Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “uranium (U)”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

The porosity surrounding carbides and second phase stringers in monolithic U-10Mo fuel plate after irradiation

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

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Infrared spectroscopic method for uranium isotopic analysis

A high performance infrared (HPIR) system was developed and demonstrated for the infrared absorption analysis of {sup 235}U and {sup 238}U isotopes in uranium hexafluoride gas samples. Sweeping the quantum cascade laser light source over the spectral range and sampling via a high-rate analog-to-digital converter provided 0.0005 cm-1 spectral resolution, which allowed for high-precision measurements of the isotopic peak shift. A data analysis method was developed using principal component analysis to predict the isotope weight % content of {sup 235}U. The HPIR precision, accuracy, and error were evaluated for a wide range of isotopic ratio samples (0.287 – 93.7 weight % {sup 235}U), and the results were compared to the International Target Values (ITVS) set forth by the International Atomic Energy Agency (IAEA) for non-destructive and destructive analytical techniques. The method meets or surpasses the IAEA ITVs for non-destructive analysis of samples with isotopic content of depleted to highly enriched. The results also demonstrated the capability of the HPIR system to correctly predict the {sup 235}U weight % content of a mislabeled sample whose isotopic distribution was validated by mass spectroscopic measurements. The HPIR measurement is nondestructive and, thus, allows for confirmatory analyses of the exact sample at a designated IAEA lab if higher-resolution or a certified analysis is needed. (author)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Developing New Fuels for High Performance Research Reactors

The National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) program works globally to minimize the civilian use of highly enriched uranium (HEU), a weapon-usable nuclear material. Supporting this effort, M3’s Office of Reactor Conversion and Uranium Supply is developing new fuels capable of converting research reactors from HEU fuel to high-assay low-enriched uranium (HALEU) fuel. Some of the remaining research and test reactors (RTRs) operating on HEU today have unique designs, fuel configurations, and demanding performance requirements that cannot be met with an existing regulatory-approved low-enriched uranium (LEU) fuel. M3, DOE’s national laboratories, and other industry partners are qualifying new high-density LEU fuels to convert these RTRs while maintaining their unique capabilities supporting a wide variety of science and technology research in areas such as medicine, industry, defense, education, and training. Current efforts are focused on two options for the remaining US high-performance research reactor conversions: a monolithic uranium 10wt% molybdenum (U-10Mo) fuel form and a dispersion uranium silicide fuel form. This paper reviews the history and status of M3’s fuel qualification efforts for the U-10Mo LEU fuel form.

Montgomery, Rose [ORNL] (ORCID:0000000286038936)↗
DOE OSTI2022

Molecular dynamics studies of grain boundary mobility and anisotropy in BCC γ-uranium

Grain morphologies such as grain size and aspect ratio in uranium-based metallic fuels are important microstructural features that can impact various fuel performance properties such as fission-gas-induced swelling, thermal transport, high burnup structure formation, and radiation resistance. Accurate prediction of the fuel grain morphologies requires knowledge of critical grain growth parameters such as grain boundary (GB) mobility and anisotropy. In this work, molecular dynamics simulations were performed to study the GB mobility and its anisotropy in pure body-centered-cubic (BCC) γ uranium. Nine GBs with different combinations of misorientation angles (20°, 30°, 45°) and rotation axes (<100>, <110>, <111>), as well as an additional <111> 38.2° GB were studied using three interatomic potentials. Here, it is found that the GB mobility anisotropy has complex trends, depending on both rotation axis and misorientation. However, in general the <110> rotation axis has the fastest GB mobility at the same misorientation. The results of this work can be used as not only baselines for future studies of GB mobility in uranium-based alloys such as uranium-molybdenum (U-Mo) fuels, but also input for mesoscale modeling of grain growth in uranium-based alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

