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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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Exploring uranium oxidation in glue vapor to help avoid catastrophic slivers on NIF targets

Depleted uranium (DU) hohlraums are utilized for indirect drive fusion experiments at Lawrence Livermore National Laboratory. The production process for DU hohlraums is intricate, involving complex sputtering, machining, and leaching techniques. Achieving low-stress and defect-free hohlraums presents significant challenges. Recent failures have revealed that some uranium flakes, known as slivers, detach from the diagnostic windows and fall onto the capsule. These catastrophic failures were observed only after target assembly, leading to the hypothesis that oxidation of the DU by glue vapor might be the cause. A study investigating the impact of glue vapor on DU hohlraums found that spallation occurred in three out of seven sample hohlraums after three weeks. The slivers observed in these samples were consistent with those seen in assembled targets. This marks the first experimental reproduction of such catastrophic slivering failures. Consequently, future studies can now focus on sliver mitigation and the precise diagnosis of the failure mode.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2024

Uranium Mononitride (UN) Handbook

Uranium Mononitride (UN) is being reexamined as a high temperature advanced nuclear fuel in many reactor applications due to its high thermal conductivity, high melting point, and high fissionable uranium density, among other desirable properties. A thorough review of the experimental data for the material and irradiation properties has been collected and is presented in this handbook. As an outcome of this review, in many cases, it is apparent that additional experimental verification must be conducted to verify the available data and to verify the suggested empirical correlations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

U.S. Department of Energy Office of Legacy Management 2024 Annual Site Inspection and Monitoring Report for Uranium Mill Tailings Radiation Control Act Title I Disposal Sites

This report, in fulfillment of a U.S. Nuclear Regulatory Commission (NRC) license requirement, presents the results of long-term surveillance and maintenance activities conducted by the U.S. Department of Energy (DOE) Office of Legacy Management (LM) in 2024. These activities occurred at the 19 uranium mill tailings disposal sites established under Title I of the Uranium Mill Tailings Radiation Control Act (UMTRCA) and verify that the UMTRCA Title I disposal sites remain in compliance with license requirements and Long-Term Surveillance Plans (LTSPs). Individual chapters for each site are available on the LM public website at https://energy.gov/lm/sites/lm-sites.

54 ENVIRONMENTAL SCIENCES↗
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

Internal-Gelation Production of Uranium Oxide Sol-Gel Particles for Forensic Applications

The purpose of this project was to develop and demonstrate a novel method for the production of uranium oxide microsphere particles with tunable chemical compositions via a sol-gel process using a 3D-printer setup. These particles can serve several purposes in research and development as a forensic training tool or as standard reference materials. A key component of the project was to demonstrate the ability to control physical and chemical parameters of the particles created. First, we demonstrated the ability to employ an internal gelation sol-gel process to create individual uranium oxide particles. The particles were successfully dispensed using a unique 3D-printing setup onto a substrate to react and then were collected and thermally processed. A series of temperatures for the annealing process was tested on individual samples to investigate the effect on the sol-gel chemical composition and physical integrity. Next, we demonstrated the ability to control matrix composition of the particles by separately incorporating fission product isotopes as well as Np-237 into the sol-gel solution. It was shown by gamma-ray spectroscopy that these matrix elements were successfully retained during the gelation process. We studied the retention of the elements across a series of annealing temperatures. Additionally, we demonstrated the ability to quantitatively control the isotopic composition of the particles by altering the U-237/U-238 ratio to a controlled value. Finally, X-ray diffraction analysis (XRD) was used to investigate the oxidation state of the sol-gel after annealing at different temperatures.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2024

Updated uranium reference material 231 Pa/ 235 U consensus model ages for nuclear forensics

Radiochronometry provides the model age of nuclear materials, which is useful for understanding the production history of materials found outside of regulatory control. Certified reference materials (CRMs) are important for radiochronometry to increase confidence in measurement quality; however, the absence of 231 Pa/ 235 U CRMs necessitates that nuclear forensic laboratories measure the 231 Pa/ 235 U model ages of preexisting uranium reference materials. Here, in this work, new consensus 231 Pa/ 235 U model ages are reported from three nuclear forensic laboratories for three reference materials using current analytical methods. These updated consensus values can be used for quality control of 231 Pa/ 235 U model age measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Neutron capture signatures determined from gadolinium isotope compositions of uranium ore concentrates using MC-ICPMS

