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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

A microfluidic electrochemical cell for studying the corrosion of uranium dioxide (UO 2 )

We have developed a specialized microfluidic electrochemical cell that enables in situ investigation of the electrochemical corrosion of microgram quantities of redox active solids. The advantage of downscaling is the reduction of hazards, waste, expense, and greatly expanding data collection for hazardous materials, including radioactive samples. Cyclic voltammetry was used to monitor the oxidation–reduction cycle of minute quantities of micron-size uraninite (UO 2 ) particles, from the formation of hexavalent uranium (U(VI)), U 3 O 7 and reduction to UO 2+x . Reaction progress was also studied in situ with scanning electron microscopy. The electrochemical measurements matched those obtained at the bulk-scale and were consistent with ex situ characterization of the run products by X-ray photoelectron spectroscopy, scanning transmission electron microscopy, and atomic force microscopy; thus, demonstrating the utility of the microfluidic approach for studying radioactive materials.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Speciation mapping of the oxidation layer on aged uranium dioxide using scanning transmission x-ray spectromicroscopy

In this study, UO 2 was aged in humid air and prepared as a thin section using a focused ion beam (FIB) instrument. The specimen was measured using synchrotron radiation spectromicroscopy techniques at the Beamline 11.0.2 STXM end station of the Advanced Light Source (ALS). Non-negative matrix factorization (NMF) methods were used to identify and map three component x-ray absorption near-edge structure (XANES) spectra in the oxygen K-edge data, revealing a surface layer of U 4 O 9 with a thickness of 206 ± 21 nm, and the bulk of the sample remaining as UO 2 . Uranium N 4,5 -edge XANES spectromicroscopy supports these results. The diffusion-controlled parabolic rate constant for UO 2 oxidation to U 4 O 9 was calculated from the observed layer thickness and compared to literature values. Complementary transmission electron microscopy (TEM) was used to image the sample and identify the phases present in various regions, confirming the STXM results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Hydrochlorination of Uranium Dioxide in a Molten Salt Mixture - Phase 1: Tube Furnace Cross Flow Experiments

In 2023, Metatomic® Inc., a South Carolina based company, was awarded a Gateway for Advanced Innovation in Nuclear (GAIN) research voucher for a proposed series of experiments aimed at demonstrating the viability of a spent nuclear fuel (SNF) recycling process patented by Met atomic® Inc. For the GAIN voucher, Metatomic® Inc. selected Savannah River National Laboratory (SRNL) as a partner in executing the proposed proof-of-concept experiments. This report outlines the proof-of-concept experiments performed by SRNL for Metatomic® Inc. during Phase 1 (of 2) experimentation. The Phase 1 hydrochlorination experiments consisted of weighing UO2 into alumina crucibles with a eutectic mixture of NaCl and CsCl, heating the uranium/salt mixture to a varied temperature (550, 650, or 750 °C), and flowing anhydrous hydrogen chloride (AHCl) gas across the surface of the uranium-bearing molten salt mixture. The hydrochlorination process conditions were maintained for 4-5 hours and the percent conversion for each batch of UO2 was determined using a suite of analytical characterization techniques. The degree of UO2 conversion was found to be greatest at the highest tested temperature, ultimately achieving 32.9% conversion of UO2 to water soluble uranium-chloro species (e.g., UO2Cl2, UCl4) after exposure to AHCl for 4.25 hours at 750 °C.