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/cm 3 ) and to high fuel particle fission densities (~6.4 × 10 21 fissions/cm 3 ). These conditions mimic the bounding conditions of the BR-2 reactor in Belgium, where a concurrent irradiation experiment was performed, and exceed those previously explored in dispersion U-Mo fuel plates. A local fuel swelling analysis, as determined through high-fidelity, post-irradiation mini-plate profilometry, was used along with statistical methods to non-destructively evaluate the overall performance and separate the effects of convoluted fabrication variables. While some effects observed with this non-destructive analysis were subtle, others had more significant, and possibly competing, effects on the fuel swelling behavior. In closing, these observations will be examined further with destructive examinations to more fully assess them as the fuel design is developed and qualified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Instability of U 3 Si 2 in pressurized water media at elevated temperatures

Following the Fukushima Daiichi accident, significant efforts from industry and the scientific community have been directed towards the development of alternative nuclear reactor fuels with enhanced accident tolerance. Among the proposed materials for such fuels is a uranium silicide compound (U 3 Si 2 ), which has been selected for its enhanced thermal conductivity and high density of uranium compared to the reference commercial light water reactor (LWR) nuclear fuel, uranium oxide (UO 2 ). To be a viable candidate LWR fuel, however, U 3 Si 2 must also demonstrate that, in the event of this fuel coming in contact with aqueous media, it will not degrade rapidly. In this contribution, we report the results of experiments investigating the stability of U 3 Si 2 in pressurized water at elevated temperatures and identify the mechanisms that control the interaction of U 3 Si 2 under these conditions. Our data indicate that the stability of this material is primarily controlled by the formation of a layer of USiO 4 (the mineral, coffinite) at the surface of U 3 Si 2 . The results also show that these layers are destabilized at T > 300 °C, leading to the complete decomposition of U 3 Si 2 and its pulverization due to its full oxidation to UO 2 .

36 MATERIALS SCIENCE↗
DOE OSTI2022

Density functional theory investigations into the magnetic ordering of U 3 O 8

Density functional theory (DFT) has been highly successful in supporting experimental materials science; however, a correct electronic ground state is required to realize the full theoretical capacity of DFT. The uranium oxides, α–U 3 O 8 in particular, are simultaneously technologically important materials and theoretically challenging for DFT because the uranium magnetic ground state is not obvious. This is true for both experiment and theory—magnetic susceptibility measurements indicate an antiferromagnetic (AFM) ground state with transitions near 4.2 and 8.0 K, but the ordering itself is not known. Theoretical literature reports are in contradiction, with independent studies finding paramagnetic, ferromagnetic (FM), and AFM states as the lowest energy configuration. However, recent inelastic neutron scattering experiments suggested an uninvestigated magnetic structure with ordering along the [0.5 1 1] plane, motivating a theoretical reinvestigation. Using this insight, we calculated the relative energy of FM and AFM orderings along [0.5 1 1], [0.5 0 0], [0 1 0], and [0 0 1] using noncollinear DFT calculations with spin-orbital coupling. In this work, we found that the [0.5 1 1] AFM structure is lower in energy than FM or AFM orderings along the low Miller index directions. We also investigated polarization of the magnetic moment along each lattice vector and found that polarization along the out-of-plane direction is the energetically preferred orientation for the AFM structures. Additionally, we found in all calculations that moments initially pointing along the in-plane lattice vectors significantly relax until they point along the coordinate between the two crystallographically distinct uranium sites with complex noncollinear magnetic configurations. The new [0.5 1 1] AFM magnetic structure provides an additional path forward toward understanding the electronic structure of α-U 3 O 8 and lends theoretical credibility to recent neutron scattering results.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Automated Calculation of U-10Mo Fuel and Zr Cladding Thickness