We present Gd isotope compositions on 25 well-characterized uranium ore concentrates (UOCs). About half the UOCs have isotope depletions in 157 Gd coupled with anticorrelated excesses in 158 Gd due to neutron capture effects. UOCs from older ore bodies and with higher U contents show larger neutron capture effects than UOCs from younger ore bodies with lower U contents. These Gd data are correlated with their previously measured Sm isotope compositions, and we use the Sm data to estimate the neutron fluence of these samples. In conclusion, this work demonstrates how Gd isotope signatures can be employed as a new tool for nuclear forensics.

Earth Sciences↗
DOE OSTI2025

Quantifying uncertainty in uranium concentration measurements via K-edge densitometry

This study quantifies the uncertainty in uranium concentration predictions of fluoride and chloride-based salts within a steel pipe using K-edge densitometry. Modeling and simulation was conducted with the Monte Carlo N-Particle Transport (MCNP) code. The quality of of this technique’s prediction in a pipe requires proper characterization of the pipe’s thickness, which is dependent on the source size and axial offset from the pipe centerline. The thickness was determined as either the center-line thickness seen by the X-ray source or an average value determined through random sampling. Generally, the predicted concentrations were slightly better at lower offset with the random sampling thickness and using the center-line thickness for the highest offsets. For a line-beam source and varying axial offsets, the relative error of concentration was within 1% of the true value but uncertainty increased by 2 orders of magnitude. Similarly, for no axial offset, the relative error was significantly less than 1% while no trend for uncertainty was found. However, at the largest possible offset for a given source size, the concentrations become erroneous and greater than the allowable 1% relative error. Furthermore, high offsets tended to increase the variance of the transmission spectra by 3 orders of magnitude.

Characterization and Analytical Technique↗
DOE OSTI2026

Characterizing microscale signatures in uranium ore concentrates using electron probe microanalyzer

Impurities in uranium ore concentrates (UOCs) serve as forensic signatures of processing history and origin. Here, this study utilizes Electron Probe Microanalyzer (EPMA) to characterize microscale compositional and textural features in individual UOC particles from both commercial and bench-scale production. At the particle scale, multiple internal phases with distinct morphologies, chemical signatures, and stoichiometries are documented. Our data shows that chemical impurities are heterogeneously distributed within single particles and among particles within a sample. These microscale heterogeneities correlate with known processing histories, indicating that microscale signatures of early fuel cycle materials can provide valuable information for nuclear forensic material analysis.

organic↗
DOE OSTI2026

Ra-226 extraction and purification from simulated uranium tailings

Growing demand for Ac-225 has outstripped current production, increasing interest in alternative production routes. Furthermore, these often rely on irradiations of Ra-226, yet Ra-226 supplies are also limited. Uranium tailings provide an abundant source of Ra-226, though separation from these complex matrices remains challenging. A laboratory-scale process was developed to recover Ra-226 from simulated U-tailings. After acid leaching, Ra remained adsorbed on refractory oxides and silicates. An additional CaCl₂ leach of the residue raised recovery to ~ 64%. Subsequent purifications, including precipitation, ion-exchange, and extraction chromatography (TK101), achieved chemical yields up to 73 ± 7%, producing 7.68 ± 0.1 µCi of high-purity Ra-226.

and nuclear chemistry↗
DOE OSTI2024

Measuring thermal diffusivity and gap conductance in uranium nitride and Zircaloy relevant for microreactor applications