Nguyen, Vinh T. [Savannah River National Laborator↗
DOE OSTI2025

Chromium-doped uranium dioxide fuels: A review

UO 2 doped with parts per million CR 2 O 3 powder is considered a potential near term accident tolerant fuel candidate. Here, the results of decades of industry and academic research into Cr-doped UO 2 are analyzed and their shortcomings are critiqued. Focusing on the incorporation mechanisms of Cr into the fuel matrix, we explore a mechanistic understanding of the characteristic properties of Cr-doped UO 2 , notably, enhanced fission gas retention attributed to enlarged grain sizes following sintering, along with marginal improvements in the thermophysical properties. The findings of recent X-ray Adsorption Near Edge Spectroscopy studies were compared and put into conversation with historic data regarding the incorporation of Cr in UO 2 . On the basis of defect mechanisms, the case is made for the substitutional incorporation of Cr governing the lattice solubility but not the enhanced U diffusivity. Instead, Cr/CR 2 O 3 redox chemistry in a well-defined oxygen potential explains the differences in the U diffusivity and O/M ratio. The primary mechanism of doping enhanced grain growth is found to be liquid assisted sintering due to a CRO (1) eutectic phase at the grain boundaries. The role of inhomogeneities in Cr concentration in UO 2 at various length scales across the materials microstructure is highlighted and connected to promising experimental and modeling work to fill in the gaps in the current understanding of Cr-doped UO 2 . The review considers both the open scientific questions and engineering applications to illustrate the deep connections between the practice and theory in the design of accident tolerant nuclear fuels. In conclusion, the review ends with an outline of future works that combine meticulous irradiation studies and high resolution experiments with next generation modeling and simulations techniques empowered by machine learning advances to accelerate the fabrication and adoption of Cr-doped UO 2 light water reactors.

Cleveland, Mack Wesley [Massachusetts Inst. of Tec↗
DOE OSTI2020

Fracture Strength Determination Methods for Ceramic Materials Applied to Uranium Dioxide

The Advanced Fuels Campaign is currently focusing on development of accident-tolerant fuels that possess a range of property modifications intended to improve fuel performance during accident and transient conditions as well as extend license limits to burnups beyond 62 MWd/kgU. Both of these drivers have identified understanding and mitigating fuel cracking as a key performance criterion. Small-scale cantilever beam testing has been identified as a plausible method by which to collect fracture data for UO 2 as a function of chemical and structural evolutions introduced either during fabrication or irradiation. While this method has been found to be capable of providing data that are in reasonable agreement with literature for unirradiated UO 2 , the inherently small sample volumes that can be sampled limit its ability to capture the statistical nature of mechanisms that govern the fracture of brittle ceramics. A biaxial flexure strength test was developed to be used for unirradiated UO 2 . This method is standard in the community, but no systems presently in use at national laboratories, universities, or private companies are available to be used for nuclear fuel materials. This report describes the operation and benchmarking of this system, which has been validated for a number of common oxides. Future work will extend this system to characterization of both doped UO 2 and other relevant microstructural modifications that can also be measured using small-scale cantilever beam testing. Comparison of datasets collected using both methods will allow the overall applicability of small-scale techniques to be assessed and provide confidence or bounds on their use for irradiated fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

ENDF/B-VIII.1: Thermal Neutron Scattering Sublibrary

The thermal neutron scattering law (TSL) sublibrary aims to describe the interaction of incident neutrons at thermal or sub-thermal energies with different compound materials such fuels, moderators and special-purpose materials. In ENDF/B-VIII.1 there was a large number of new and updated TSL evaluations, including traditional moderators (light water, Beryllium metal, Beryllium Oxide, Calcium Hydride, plastics (Polystyrene and Lucite), graphite (reactor-grade and crystalline), anhydrous Hydrogen Fluoride, and heavy paraffinic oil); exotic moderators (Beryllium Carbide, Zirconium Hydride, Yttrium Hydride, Lithium-7 Hydride and Deuteride), FLiBe molten salt, structural materials and cladding (Silicon Carbide, Silicon Dioxide, Zirconium Carbide), fuels (Plutonium Dioxide, Uranium Carbide, Uranium metal, Uranium Nitride, Uranium Dioxide, Uranium Hydride), and special purpose materials. In ENDF/B-VIII.1 we also distribute alongside the evaluated files, a comma-separated file (CSV), named TSL_MAT_numbers.csv, which lists all evaluated files in the current release and their corresponding unique MAT number.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Benchmark Gap Assessment for the Manufacturing of High-Assay Low-Enriched Uranium Fuels

This document develops basic critical conditions for spheres—moderated and unmoderated, as well as reflected and unreflected—in consideration of nuclear criticality safety of a potential fuel production facility producing high-assay low-enriched uranium (HALEU) fuel of several different types like tristructural-isotropic (TRISO), uranium metal and alloys, oxide and non-metallic forms. In addition to spherical arrangements, TRISO particle manufacturing process–specific equipment is modeled as it would be for the criticality safety analysis. The objective is to develop representative systems that can then be used for comparison with existing benchmarks. SCALE/TSUNAMI is used to assess the similarity index between these systems to assess validation gaps for possible fuel production applications of proposed advanced reactors. Several different fuel types were evaluated, including TRISO, uranium metal, uranium molybdenum, uranium zirconium, uranium dioxide, uranium nitride, uranium hydride, U-ZrH, and uranium chloride. This selection of fuel types covers a breadth of proposed reactor types, as well as intermediate steps in the production and fabrication of the fuel

