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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Early-stage uranium-hydrogen corrosion kinetics and mechanism

Research into the early-stage uranium-hydrogen degradation mechanism lacks unified interpretation and understanding due to inadequate temporal and spatial characterization resolution required to unify reaction observations. In this study, white-light interferometry was adapted to characterize the uranium-hydrogen reaction at a temperature of 50 °C and hydrogen pressure of 13.8 kPa. The uranium surface remained unchanged for 62 min during the hydride induction period before the first hydride blister appeared. This blister was monitored from induction through critical spallation and UH₃ powder release during rapid surface hydriding. The critical spallation dimensions measured for height, diameter, area, and perimeter of the first blister were 1.79 µm, 17.58 µm, 187.7 µm 2 and 59.79 µm. Hydriding growth kinetics were assessed by tracking the first hydride site over the reaction time. Analysis shows that the spall-front velocity of the hydride was changing at a rate of 0.91 µm/min which compares favorably with legacy Sievert’s experiments in literature. Total percentage of the surface area hydrided after 242 min was 42.7%, and post-characterization of the UH 3 powders shows both α-UH 3 and β-UH 3 . This work highlights the strength of characterizing the early-stage uranium-hydrogen reaction using white-light interferometry, but more importantly, unifying the understanding and mechanism of the uranium-hydrogen reaction kinetics.

Materials science↗
DOE OSTI2024

Uniaxial compressive creep tests by spark plasma sintering of 70% theoretical density α -uranium and U-10Zr

Metallic fuels hold numerous advantages over conventional uranium dioxide fuels and are a key component of several liquid metal-cooled advanced reactor concepts including sodium fast reactors. These fuels undergo rapid swelling during early burnup; consequently, they spend most of their reactor lifetime in a porous state. The presence of this porosity alters many of the mechanical properties of the fuel including creep impacting fuel deformation during axial swelling. This work investigates the creep behavior of the porous fuel using a spark plasma sintering technique. Creep tests were performed for the first time on porous α-phase uranium and uranium with 10 wt. % zirconium (U-10Zr) samples. The samples of α-phase uranium and U-10Zr were fabricated from depleted uranium by spark plasma sintering and subjected to uniaxial compressive creep testing. Calculated stress exponents were found to be 2.6±1.6 and 5.7±1.4 for α-U and U-10Zr, respectively, and calculated activation energies were found to be 61.6±1.1kJ/mol for α-U. The creep data were also used to evaluate existing porosity inclusive in creep models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Acceleration of uranium beam to record power of 10.4 kW and observation of new isotopes at Facility for Rare Isotope Beams

The Facility for Rare Isotope Beams (FRIB) is a major nuclear physics facility for research with fast, stopped, and reaccelerated beams that was successfully commissioned in May 2022. A key capability of FRIB is the production of an acceleration of the uranium beam, but this capability requires the facility to work at the design limits of the lowest charge-to-mass ratio and the highest power density on the beam intercepting devices. This paper presents techniques for overcoming the significant challenges in accelerating the uranium beam, culminating in the demonstration of 10.4 kW on target, and the discovery of three new isotopes. The high-power uranium beam enabled us to produce and identify G 88 a , A 93 s , and S 96 e , within the first 24 h of operation. The successful uranium operation at FRIB sets a new record for accelerated uranium beam power above 10 kW and opens a new avenue of research with rare isotopes. Published by the American Physical Society 2024

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Rapid Oxidation of Uranium Steel Alloys: Combustion Synthesis

This project explored the use of combustion synthesis as a rapid, high-temperature method for oxidizing uranium-bearing steel alloys. Traditional laboratory-scale synthesis methods often fail to replicate the thermal and kinetic conditions experienced by real-world particulates, particularly those formed under rapid quenching or high-temperature scenarios. Combustion synthesis offers a promising alternative by enabling fast, localized heating and flexible precursor selection. A series of targeted experiments were conducted using a U 2 NiCrFe 4 alloy as the precursor. The alloy was oxidized using combustion synthesis reactions fueled by uranyl nitrate and glycine, achieving peak temperatures exceeding 1,200 °C. Postreaction analysis using scanning electron microscopy (SEM), elemental mapping, and Raman spectroscopy revealed the formation of iron-based oxides, with limited but detectable evidence of uranium oxide phases such as UO 2 . The results indicate that under the rapid reaction and cooling conditions of combustion synthesis, iron oxides form preferentially, but uranium oxide formation is kinetically limited. These findings validate combustion synthesis as a viable method for simulating the oxidation behavior of uranium steels in extreme environments and lay the groundwork for future studies aimed at enhancing uranium oxide formation through higher temperatures or modified precursor compositions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Understanding neutron capture processes in uranium deposits using combined U-Sm-Nd isotopic compositions