Hand calculation and verification of cladding and fuel thickness in a hot-isostatically pressed uranium-molybdenum (U-10Mo) alloy can incur human errors and longer image processing time when analyzing a large set of cross-sectional images. To help alleviate both time cost and errors made in hand measurements, an automated image processing procedure was developed using Octave, an open-source MATLAB alternative, to repeatedly determine the thickness of Zr and U-10Mo layers. U-10Mo specimens were imaged using standard secondary electron and backscattered-electron imaging at 250× magnification to capture the various layers present in U-10Mo. Further image processing used pixel-by-pixel calculation of the Zr and U-10Mo layers to gather quantitative statistics on the thickness variations associated with each of the layers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 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 OSTI2021

U M subshell X-ray emission spectroscopy of uranium dioxide: the effect of excitation energy

There are very great differences between the non-resonant U M subshell X-ray Emission Spectra of Uranium Dioxide generated with low and high-energy excitation. Additionally, these different spectra are each supported by theoretical constructs. It will be shown that the differences between these experimental spectra can be reconciled and the utility of both theoretical approaches confirmed, including the use of FEFF in the low energy and high-energy regimes.

36 MATERIALS SCIENCE↗
DOE OSTI2021

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. This work demonstrates the ability to provide more homogeneous U-235 distributions during downblend casting by using zirconia for crucibles rather than graphite. When graphite is used as a crucible, the electromagnetic field produced by the induction heater couples directly with the graphite. Zirconia is nonconducting in this study, and therefore does not couple with the field. The induction field couples directly with the metal, and causes electromagnetic stirring (EMS) in the molten metal pool. Eight downblend casting experiments were carried out in this work. Four were performed with zirconia crucibles and four with graphite crucibles. The U-235 enrichment was measured at nine discrete points in each casting using laser ablation multi-collector inductively coupled mass spectrometry (LA-MC-ICP-MS). The enrichment homogeneity was approximately eight times better in zirconia crucible castings than in graphite crucible castings as measured by the plate enrichment range and enrichment coefficient of variation. The results show a single casting step could be used to meet USHPRR enrichment specifications if electromagnetic stirring is present during casting. A related investigation was carried out to determine whether it is appropriate to measure enrichment on as-cast ingots or whether heat treated specimens should be used to accurately characterize enrichment. The results show that the as-cast microstructure plays a significant role in enrichment homogeneity. It is recommended that all enrichment measurements be done after the homogenization heat treatment typical of U-10Mo processing.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds

The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. In conclusion, the DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fission Induced Radiolysis of Uranyl Sulfate Solutions

At Argonne National Laboratory, an experimental pilot facility is being tested for the production of the medical isotope molybdenum-99 ( 99 Mo). AMORE (Argonne Molybdenum Research Experiment) is an experimental pilot facility in which a 40 MeV electron beam from the LINAC strikes a depleted uranium (DU) target. The target generates high energy neutrons, which are thermalized in an 18-liter uranyl sulfate solution of low enriched uranium (LEU). Thermalized neutrons fission the uranium-235 ( 235 U) to produce 99 Mo as well as many other fission products. The 99 Mo is recovered from the uranium solution and purified. The goal of our work is to reduce the use of highly enriched uranium (HEU) for the production of this isotope.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Mechanochemical α to β phase transition of U 3 O 8

Recognized as the most stable phase of uranium oxide, α-U 3 O 8 is widely used throughout the nuclear fuel cycle for safe storage and transportation. There is, however, a lack of understanding of the thermodynamic relationship between α-U 3 O 8 and its polymorph, β-U 3 O 8 . This research contributes new knowledge regarding the kinetics and mechanism of a previously described mechanochemical phase transition between the α- and β-U 3 O 8 polymorphs to improve the understanding of their thermodynamic relationship. In this work, tumble milling of α-U 3 O 8 using different milling media shows an ingrowth of the β-phase over time, with an observed correlation between media density, percent conversion and lattice strain. Anisotropic peak broadening observed within collected X-ray diffraction powder patterns suggests the preservation of uranium polyhedral layered sheets throughout the milling process. Preservation of the sheets implies a shear-induced slip mechanism occurring through in-plane shifting of uranium polyhedra along lattice planes perpendicular to the polyhedral sheets.