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Direct measurement of 5f delocalization with U XES

Delocalization of the 5f states in the early actinides in general and U metal in particular is significantly important and yet poorly understood. Here, extant spectroscopic techniques have failed to resolve the situation. Here it will be shown that X-Ray Emission Spectroscopy (XES) of the M 4,5 levels can provide the needed information, with a distinct difference between the delocalized U metal and localized uranium dioxide and uranium tetrafluoride cases. A Peak Ratio (PR) model, built upon electric dipole selection rules, has been developed and utilized, with quantitative agreement between experiment and theory. Possible expansion to other types of 5f mixing systems will be discussed.

36 MATERIALS SCIENCE↗
DOE OSTI2021

High Level Gap Analysis for Accident Tolerant and Advanced Fuels for Storage and Transportation

This initial gap analysis considers proposed accident tolerant fuel (ATF) options currently being irradiated in commercial reactors, since these are most likely for future batch implementation. Also, advanced fuel (AF) options that may be likely for use in advanced reactors are considered. The cladding technologies considered were chromium-coated zirconium-based alloys, FeCrAl, and both monolithic and matrix composite Silicide carbide (SiC). The fuel technologies considered were chromium-doped uranium dioxide fuel, uranium alloys, uranium nitride, and uranium silicide. Numerous national labs, industry, and countries are performing significant testing and modeling on these proposed technologies to establish performance, but at this time none of the prototypes being irradiated have achieved end-of-life (EOL) burnup. There are some testing results after one burnup cycle to verify in-reactor performance, but little data beyond that. As the ATF prototypes acquire more burnup, data will be produced that is relevant to storage and transportation. The DOE:NE Spent Fuel and Waste Science and Technology (SWFST) Storage and Transportation (ST) Control Account will evaluate the performance data as it becomes available for application to the identified gaps for ST.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Microfluidic Uranium Microspheres Production for TRISO and Advanced Fuel Concepts

The purpose of this project was to expand existing internal gelation sol-gel capabilities at PNNL to explore producing uranium dioxide spheres for potential use as fuel kernels in next generation Tri-isotropic (TRISO) particle fuel. This project expands on previous sol-gel efforts at PNNL by (1) increasing the size regime of sphere production from the micro-fluidic range to the milli-fluidic range, and (2) producing uranium spheres. The approach involved first scaling up the channel size of the fluidic system to the millimeter range, with radiation safety considerations in mind; testing and demonstration on non-radioactive surrogate material, cerium oxide; then transitioning to uranium production and finally optimizing system parameters. Commercially available fluidic chips in the desired size range were could not be found, therefore the project designed and fabricated a T-junction with 1mm channels for droplet production. Because the production process is temperature sensitive, prior efforts have involved performing droplet production in a lab freezer. To reduce radiological waste and footprint, two alternative chilling methods were explored using aluminum thermal beads as a chill bath and a custom aluminum block fit to reagent reservoir sizes. Both were successful in the cerium tests, however the aluminum block design outperformed the thermal bead bath and was further adapted for the radiological test and production run in the Radiochemical Processing Laboratory (RPL). Gelation trials were performed to determine an acceptable range of feed solution parameters for the uranium dioxide gels, characterized by R-values, which is determined by the ratio of uranium nitrate to Hexamethylenetetramine (HMTA)/urea in the feed solution. R-values ranging from 1.6-2 were examined in the gelation trials, with only the 1.8 condition being tested in production. The project was successful in demonstrating a proof of concept design for producing uranium dioxide spheres, however further optimization is needed to dial in production parameters and improve sphere quality and homogeneity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Atomistic and cluster dynamics modeling of fission gas (Xe) diffusivity in TRISO fuel kernels