Valuable insights into the history and evolution of a geologic deposit can be found by investigating neutron capture reactions. Thermal neutron capture reactions occur within both the samarium (Sm) and the uranium (U) systems, where 149 Sm and 235 U can capture neutrons to become 150 Sm and 236 U, respectively. Although largely unexplored, paired measurements of 150 Sm and 236 U could be important for understanding neutron capture effects within uranium ore bodies, and such measurements are potentially useful in nuclear forensics for assessing a material's provenance or mineral exploration. In this work, we refined measurement procedures of Sm isotope compositions utilizing MC-ICPMS. While geologic reference materials were found to have indistinguishable Sm isotope compositions, we found significant isotope variations consistent with nuclear field shift among synthetic Sm standards. Here, this observation highlights that future high-precision Sm isotope investigations need to carefully evaluate synthetic standard(s) against geologic reference materials until an unfractionated and agreed-upon standard is identified. Here, we applied this method to a set of nine uranium ores from the South Australian Beverley North uranium deposits. Although 236 U excesses had been previously reported for these U ores, we found no measurable isotopic shifts in 149 Sm- 150 Sm at the current level of precision (±5 parts per million). One possible explanation for this disparity in the observed neutron capture signatures between U and Sm is that the source(s) of the U and Sm in these ores may be decoupled. This is consistent with the finding that these ores have variable 143 Nd/ 144 Nd, thus demonstrating that diverse sources were involved in the formation of the Beverley North deposits. Alternatively, this deposit may be too young (<50Ma) to have accumulated measurable neutron capture effects in Sm to be detected with the methods employed here.

Mineral exploration↗
DOE OSTI2025

Development of the technological process for the IGR reactor's highly-enriched irradiated uranium-graphite fuel immobilization

The immobilization of irradiated highly enriched uranium (HEU) fuel is a critical component of nuclear waste management and non-proliferation efforts. In Kazakhstan, at National Nuclear Center of the Republic of Kazakhstan special attention is given to managing legacy HEU fuel from research reactors. One such case involves the IGR research reactor, whose first core containing irradiated HEU uranium-graphite fuel was operated from 1961 to 1966 and removed following reactor modernization. This fuel now requires a reliable and secure immobilization strategy. Here, this paper presents the development of a technological process for immobilizing this fuel to reduce its enrichment to below 20 % in terms of 235U content. The proposed method involves down-blending irradiated HEU fuel with depleted uranium, followed by encapsulation in a Portland cement matrix. Full-scale experiments were conducted to assess the uniformity of uranium distribution within the matrix. The results confirm the effectiveness of this approach, ensuring reliable immobilization of fuel in accordance with international requirements, including IAEA standards and Kazakhstan's regulatory framework. These findings contribute to the broader effort of adapting immobilization strategies for the safe management of spent fuel from research reactors.

12 - MGMT OF RADIOACTIVE AND NON-RADIOACTIVE WASTE↗
DOE OSTI2025

Chemical speciation correlated with microstructural heterogeneity of interdicted uranium materials

Two uranium powders seized by law enforcement in Victoria, Australia, have been characterized by established nuclear forensic methods in a previously published study. Here, in this work, the results of further characterization by a scanning transmission x-ray microscope (STXM) operating in the soft x-ray regime are reported. STXM images are used to estimate the elemental distribution in micrometer-scale particles of each powder, and oxygen K-edge absorption spectra are used to determine the chemical state of uranium. The results of the current study are consistent with the previous analysis; the first powder is found to be a potassium-uranium hydrate, while the second powder is determined to be a mixture of uranium oxides primarily consisting of UO 3 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Inverse Trans Influence and Uranium-Arene σ-Bonding Drive Molecular Geometry: Ligand Modification from Hard to Soft Flips the Oxide from Axial to Equatorial

