Physics Research Reactor (HPRR). The HPRR was similar to the Godiva burst reactor at Los Alamos National Laboratory and was designed to produce 50 microseconds burst of 10 17 fission pulses of radiation for dosimetry measurements, initially in support of the determination of the doses from the nuclear detonations in Japan during World War II. These experiments reported here were used to verify the calculational methods used to design the HPRR. These delayed critical measurements were:1) a solid unreflected and unmoderated 8-in.-dimeter U-Mo cylinder, 2) an unmoderated and unreflected annulus with 8-in.-outside diameter, 2-in.-inside diameter cylinder with a central void, 3) an unmoderated and unreflected annulus with 8-in.-outside diameter, 2-in.-inside diameter cylinder with a central void filled with stainless steel, 4) Same as 3) but with 3-in-thick Plexiglas reflector on top with and without cadmium between the reflector and the U-Mo alloy assembly with steel in the center, and 5) an unmoderated and unreflected annulus which was a modification of the second but with the lower 5 inches of the central hole enlarged to 3.5 in. with various reflector conditions. The reflector conditions were: 1-in.-thick Plexiglas on all outer surfaces-void in the center; 1-in.-thick Plexiglas on all outer surfaces-Plexiglas in the center; 2-in.- thick Plexiglas on radial surface-void in the center; 6-in.-thick Plexiglas on the bottom only-Plexiglas in the center; and 6-in.-thick Plexiglas on bottom, 1-in.-thick on top and on the lower 8.25-cm.-section of the radial surface-void in the center. For some of these reflector conditions 0.025-cm.thick cadmium was located between the reflector and the U-Mo alloy. The uranium contained 93.17 wt. % 235 U. Reflection was a safety concern for this unmoderated and unreflected reactor and reduction of reflection effects was also investigated by insertion of neutron absorber around the U-Mo alloy. The stainless steel 304 contained 18% nickel and 8% chromium and the rest iron. The reflector material was a methacrylate plastic (Plexiglas) containing 5.8 x 10 22 atoms/cm 3 of hydrogen and 3.6 x 10 22 atoms/cm 3 of carbon with a density of 1.20 g/cm 3 . The purpose of this report is to document the experimental information for the measurements performed so that at a later date researchers could perform the required uncertainty and calculational analyses and documentation to use these data for an International Nuclear Criticality Safety Benchmark Program (ICSBEP) or a EURATON Nuclear Energy Agency (NEA) benchmark. The data from the experiments described should be acceptable for use as criticality safety benchmark experiments for the ICSBEP and the NEA nuclear criticality safety benchmark program, once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, the uncertainties in k eff could be as low as ±0.0002 for some configurations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Materials Data on UMo by Materials Project

U(Mo) crystallizes in the monoclinic P2_1/c space group. The structure is one-dimensional and consists of four U(Mo) ribbons oriented in the (1, 0, 0) direction. U is bonded in a 2-coordinate geometry to two equivalent Mo atoms. There are one shorter (2.63 Å) and one longer (2.67 Å) U–Mo bond lengths. Mo is bonded in a 2-coordinate geometry to two equivalent U atoms.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Xenon mobility in γ-uranium and uranium–molybdenum alloys

Diffusion in bcc uranium and U–Mo alloys is of great interest because fission gas and other fission products impact the performance of nuclear fuels. We investigate the mobility of xenon and molybdenum in bcc uranium (⁠$γ$-U) and metallic U–Mo alloys by calculating the migration energies of xenon and molybdenum for various local compositions using density functional theory. We also calculate the solute–vacancy binding energies of different solutes to vacancies in bcc uranium. We find that the solute–vacancy binding energy in bcc uranium is significantly higher than it is in other bcc metals (e.g., Fe and W). We also find that the migration energy of molybdenum is substantially higher than the migration energy of xenon, indicating that xenon is much more mobile than molybdenum in bcc uranium. The presence of molybdenum in the nearest-neighbor shell around a xenon atom typically increases the migration energy of xenon, which indicates a reduction of xenon mobility in U–Mo alloys compared to pure bcc uranium.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Purification of U from U-10Mo scrap generated during the fabrication of high performance research reactor fuel

A low enriched U-Mo alloy fuel is under development to replace highly enriched U fuels currently used in United States high performance research reactors. The alloy casting and fuel fabrication processes will generate scrap streams containing low enriched U (LEU) which must be recovered. Solvent extraction processes were designed using the Argonne Model for Universal Solvent Extraction (AMUSE) to purify solutions containing 20 and 50 g/L U. The feed for the solvent extraction processes was prepared from solutions generated from the dissolution of U-10Mo-Zr foils and U-10Mo-Zr-Al mini-plates. The U purification processes were demonstrated using two, 16-stage banks of miniature mixer-settlers. The solvent extraction experiments demonstrated that all design objectives for the U purification processes could be met. The U recovery in the product stream for each flowsheet was ≥99.9%. The flowsheet demonstrations also showed that the purity of the U Product will meet the requirements of the ASTM International C1462-21 specification for LEU metal enriched to less than 20% 235 U. In conclusion, the AMUSE modeling for both flowsheet demonstrations was validated by comparing predicted and measured concentrations of U, Mo, and Zr in the exit streams and stage samples at steady-state conditions in the mixer-settlers.

modified PUREX process↗
DOE OSTI2021

Microstructural Changes and Chemical Analysis of Fission Products in Irradiated Uranium-7 wt.% Molybdenum Metallic Fuel Using Atom Probe Tomography

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.

Bachhav, Mukesh↗
DOE OSTI2025

Entrapment Behavior of Solid Surrogate Fission Products at Engineered UN Nano‐Hetero‐Interfaces Within Metallic Nuclear Fuels

Nanometric hetero-interfaces provide a wealth of scientific and engineering opportunities due to their complex and often misunderstood properties that can differ from their respective bulk constituents. In this work, the ability for engineered nanostructures within a bulk U─Mo alloy to arrest simulant fission products is investigated experimentally and computationally. Nanostructured 90 wt% U/ 10 wt% Mo (U-10Mo) with 7.1 at% Nd is consolidated using spark-plasma- sintering (SPS) techniques and is heat-treated at 500 °C under vacuum for 24, 100, 500, and 1000 h. Analysis on the sintered and heat-treated U-10Mo reveals rapid kinetics in Nd diffusion to nanocluster sites, with evidence of Nd diffusion occurring during sintering and during the following heat-treatment. The segregation behavior of Nd at two different U─Mo/UN interfaces is computationally verified using density functional theory (DFT) to reinforce experimental data.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