TRISO fuel particles are candidates for use in next generation reactors including gas reactors, fluoride salt-cooled high temperature reactors, and micro-reactors. The UCO fuel kernel consists of a uranium dioxide (UO) and uranium carbide mixture. The addition of UC helps suppress the formation of carbon monoxide gas, which led to failures during initial TRISO development. The addition of uranium carbide alters the chemistry of the UO kernel, which is known to influence performance parameters such as fission gas diffusivity, although the impact has not been quantified and no models exist that take the change in chemistry into account. Therefore, better understanding and more accurate models of the impact of chemistry on fuel performance are of high priority. In this paper, a first-principles density functional theory (DFT) and empirical potential based multi-scale study has been carried out to model the diffusivity of fission gas xenon (Xe) in UCO TRISO fuel kernels. The focus is on the UO component in the UCO fuel kernels, as that represents the largest volume fraction of the fuel kernels. The study relies on DFT and empirical potential calculations to determine Xe and point defect properties, which are then used in thermodynamic and kinetic models to predict diffusion for intrinsic conditions. In addition, the information is utilized in cluster dynamics simulations using the Centipede code to estimate the impact of irradiation on defect transport. Additionally, the presence of UC or UC in the UCO fuel kernels is shown to have a substantial impact on the UO non-stoichiometry by inducing oxygen vacancies and driving UO sub-stoichiometric, which causes much slower Xe diffusion in UCO compared to light water reactor UO fuel. The application of this model in fuel performance simulations using the Bison code is also demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Thermal Neutron Scattering Improvements and Fixes for MCNP6.3 [Slides]

Major error (10,000+ pcm) discovered when using uranium-dioxide or uranium-nitride included in ENDF80SaB2 release. Improvements are being made towards preparing the code for ENDF/B-VIII.1 (ACE) format updates. A minor error (0-10’s pcm) was discovered while working on new capability development for Lab Directed Research & Development (LDRD) project. Previously, only a single coherent or incoherent elastic channel (along with an inelastic channel) was allowed in a thermal neutron scattering evaluation and subsequent processed ACE file. Based on what is planned in ENDF/B-VIII.1 release, there will be (a) new thermal scattering evaluation(s) which uses this mixed-mode coherent and incoherent elastic scattering blocks. MCNP6.3 has been modified to be able to handle such data in the future. Some reorganization has been made in the cross section calculation (acetot, sabcol, colidn) in the presence of TSL data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Complete Survey of Fuel Candidates for Microreactor Purposes

This report summarizes various fuel types that may be applicable to several microreactor concepts, which are defined as (1) very high temperature reactors (VHTR), (2) sodium fast reactors (SFR), (3) system for nuclear auxiliary power (SNAP) reactors, (4) gas fast reactors (GFR), and (5) molten salt reactors (MSR). The fuel systems that were assessed include: uranium mononitride (UN), uranium monocarbide (UC), uranium dioxide (UO 2 ), uranium oxycarbide (UCO) Tristructural Isotropic (TRISO), UN TRISO, mixed oxide (MOX), metallic fuels, and metal hydrides. While UCO TRISO has undergone significant testing through the advanced gas reactor (AGR) program, the very high cost necessitates consideration of other fuel types. UN and UC were identified as fuels that should receive further investigation due to their thermophysical and mechanical properties. Minimal irradiation performance data shows that these fuels are good candidate fuels for microreactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Generation of Enrichment-Dependent Thermal Neutron Scattering Data

This work details the generation of enrichment-dependent thermal neutron scattering cross sections for several crucial uranium fuel compounds. The evaluations of the thermal scattering law (TSL) and associated cross sections for uranium dioxide (UO 2 ), uranium carbide (UC), and uranium nitride (UN) were performed using standard ab initio lattice dynamics (AILD) methods. The data for uranium metal was produced using a novel hybrid approach of molecular dynamics combined with lattice dynamics methods. 235 U enrichments of 5%, 10% (LEU+), 19.75% (HALEU), 93% (HEU), and 100% were considered, in addition to natural uranium. The enrichment-dependent masses and free atom cross sections were used in the generation of elastic and inelastic thermal neutron scattering cross sections, while the calculation of the phonon density of states (DOS) and resulting TSL considered only the natural isotopic composition of uranium. The use of an identical DOS for all enrichments is expected to have minimal impact on the final data, as the small change in uranium mass should not significantly affect lattice vibrations. The cross sections are shown to exhibit significant dependence on 235 U enrichment. The submission of this data to the National Nuclear Data Center (NNDC) for release in the ENDF/B-VIII.1 database should support the design of advanced reactor concepts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