A rare example of an equatorially bound terminal uranium(V) oxo complex in a chelating sulfur-based ligand environment, namely [(mes( Me,Ad ArS) 3 )U V (O eq )(THF)] (2), is presented. Octahedrally coordinated 2 is obtained by reaction of the mesitylene-anchored tris-thiophenolate-coordinated uranium(III) complex [U III ((SAr Ad,Me ) 3 mes)] (1) with the oxygen-atom transfer reagent N 2 O. The observed, equatorially bound oxo ligand in 2 is in stark contrast to its known tris-aryloxide analog, [(mes( Me,Ad ArO) 3 )U V (O ax )(THF)] (A), where the oxo ligand occupies the typically observed axial coordination site. Complexes 1 and 2 are characterized by single-crystal X-ray diffraction analyses and spectroscopic and magnetochemical methods, including 1 H NMR, UV/vis/NIR electronic absorption, as well as EPR spectroscopy and SQUID magnetometry, thus confirming the C S symmetry and the pentavalent oxidation state of 2. Encompassing quantum chemical calculations (DFT and CASPT2) on 2 and its tris-phenolate analog A, support and rationalize the structural and electronic differences. The molecular orbital pictures show that a stabilizing σ-bonding interaction arising from the U–O eq inverse trans influence (ITI) is present in 2 but missing in A. In 2, the sulfur 3p orbitals are closer in energy to the uranium 5f manifold than the arene π-system, leading to an ITI, while U–arene σ- or δ-bonding is not observed. Although the arene orbitals remain separated from the uranium 5f orbitals in A, the absence of an ITI allows the arene a 2u orbital to engage in a σ-type interaction with the metal. Thus, incorporating a tris-thiophenolate to an arene anchor introduces a new design concept in molecular f-element chemistry. This approach stabilizes an equatorially bound U(V) oxo center, contrasting with its tris-phenolate counterpart, where oxo coordination is axial. The observed geometric divergence, driven by competing ITI and U–arene interactions, not only tunes electronic structure but also leads to differentiated reactivity: only the phenolate analogs activate H 2 O, while the thiolates do not.

Hydrocarbons↗
DOE OSTI2024

Photoluminescence of a Uranium(IV) Alkoxide Complex

In this report, we describe the photoluminescence of a homoleptic uranium(IV) alkoxide complex. Excitation of [Li(THF)] 2 [U IV (O t Bu) 6 ] leads to the first example of photoluminescence from a well-defined actinide complex originating from an f–f excitation, supported by second order multiconfigurational electronic structure calculations including spin–orbit coupling. These calculations show strong spin–orbit coupling between the excited triplet and singlet states for the 5f-orbital manifold, which leads to a long-lived excited state lifetime of 0.85 s at low temperature. The photophysical properties of homoleptic uranium(V) and uranium(VI) tertbutoxide complexes are also presented; we find that oxidation of the uranium(IV) alkoxide results in quenching of luminescence in [Li(THF)][U V (O t Bu) 6 ] and [U VI (O t Bu) 6 ]. This is attributed to competing ligand to metal charge transfer absorption processes shifted to lower energy upon oxidation of the actinide center, which mask the relevant f–f transitions in the visible region of the electronic absorption spectrum.

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

Novel Synthesis and Characterization of Uranium-Zirconium Carbonitride by Direct Casting

UZrCN has exhibited thermophysical properties beneficial to high temperature reactor applications such as nuclear thermal rockets. The present work investigates a novel liquid phase synthesis method involving the admission of nitrogen gas during arc melting of uranium, zirconium, and carbon. Initial microstructural examinations using SEM indicated that the samples remain heterogenous with zirconium-rich cores in a uranium-rich matrix. Heterogeneity resulted from large differences in melting temperature of the major constituents and rapid solidification. Additional EDS, combustion analysis, and inert gas fusion analysis proved that the core regions have both uranium and zirconium and establish that carbon and nitrogen are retained during the fabrication process. XRD clarifies that the light elements did not form compounds with the uranium matrix but rather incorporated into the core region forming a sub-stoichiometric UZrCN. Heat treatment performed on equimolar U-Zr-C resulted in an increase in the homogenous phase present with an XRD pattern reflecting a ternary UZrCN despite small amounts of segregation in the final button.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Utilizing Gamma Signals to Find Optimal Uranium Wells

Uranium is a very important resource when using nuclear power. Only a small fraction of the uranium we use in the US is domestically sourced. Our goal is to effectively find and mine uranium in a way that is generally accurate and not difficult. Using computer science and machine learning, we want to automate a reasoning system that geologists use to analyze where the uranium ore bodies are. Presenting a general explanation of the goals and impacts of this project for the High School Intern showcase.