Heat transfer across nuclear fuels and structural interfaces is an important factor for evaluating the performance of nuclear power systems. Specifically, heat generated as nuclear fuel fissions must be transported through the cladding material and through the reactor to reach the steam turbine for power generation. As new microreactor designs emerge, maximizing the efficiency of this heat transfer process becomes crucial to make them commercially viable. This article examines thermal diffusivity and gap conductance in uranium nitride (UN) fuel and Zircaloy-4 (Zry4) cladding using light flash analysis (LFA). Thermal diffusivity measurements were made on monolithic UN pellets and Zry4 exposed to carbon at peak operating temperatures of microreactors and show that carbon ingress has a minimal effect on thermal diffusivity when compared with identical materials not exposed to carbon. Evaluation of gap conductance at the UN-Zry4 interface was done using one-dimensional two-layer thermal transport models as a function of applied pressure. Here the results show that increasing pressure on the UN-Zry4 interface leads to gains in gap conductance per unit area in fuel-cladding assemblies at microreactor operating temperatures. While many other variables are expected to influence UN-Zry4 interfacial gap conductance (e.g. contact surface roughness, porosity, localized heating, environmental gas pressure), the work offers a demonstration of using a conventional LFA apparatus to determine this parameter at elevated temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microstructurally validated stable and predictable swelling in low-enriched uranium monolithic U-10Mo fuel mini-plates

Qualification of the low-enriched uranium (LEU) monolithic U-10 wt%Mo (U-10Mo) plate-type fuel system requires a demonstration of a stable and predictable fuel swelling behavior over the anticipated operating conditions of the United States high-performance research reactors (USHPRRs) selected for conversion to LEU operation. This will allow each reactor to develop appropriate safety margins that will retain fuel element lifetime coolability. Additionally, the fuel system must maintain performance attributes when fabricated at a commercial scale. The Mini-plate 1 experiment represents the first irradiation test of commercially fabricated miniaturized monolithic LEU U-10Mo fuel plates. Here, the swelling behavior within this experiment was compared against that of historical fuel developmental tests to reveal that the commercially fabricated fuel performed within the current recommended U-10Mo swelling model's predictions. Additionally, the fuel microstructural evolution was evaluated to link initial conditions to subtle variations detected in the swelling response, providing validation and confidence that the fuel system is robust.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Chemical interaction and compatibility of uranium mononitride and alumina forming austenitic stainless steel

Uranium mononitride (UN) and alumina forming austenitic (AFA) stainless steel are a potential fuel-cladding combination for the lead-cooled fast reactor (LFR). Chemical compatibility between UN and AFA steel needs to be verified before implementation in a nuclear reactor. Diffusion couple experiments at 823 K and 1023 K were conducted for nonirradiated UN samples in contact with as-cast (no thermally grown Al 2 O 3 ) and preoxidized (with thermally grown Al 2 O 3 ) AFA for 500 and 1000 h in an inert environment. Preoxidized AFA exhibited little to no interaction with all UN samples tested at both 823 K and 1023 K, displaying the stability and capability of the Al 2 O 3 layer to prevent chemical interaction and inter-diffusion with UN. Chemical interaction occurs between UN and as-cast AFA. At 1023 K, an aluminum and nitrogen rich phase (likely AlN) formed along the interface of as-cast AFA and UN samples. At 823 K the AlN phase was not prominently observed due to the reduced diffusivity of aluminum through AFA. The aluminum and nitrogen-enriched phase was also observed in a high temperature pressure-assisted test sample of UN and as-cast AFA thermally treated at 1373 K. Finally, in UN samples doped with a low weight percent of UO 2 (< 3 wt%), AlN was not detected along the interface at either temperatures, and an Al 2 O 3 layer likely formed along the interface and prevented further chemical interaction between UN and as-cast AFA.

36 MATERIALS SCIENCE↗
DOE OSTI2025

A systematic exploration of the hydrolysis products of the uranium trioxide polymorphs and their optical vibrational spectra

Study of the uranium trioxide (UO 3 )-water system is complex with inconclusive results and limited details in the literature. The UO 3 system is home to at least seven structural polymorphs and an amorphous phase. The proposed hydrolysis products of UO 3 are just as numerous, yet investigations of these alteration products are sporadic and generally antiquated, thus requiring systematic investigations. Recent developments in the understanding of UO 3 phase space, aided by improvements in analytical and computational techniques, necessitate more modern investigations into the uranyl hydroxide family and their naturally occurring mineral counterparts. We present findings from a systematic investigation of the products formed via hydrothermal reactions of common UO 3 polymorphs and discuss how the equatorial coordination of the uranyl/uranyl-like ions within the UO 3 precursors leads to differences in the optical vibrational spectra of the resulting hydrolysis products. The hypo-stochiometric nature of α-UO 2 (OH) 2 allows for the formation of multiple unique uranyl sites and a distortion of the unit cell to a lower symmetry. This study provides, for the first time, an analysis of β-UO 2 (OH) 2 using modern techniques and instrumentation (Raman/infrared spectroscopy and powder x-ray diffraction) and lays a foundation for future time-dependent investigations into the structural dependence of the hydrolysis kinetics of the UO 3 phases. In conclusion, given the prevalence of UO 3 at both ends of the nuclear fuel cycle, an understanding of its behavior with water has applications ranging from nuclear forensics to waste management and environmental transport.

