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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 289 records · Page 16
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Analysis of the 30B UF 6 Container for Use with Increased Enrichment

Commercial light-water reactor (LWR) operators and fuel vendors are pursuing advancements in fuel and reactor design that include increasing 235 U enrichment from low enrichment (<5 wt.% 235 U) to low-enriched uranium+ (LEU+) (10 wt.% > 235 U > 5 wt.%) and high-assay low-enriched uranium (HALEU) (20 wt.% > 235 U > 10 wt.%). These advancements will necessitate the ability to transport LEU+ and HALEU fuel materials. Assessment of Existing Transportation Packages for Use with HALEU(ORNL/TM-2020/1725) assesses the potential to use currently licensed transportation packages for the transportation of HALEU by evaluating a representative package for each uranium fuel form category (for example, oxide and/or metal in fuel assemblies, fuel pins, powder, pellets, and uranium hexafluoride [UF 6 ].) This report expands on the UF 6 analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Accelerated and heterogeneous corrosion of Cr-doped uranium nitride fuel pellets

The steam oxidation of Cr-doped UN fuel pellets is analyzed during sequential isothermal holds up to 720 °C. In situ neutron diffraction results show how Cr is accommodated in a secondary U 2 CrN 3 phase, leading to the formation of a duplex UN/U 2 CrN 3 microstructure. Under corrosion, the oxidation of the two phases begins at 400 °C for UN and 430 °C for U 2 CrN 3 , respectively. Because the UN phase is preferentially oxidized in the presence of U 2 CrN 3 , addition of Cr in UN based nuclear fuel is found to accelerate the corrosion rate. At 430 °C the oxidation of UN in the UN/U 2 CrN 3 microstructure is ~ 5 times faster than pure UN, increasing to ~19 times faster at 460 °C. The oxidation of U 2 CrN 3 produces UO 2 via the formation of two transient intermediate phases. In situ neutron diffraction enables oxidation processes of UN and U 2 CrN 3 components to be followed separately within the two-phase system.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Metallic Coating of Cerium Oxide Microspheres

The ability to remove heat is paramount to nuclear fuel performance and longevity. Retaining fission product and separating fuel from reactor coolant and the environment is also necessary to prevent radiological contamination. Conventional nuclear fuel for commercial light water reactors and radioisotope power systems (RPS) is composed of oxide powders pressed into a pellet (cm-scale) and then sealed into a metal cladding to confine the fuel. What typical fuels lack is a method to surround each particle of nuclear fuel in metal, thus providing a more intimate protection layer for accident tolerance and boosting the thermal extraction from the fuel element. In such a way, metal-coated fuel particles increase heat extraction efficiency over clad-pellet designs while increasing the accident tolerance of the fuel. Metal oxide microspheres have wide-ranging applications, including the realm of fuels for nuclear reactors and RPS. Microspheres of uranium oxide/uranium carbide, mixed uranium/plutonium oxides, transuranics, and thorium fuels have been extensively studied. Pacific Northwest National Laboratory has also demonstrated the production of 238 PuO 2 microspheres for RPS applications. Metal-coated oxide microsphere fuels may also be attractive for other applications such as nuclear thermal rockets, future nuclear reactor designs, and catalysts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Synthesis and thermodynamics of uranium-incorporated α-Fe 2 O 3 nanoparticles

Hematite nanoparticles were synthesized with U(VI) in circumneutral water through a coprecipitation and hydrothermal treatment process. XRD, TEM, and EXAFS analyses reveal that uranium may aggregate along grain boundaries and occupy Fe sites within hematite. The described synthesis method produces crystalline, single-phase iron oxide nanoparticles absent of surface-bound uranyl complexes. EXAFS data were comparable to spectra from existing studies whose syntheses were more representative of naturally occurring, extended aging processes. Herein this work provides and validates an accelerated method of synthesizing uranium-immobilized iron oxide nanoparticles for further mechanistic studies. High temperature oxide melt solution calorimetry measurements were performed to calculate the thermodynamic stability of uranium-incorporated iron oxide nanoparticles. Increasing uranium content within hematite resulted in more positive formation enthalpies. Standard formation enthalpies of U x Fe 2–2x O 3 were as high as 76.88 ± 2.83 kJ/mol relative to their binary oxides, or -764.04 ± 3.74 kJ/mol relative to their constituent elements, at x = 0.037. Data on the thermodynamic stability of uranium retention pathways may assist in predicting waste uranyl remobilization, as well as in developing more effective methods to retain uranium captured from aqueous environments.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Alkali Counterion-Dependent Crystallization of Uranium(IV)–Chloro Structural Units