58 - GEOSCIENCES↗
DOE OSTI2025

On Uranium-Zirconium Carbonitride Formation by Direct Casting - A Novel Synthesis Study

UZrCN has exhibited thermophysical properties beneficial to high temperature reactor applications such as nuclear thermal rockets. The present work investigates a novel liquid phase synthesis method involving the admission of nitrogen gas during arc melting of uranium, zirconium, and carbon. Initial microstructural examinations using SEM indicated that the samples remain heterogenous with zirconium-rich cores in a uranium-rich matrix. Heterogeneity resulted from large differences in melting temperature of the major constituents and rapid solidification. Additional EDS, combustion analysis, and inert gas fusion analysis proved that the core regions have both uranium and zirconium and establish that carbon and nitrogen are retained during the fabrication process. XRD clarifies that the light elements did not form compounds with the uranium matrix but rather incorporated into the core region forming a sub-stoichiometric UZrCN. Heat treatment performed on equimolar U-Zr-C resulted in an increase in the homogenous phase present with an XRD pattern reflecting a ternary UZrCN despite small amounts of segregation in the final button.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Rapid separation of radiopure yttrium-91 for tracer studies from mixed fission products and uranium

A novel, single step method for isolating 91 Y from irradiated uranium and mixed fission products has been developed based on the commercially available Eichrom LN resin. The separation procedure allows for loading an LN resin column with an irradiated uranium solution, containing mixed fission products, in dilute HCl where both uranium and yttrium are retained on the resin. Eluting dilute (0.5–1 M) HCl will strip the majority of fission products. Furthermore, increasing the concentration to 3 M HCl will elute 91 Y without the presence of any other fission products. Finally, uranium can be recovered by passing concentrated HCl through the column.

Chromatography↗
DOE OSTI2025

Multiple oxidation states of uranium stabilized by an O , N , O -ligand

Elements of the 5f row maintain a wide variety of oxidation states that have been exploited in synthesis, catalysis, and separations. Herein, we describe the complexation of an O,N,O chelator, ExPh, with uranium in the (IV) and (VI) (uranyl) oxidation states. These two uranium complexes, U(IV)ExPh and UO 2 ExPh, respectively, were characterized in the solid state via single-crystal X-ray diffraction (SC-XRD) analysis. Electrochemical studies using cyclic voltammetry were employed to investigate the multiple redox events associated with the U(IV) complex. Spectro-electrochemical analysis of U(IV)ExPh provided spectroscopic evidence of a stable U(V) species. Chemical oxidation of U(IV)ExPh allowed isolation of the U(V) complex, U(V)ExPh. All three stable uranium complexes produced in this study were characterized via IR, UV-vis and NMR spectroscopies, and micro-spectrophotometry. On the other hand, efforts to reduce U(IV)ExPh to the corresponding U(III) species or produce this putative complex directly from ExPh failed to yield an isolable product. The stabilization of three formal oxidation states of uranium, coupled with previous lanthanide-row results, paves the way for studies of ExPh and its analogues in minor actinide chemistry.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2024

Computational study of tungsten and depleted uranium photoneutron targets for a 20 MeV electron linear accelerator

Neutron production can be realized with a high energy electron linear accelerator by using Bremsstrahlung and photoneutron converters. In this study, Monte Carlo N-Particle Code (MCNP) was used to evaluate potential photonuclear target designs for a high energy electron linear accelerator for applications such as neutron radiography and neutron resonance spectroscopy. A computational model was developed to inform a target design that would yield a high number of neutrons. It consists of a 20 MeV electron beam incident on a Bremsstrahlung target and a photonuclear target to generate neutrons. This computational model showed that a thickness of 0.75 inches for both tungsten and depleted uranium yields the most neutrons from photoneutron reactions. Saturation in the total number of generated neutrons was observed at over 0.75-inch thickness for both evaluated materials. Depleted uranium yielded approximately twice the number of neutrons overall compared to tungsten. The highest neutron surface flux for Depleted Uranium was 1.06 × 10-4 neutrons/cm2/source electron, and for Tungsten it was 5.12 × 10-5 neutrons/cm2/source electron. The optimal target design for this study’s application would consist of a 0.75 inch-thick block of depleted uranium with the length, width, and/or diameter varying dependent on application.