Uranium trioxide↗
DOE OSTI2025

Molecular-dynamics study of diffusional creep in uranium mononitride

Uranium mononitride (UN) is a promising advanced nuclear fuel due to its high thermal conductivity and high fissile density. Yet, many aspects of its mechanical behavior and microstructural features are currently unknown. In this paper, molecular dynamics (MD) simulations are used to study UN's diffusional creep. Nanometer-sized polycrystals are used to simulate diffusional creep and to calculate an effective GB width. It is found that Nabarro-Herring creep is not dominant in the temperature range of 1700–2000 K and that the dominant diffusional creep mechanism is Coble creep with an activation energy of 2.28 ± 0.09 eV. A method is proposed to calculate the diffusional GB width and its temperature dependence in polycrystals. The effective GB width of UN is calculated as 2.69 ± 0.08 nm. This value fits very well with the prefactor of the phenomenological Coble creep formula. It is demonstrated that the most comprehensive thermal creep model for UN can be represented as the combination of our Coble creep model and the dislocation creep model proposed by Hayes et al.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Machine learning framework for predicting uranium enrichments from M400 CZT gamma spectra

A machine learning framework was developed for predicting uranium enrichments from M400 CZT gamma spectra. This framework leverages the availability of a large amount of measured M400 gamma spectra and uses a recently updated version of Gamma Detector Response and Analysis Software (GADRAS) for gamma spectrum analysis and generation. It also leverages the existing machine learning modules in Python for gamma spectrum data processing, curation, model training, benchmarking, and optimization of the deep machine learning models. The framework is used to develop a deep learning model to analyze gamma spectra from a set of U 3 O 8 samples with enrichments ranging from 0.31 to 93.17% and UF 6 cylinders with enrichments ranging from 0.2 to 4.95%, and the model performance is tested using a set of measured spectra and the respective declared enrichment values. Results show that the model can correctly classify 99.35% of the U 3 O 8 sample enrichments, and can predict the samples’ enrichments within an average absolute error of 0.099% (in percentage points of enrichment). For the UF 6 cylinders, the average absolute error was approximately 0.03%, with an accuracy of 98% in classifying discrete enrichment values of UF 6 samples. Finally, the results also show that the model has performed significantly better in terms of predicting enrichments in UF 6 cylinders based on measured gamma spectra than the GEM code, with a standard deviation (of the relative errors) of 2.23% (compared with the 11.51% value for the GEM code) based on results from a set of test data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Electronic and Geometric Contributors to Hydrogen Binding in Uranium Oxide Grain Boundaries

Hydrogen induced corrosion of uranium, which leads to the formation of toxic and pyrophoric UH 3 , raises significant safety concerns for long-term storage of nuclear materials. Previous work suggests hydrogen diffuses through the grain boundaries (GBs) of the passivating oxide layer to initiate hydriding reactions. However, the atomistic mechanisms underlying this phenomenon and the structural factors that control its initiation are not well understood. To address this knowledge gap, here we use a high-throughput density functional theory (DFT) workflow to investigate the adsorption of H and H 2 in the defective bulk UO 2 . Specifically, we have exhaustively investigated the adsorption of H (107 sites) and H 2 (26 sites) in three different coincident site lattice (CSL) GBs: Σ3, Σ5, and Σ9. Compared to the binding energies in pristine UO 2 , we observe significantly stronger hydrogen adsorption at these GB sites. Interestingly, we find that the trends in H and H 2 adsorption vary considerably across the three GB models. In particular, while a small number of sites in Σ5 and Σ9 show exothermic adsorption of H and H 2 , respectively, no such sites are found in Σ3. These results provide fundamental atomistic insights that could guide the development of future corrosion mitigation strategies for the storage of nuclear materials.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