The synthesis, structural characterization, and spectroscopic properties of five tetravalent uranium (U) phases including Li 6 [U 4 (μ 3 -O) 2 Cl 18 (H 2 O) 2 ]·10H 2 O (1), [U(H 2 O) 4 Cl 4 ] (2), [U(H 2 O) 4 Cl 4 ]·KCl (3), Rb 2 UCl 6 (4), and Cs 2 UCl 6 (5) are reported. Notably, a change in the U 4+ solid-state structural unit was observed based on the identity of the alkali counterion used in the synthesis. Li 1+ yielded a tetranuclear oxo-bridged cluster, [U 4 (μ 3 -O) 2 Cl 18 (H 2 O) 2 ] 6– , Na 1+ and K 1+ yielded two structurally distinct [U(H 2 O) 4 Cl 4 ] complexes, and Rb 1+ and Cs 1+ resulted in [UCl 6 ] 2– as the dominant phases. The spectroscopic properties of the compounds were analyzed using Raman and UV–vis–NIR absorption spectroscopy. The UV–vis–NIR spectra of compounds 1–5 exhibited transitions consistent with uranium in the +4 oxidation state. Clear differences in the absorption band splitting were observed and are likely attributed to differences in metal ion coordination, crystal field effects, and outer sphere interactions Overall, this work demonstrates the utility of noncovalent interactions in tuning the crystallization of various metal complexes from otherwise identical reaction solutions and provides further evidence that counterions impact the composition and structure of actinide complexes isolated in the solid state. In this way, this work affords important insight into directing and controlling the structure of actinide complexes and clusters.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

International database of reference gamma spectra for nuclear safeguards applications

Nuclear safeguards missions use gamma spectroscopy as a non-destructive measurement technique for examining nuclear materials. Despite advances in the development of detection equipment as well as software codes, one of the concerns is the lack of well-documented spectra needed to test and validate isotopic analysis codes for their applicability. To address this need, this work introduces IDB, an international database of the reference gamma spectra of uranium, plutonium and mixed oxide nuclear material samples. IDB provides access to well-characterized sets of gamma spectra described by rich metadata, including information on the sample, measurement configuration and detector specifications. These spectra are accessible in different formats, also compatible with analysis code standards, thus promoting their sustainability and maintenance.

None, Dipti [International Atomic Energy Agency (I↗
DOE OSTI2021

Integral Experiment Request 523 Feasibility Study (Summary Report)

This report documents the feasibility phase of the Critical Experiment Design (CED) conducted as part of integral experiment request (IER) 523. The purpose of IER-523 is to explore the effects of using 35 weight percent enriched uranium dioxide-beryllium oxide (UO 2 -BeO) material on critical configurations using the Seven Percent Critical Experiment (7uPCX) at Sandia National Laboratories (Sandia). Preliminary experiment design concepts, neutronic analysis results, and proposed paths for continuing the CED process are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Centrus/Y-12 CRADA Summary [Abstract]

The goal of this Cooperative Research and Development Agreement (CRADA) is to deliver a capability for producing High-Assay Low-Enriched Uranium (HALEU) metal and oxide to support the various national interests and projects of Department of Energy/National Nuclear Security Administration (DOE/NNSA), Y-12 National Security Complex, and Centrus Energy. HALEU is required to support DOE nonproliferation objectives, research reactors, advanced commercial reactors, and other projects of importance to U.S. national and energy security.

36 MATERIALS SCIENCE↗
NASA NTRS1982

UV radiation from the young sun and oxygen and ozone levels in the prebiological palaeoatmosphere

UV measurements of young T-Tauri stars, resembling the sun at an age of a few million years, have recently been made with the International Ultraviolet Explorer. They indicate that young stars emit up to 10,000 times more UV than the present sun. The implications for the origin and evolution of O2 and O3 in the prebiological palaeoatmosphere are presented here. The results of photochemical calculations indicate that the O2 surface mixing ratio was a factor 10,000-1,000,000 times greater than the standard value of 10 to the -15. This new value reconciles the simultaneous existence of oxidized iron and reduced uranium.

Canuto, V. M.↗
DOE OSTI2020

The role of ambient gas confinement, plasma chemistry, and focusing conditions on emission features of femtosecond laser-produced plasmas