43 PARTICLE ACCELERATORS↗
DOE OSTI2024

Radiation Dose Modeling for Niowave’s Accelerator Driven Uranium Target Assembly 3

Molybdenum-99 is a high-value radionuclide commonly used for medical purposes within the United States. The National Nuclear Security Administration (NNSA) seeks to reliably produce the radioisotope 99 Mo without the use of highly enriched uranium. NNSA’s Office of Material Management and Minimization (M3) provides funding and government laboratory expertise to private companies to expedite the production process domestically and currently funds designs that use low-enriched uranium or other 99 Mo production pathways. Several production designs are being explored across the industry, including uranium fission and photonuclear conversion of 100 Mo targets. Niowave Inc. seeks to produce 99 Mo via a high-energy electron accelerator that strikes a lead-bismuth eutectic target that ultimately produces a consistent neutron flux. The neutron flux then interacts in a subcritical reactor core configuration to produce fission in low-enriched or natural uranium targets. These fissionable targets are then processed to extract 99 Mo. The purpose of this work is to estimate the neutron and photon dose response across Niowave’s proposed facility for worker safety during operation. Owing to the size of the proposed Niowave facility and necessary shielding, unbiased Monte Carlo radiation transport is impractical, and variance reduction methods are required. This work focuses on the weight window variance reduction method to produce high confidence dose response results within a Monte Carlo radiation transport code. Specifically, an adjoint-informed weight window methodology was created to improve the dose response estimates for accelerator-driven subcritical reactor designs. This adjoint-informed methodology was implemented for Niowave’s proposed design and improved dose results at far-field locations across the facility. Acceptable dose rate contours for the proposed facility were generated across the facility and are presented in this work.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2026

Development of a Chlorine Cathode and Anode Basket Assembly for Production of Uranium Chloride

A chlorine cathode has been developed for in situ chlorination of metals, oxides, and oxychlorides in molten chloride electrolytes that could be used to support synthesis of chloride fuel salts for molten salt reactors. The chlorine cathode is designed to electrochemically reduce chlorine gas to generate chloride ions that, when paired with a metal anode, chlorinate that metal as it is oxidized into the salt. The designed porous carbon electrode effectively distributes Cl 2 to the electrode surface and efficiently generates Clions in the molten chloride electrolyte. This report highlights recent improvements made to the electrode assembly, with a focus on operational control of the anode basket stability. Synthesis tests demonstrated the successful chlorination of uranium metal, resulting in 3.6 wt% uranium generated in the LiCl-KCl base salt in 45 minutes, performed in a bench-scale chlorination apparatus. Higher concentrations could be achieved by applying longer chlorination times or increasing the amount of uranium loaded in the anode basket. Overall, the chlorine cathode can be used with an appropriately designed anode to chlorinate uranium in situ for synthesis of molten salt reactor fuel salts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Commercialization of High-Density High Assay Low Enriched Uranium Fuel Systems

The Office of Reactor Conversion and Uranium Supply (NA 231) at the National Nuclear Security Administration leads the conversion effort for the United States High Performance Research Reactors (USHPRR). These reactors are the final civilian reactors in the US to transition from High Enriched Uranium (HEU) to high assay low enriched uranium (HALEU). Each of these reactors represents unique capabilities and no currently available fuel system meets their needs for conversion. The Fuel Fabrication (FF) Pillar of the USHPRR project is responsible for the fabrication of experimental elements, conversion elements, and establishing a commercial economical production capability. FF is also responsible to share with the domestic and international community the theoretical knowledge gained. Other pillars within the USHPRR project provide the experimental and conversion fuel designs, assist the reactors with licensing activities, and ensure the entire fuel cycle is evaluated. Over the last decade, FF has worked with the production partners at Y-12 National Security Complex (Y-12) and BWXT Nuclear Operations Group, Research and Test Reactors (BWXT). Y-12 has begun processing the alloy feedstock for the conversion elements with a qualified process. BWXT has started the final fabrication of the experimental elements. Once the experimental elements are complete, BWXT will begin conversion element fabrication. The FF Pillar resides at Pacific Northwest National Laboratory (PNNL) and uses PNNL, universities, commercial vendors, and the DOE national laboratory system to evaluate process development activities to improve the process steps. FF supports the fabrication of two high density fuel systems, monolithic U-10Mo (Figure 1) and Uranium Silicide (Figure 2). The U-10Mo fuel system is further along the development process. FF assists in long term planning with the production partners. This includes ramping production of the elements from experimental quantities to annual steady state needs. As part of the ramp up, opportunities to improve yield and product quality are identified to ensure the fuel systems are cost effective.

Catalan, Michael A. [BATTELLE (PACIFIC NW LAB)]↗