Roach, Jordan M. [Oak Ridge National Laboratory (O↗
DOE OSTI2022

Optical vibrational spectra of β -U 3 O 8

Technogenic uranium oxides are ubiquitous in the nuclear fuel cycle. Identifying the chemical composition of these oxides is frequently done with optical vibrational spectroscopy (infrared and Raman spectroscopy). The α phase of U 3 O 8 is the most stable and common uranium oxide, but a β phase of U 3 O 8 can be produced via very slow cooling of the α phase from elevated temperatures or the application of small shear forces. However, optical spectra of this compound have not been reported. In this work, we use high-resolution attenuated total reflection infrared spectroscopy and Raman spectroscopy to collect the optical vibrational spectra. Despite similarities in the crystal structure of α and β – U 3 O 8 , the optical spectra of these compounds differ substantially, with additional Raman active modes in β – U 3 O 8 that are likely originating from differences in the U–O coordination units. Measurements presented here provide spectral characterization that can be used for chemical identification in future studies of uranium oxides.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Electropolishing of a Full-Sized U-10Mo Plate

Electropolishing is used to remove material from the surface of a metal using electric potential and current. U-10Mo fuel is produced from a low enriched uranium plate alloyed with 10% molybdenum (U-10Mo). Electropolishing is being considered in two steps of the process of fabricating this U-10Mo fuel. First, it could be used as a means initially improving the surface finish of the cast plate while removing smut from the casting process from the surface prior to homogenization. Second, it could be used again after homogenization to remove oxidation and continue to improve the surface finish prior to hot rolling. The purpose of this work is to demonstrate the process on full-size ingots and optimize the process. A depleted uranium U-10Mo plate was electropolished three times: the first two times were prior to homogenization and the third after a homogenization step. The first two polishes decreased the overall roughness, as measured by laser confocal microscopy. After each polish, a smoother and shinier blue-gold surface was left on the plate. Upon homogenization, the surface had a more matte appearance and the surface roughness increased back to the pre-homogenization values obtained prior to the initial polish. After electropolishing the homogenized pieces, the changes in surface roughness increased, as measured by laser confocal microscopy. The final polishing step was not long enough to return the surface roughness to pre-homogenization conditions, implying less polishing is needed before the homogenization versus afterwards. Heat generation during the polish was addressed by circulating the electrolyte solution through an external heat exchanger; however, solubility and anodization challenges require further study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Materials Data on U by Materials Project

U is beta Uranium structured and crystallizes in the tetragonal P4_2nm space group. The structure is three-dimensional. there are eight inequivalent U sites. In the first U site, U is bonded to twelve U atoms to form distorted edge-sharing UU12 cuboctahedra. There are a spread of U–U bond distances ranging from 2.89–3.46 Å. In the second U site, U is bonded in a 2-coordinate geometry to eleven U atoms. There are a spread of U–U bond distances ranging from 2.70–3.45 Å. In the third U site, U is bonded in a 1-coordinate geometry to eleven U atoms. There are a spread of U–U bond distances ranging from 2.56–3.36 Å. In the fourth U site, U is bonded in a 1-coordinate geometry to fourteen U atoms. There are a spread of U–U bond distances ranging from 2.98–3.37 Å. In the fifth U site, U is bonded in a 12-coordinate geometry to twelve U atoms. There are a spread of U–U bond distances ranging from 2.68–3.32 Å. In the sixth U site, U is bonded in a 3-coordinate geometry to fourteen U atoms. There are one shorter (3.12 Å) and four longer (3.61 Å) U–U bond lengths. In the seventh U site, U is bonded in a 3-coordinate geometry to fourteen U atoms. There are a spread of U–U bond distances ranging from 2.68–3.32 Å. In the eighth U site, U is bonded in a 3-coordinate geometry to fourteen U atoms. There are a spread of U–U bond distances ranging from 2.68–3.61 Å.

36 MATERIALS SCIENCE↗