Ultrafast laser ablation coupled with optical emission spectroscopy is currently under development for standoff detection of elements and their isotopes for a wide range of application areas. In this work, we explore the role of ambient air pressure during the interaction of an ultrafast laser beam with two metallic targets (brass and uranium) with significantly different oxygen gas-phase reactivity. Plasma plumes were generated by focusing ~800 nm, ~35 fs pulses from an ultrafast laser system using a long focal length lens with f = 1 m. Two-dimensional spectral imaging was performed over the pressure range of 30 mTorr-700 Torr air to evaluate emission dynamics, plasma chemistry, signal-to-background ratio, and characteristic parameters (i.e. excitation temperature, electron density). An increase in ambient air pressure during ultrafast laser ablation leads to plume confinement and subsequent changes in the emission dynamics due to plasma chemistry as well as changes in plasma generation conditions (i.e. focused fs laser ablation versus filament ablation). Ablation crater morphologies were also investigated via scanning electron microscopy. Results indicate atomic emission intensity and signal-to-background ratios peak at moderate pressure levels (~50-100 Torr air) for both targets studied, although plasma chemistry influences uranium emission signatures. The emission features of the uranium plasma at pressures = 10 Torr showed the presence of atoms and oxide molecules which are co-located within the plume. We also find filament ablation leads to wider, more shallow craters compared to focused laser ablation. Our study provides unique insight into the interplay between plume dynamics, confinement, and oxygen chemistry of fs laser-produced plasmas and how these phenomena evolve with changing ambient air pressure.

Laser-produced plasma↗
DOE OSTI2025

UN Synthesis, Fabrication, and Characterization and UC Oxidation Study in support of Advanced LEU Fuel Concepts

Uranium nitride and carbide-based fuels are proposed for use in advanced LEU fuel systems. This report details recent development work to study synthesis, sintering and fabrication methods for single phase UN in monolithic geometries. These samples were intended to support a variety of irradiation testing conditions by tailoring enrichment, geometry, and compositional requirements. Two methods were under evaluation for synthesizing uranium nitride powder – carbothermic reduction-nitridation (CTR-N) of UO 2 powder feedstock and hydride-dehydride/nitridation (HDN) of uranium metal feedstock. Pellets were fabricated from powder feedstocks via conventional pressing and sintering methods. As-fabricated feedstocks and pellets were evaluated via X-ray diffraction (XRD) for phase purity and sintered materials were characterized for density. Both the CTR N and the HDN methods were successfully demonstrated as an effective process for fabricating phase pure UN. Additionally, results from machining efforts for UN are detailed. This report also contains information on a study trying to elucidate the oxidation mechanisms of sintered uranium carbide samples fabricated from feedstock synthesized via an arc-melting method.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Structure characterization and steam oxidation performance of U 3 Si 2 with Zr alloying additions

Research and development of high uranium density fuels supports the on-going effort to enhance reactor technology, safety, performance, and economics. Among high uranium density fuels, U 3 Si 2 is a favored fuel candidate due to its higher uranium density and improved thermal conductivity when compared to traditional UO 2 . While U 3 Si 2 demonstrates an improved fuel economy, it suffers from rapid degradation when exposed to water at T > 300°C. In order for U 3 Si 2 to be implemented in water cooled reactors, the water reaction must be understood and mitigated. Reported here is the performance of Zr alloying additions to U 3 Si 2 exposed to high temperature steam oxidation environments. Here, the investigation shows that homogenization heat treatment contributes to a delayed onset of oxidation and an improved homogeneity throughout the arc-melt fabricated alloyed compositions. Of the pre- and post-annealed samples, the post-annealed 7 vol% Zr addition exhibited the highest onset temperature of oxidation. A preliminary computational investigation to complement the experimental analysis assesses the solubility of Zr in U 3 Si 2 and its incorporation in the crystal lattice.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Uranium-hydrogen reaction mechanism and numerical model

We have developed a numerical model to help predict lifetimes of uranium parts in situations where the uranium surface is exposed to hydrogen in the gas headspace. Assessments can be made based on hydrogen pressure and on an upper limit of size of a surface breached hydride spot (volume of hydride corrosion product produced). The model assumes development of a hydride nucleus at a single arbitrary subsurface location associated with an arbitrary surface defect and follows the development of the hydride nucleus to the break-through phase and further growth to the upper limit of acceptable volume of corrosion product. The model has been developed from an understanding of the hydriding mechanism outlined below in 10 steps and measured rates of hydrogen ingress at several types (chemically) of uranium surfaces at ambient temperature and variable pressure. The key points are 1) that the hydrides nucleate only at locations where a surface defect allows hydrogen ingress (reactant delivery) into the uranium metal subsurface (beyond the oxide), and 2) that the hydrides which eventually reach surface break-through status, nucleate and grow in the near surface (few 10s of micron depth maximum) – those beyond ~50 micron have arrested growth and never reach break-through status.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Restoration Insights Gained from a Field Deployment of Dithionite and Acetate at a Uranium In Situ Recovery Mine

Mining uranium by in situ recovery (ISR) typically involves injecting an oxidant and a complexing agent to mobilize and extract uranium in a saturated ore zone. This strategy involves less infrastructure and invasive techniques than traditional mining, but ISR often results in persistently elevated concentrations of U and other contaminants of concern in groundwater after mining. These concentrations may remain elevated for an extended period without remediation. Here, we describe a field experiment at an ISR facility in which both a chemical reductant (sodium dithionite) and a biostimulant (sodium acetate) were sequentially introduced into a previously mined ore zone in an attempt to establish reducing geochemical conditions that, in principle, should decrease and stabilize aqueous U concentrations. While several lines of evidence indicated that reducing conditions were established, U concentrations did not decrease, and in fact increased after the amendment deployments. We discuss likely reasons for this behavior, and we also discuss how the results provide insights into improvements that could be made to the restoration process to benefit from the seemingly detrimental behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Delayed Critical and Subcritical Experiments with Polyethylene Moderated Unreflected Thin 15 in. Diameter HEU Metal Plates

The thin ~15 in. diameter highly enriched uranium (HEU) metal plates were assembled to delayed criticality at the Oak Ridge Critical Experiments Facility (ORCEF) in 1969 with various thicknesses of polyethylene (varying from 1/16 to 2$\frac{3}{8}$ inches) between uranium metal plates. The average 235 U enrichment was 93.27 wt. %. These unreflected critical configurations contained 4$\frac{2}{3}$ to 20$\frac{5}{6}$ thin 15 in. diameter HEU metal plates (on loan from Los Alamos National Laboratory [LANL] and shipped to Oak Ridge National Laboratory [ORNL] on June 3, 1969). Depending on the thickness of polyethylene, the enriched uranium masses varying from 28,053 to 135,148 grams. Fractional plate sections consisted of the appropriate number of 60° pie sections. In addition to the measurement at delayed criticality, subcritical measurements were also performed by the inverse kinetic rod drop method. Prompt neutron decay constant measurements were also performed by the Rossi alpha and randomly pulsed neutron method using a time-tagged spontaneous fission californium neutron source; these are briefly reported here. At the time of these measurements in 1969, the thin HEU metal plates were in near-pristine condition with extremely little oxidation, allowing better descriptions of the uranium plates than the use of these plates in a heavily oxidized and deteriorated condition in recent reflected benchmark experiments at the LANL facility at the Nevada Test Site with these same thin highly enriched uranium metal plates. This report documents the experimental information for the measurements performed so that later researchers can perform the required uncertainty and calculational analyses and documentation to use these data for an International Nuclear Criticality Safety Benchmark Evaluation Program (ICSBEP) or a Nuclear Energy Agency (NEA) benchmark. 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 on the measured neutron multiplication factors is completed. Additional data—such as the dimensional inspection reports, uranium isotopic information, and other relevant particulars—should be retrieved from the Y-12 Plant or LANL and incorporated in the final ICSBEP benchmark. 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. Other experiments with smaller-diameter than 15 in. diameter HEU metal plates have been benchmarked in HEU-METFAST-001. The prompt neutron time decay measurements could be the basis for an International Reactor Physics Benchmark Program. Preparation of the present report is part of an effort at ORNL to document more than 15 undocumented critical and subcritical experiments enumerated in ORNL/TM-2019/18 and performed by ORNL at ORCEF and other US Department of Energy critical experiments facilities using more than 500 operational days of critical facility time. This work for this report publication was supported by the Nuclear Criticality, Radiation Transport and Safety NCSP Program at ORNL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

In situ monitoring of lanthanide reactions with oxide species via combined absorption spectroscopy and electrochemical methods

Molten salts for engineering scale applications of spent nuclear fuel pyrochemical processing will inevitably have some level of oxygen impurities which can form various insoluble oxide and oxychloride species with uranium and fission products. This work demonstrates real-time concentration monitoring of two trivalent lanthanide (Ln3+) fission products, Nd3+ and Pr3+, and their reactions with oxygen (O2-) impurities to form insoluble products in LiCl-NaCl-KCl eutectic salt. A combination of high-temperature absorption spectroscopy and electrochemical testing were used to track lanthanide concentrations. O2- impurity levels were controlled in the range of 0.001 M to 0.5 M by adding Li2O to lanthanide-salt solutions. After the introduction of O2- impurities, Ln3+ concentrations were monitored via time-resolved absorption spectroscopy. Concentrations of both Ln3+ species in solution decreased with time as insoluble products formed. The initial impurity concentration controlled whether insoluble products were predominantly oxychlorides (LnOCl) or mixtures of oxychlorides and oxides (Ln2O3). However, absorption spectroscopy is limited for weakly absorbing species, such as Pr3+, and under conditions of high impurity concentrations where solutions can be turbid. To circumvent this limitation, the concentrations of Pr3+ were monitored with square wave voltammetry (SWV). Estimated reaction rates and extent of Pr3+ removal from solution as monitored by SWV agreed to within ~10% of the results found from absorption spectroscopy monitoring. This demonstrates that simultaneous electrochemical testing complements and expands the capabilities of absorption spectroscopy to monitor reactions of fission products in molten salts.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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↗