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Results for “uranium metal”

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 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Application of HRGS for forensic characterization of uranium oxides, pure uranium metals and uranium alloys

A nondestructive iterative method for uranium-bearing material characterization with HRGS developed earlier in (J. appl. Rad. Isotopes, 166 (2020) 109433) is applied to determine matrix densities and uranium isotope masses of a variety of uranium materials, namely uranium ore, UO 2 and U 3 O 8 powders, fuel elements in the form of UO2 microspheres, uranium metal and uranium alloys. It is shown that U 3 O 8 powders with uranium mass fraction of about 84% can be distinguished from the powders of UO 2 with uranium mass fraction of about 87%; uranium products in the form of liquid or loose powder with matrix density of 0.5-2.0 g/cm 3 can be distinguished from uranium products in the form of compacted fuel elements with matrix density of 6.0-10.0 g/cm 3 and from pure metal uranium and uranium alloys with matrix density of 14.0- 19.0 g/cm 3 . In fuel microspheres based on UO2 the uranium mass fraction 88.02% measured by HRGS is consistent, within the measurement uncertainties, with the results of isotope dilution mass spectrometry 87.76±0.64 % and also is confirmed by X-ray diffraction technique. The uranium mass fraction of the uranium ore estimated as 0.08% by HRGS is consistent, within the measurement uncertainties, with the value 0.09±0.01% determined with WDXRF. Densities of two different uranium metal samples, estimated as 18.42 g/cm 3 and 19.33 g/cm 3 by HRGS are consistent with values 18.24±0.55 g/cm 3 and 18.86±0.59 g/cm 3 , respectively, obtained by the gas pycnometry technique.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Synthetic diversity in the preparation of metallic uranium

Uranium metal is associated with several aspects of nuclear technology; it is used as fuel for research and power reactors, targets for medical isotope productions, explosive for nuclear weapons and precursors in synthetic chemistry. The study of uranium metal at the laboratory scale presents the opportunity to evaluate metallic nuclear fuels, develop new methods for metallic spent fuel reprocessing and advance the science relevant to nuclear forensics and medical isotope production. Since its first isolation in 1841, from the reaction of uranium chloride and potassium metal, uranium metal has been prepared by solid-state reactions and in solution by electrochemical, chemical and radiochemical methods. The present review summarizes the methods outlined above and describes the chemistry associated with each preparation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Quantitative Morphological Characterization of Carbide Inclusions in Uranium Metal

Uranium carbides (UCs) are prevalent inclusions in U metal that form during melting operations from interactions with crucible walls and the casting chamber atmosphere. Although UCs have been studied extensively since the beginning of U metal foundry operations, there are still unknowns regarding the effects of thermomechanical processing on their sizes and morphology. Here, we present the results of a series of controlled cooling experiments with molten uranium to elucidate the effect of cooling rate on inclusion morphology in a-U. Samples were melted using a vacuum induction melter and manually cooled at rates of 2.5, 1.7, 1.1, 0.8, 0.3 (± 1%) K/s from ~1600 K to <700 K in under 1 hour. Subsequent scanning electron microscopy (SEM) was performed on cross-sections of the samples, revealing a complex mixture of UC morphologies that are indicative of diffusion and growth influenced by the thermal processing of the U matrix. Image analysis using the Morphological Analysis of Materials (MAMA) software showed that UC sizes generally grew larger with slower cooling rates, and the two slowest cooling rates noticeably impacted the inclusion circularity and ellipse aspect ratio. These results indicate that UC morphology is sensitive to short cooling rates (<1 hour) and could therefore be controlled in the production of metallic nuclear fuels. Additionally, inclusion speciation and morphologies could potentially provide forensic clues about processing history of unknown metal samples. Understanding the driving forces involved in UC morphology evolution is beneficial for evaluating metal fuels for next generation nuclear reactors and for identifying signatures for nuclear forensics.

Athon, Matthew T.↗
DOE OSTI2024

Quantification of uranium metal and total uranium in U-UO 2 mixtures using H 3 PO 4 -HCl

Here, the selective dissolution of uranium metal from its mixtures with UO 2 was studied to establish a simple and rapid quantitative method. Uranium metal was found to selectively dissolve as U(IV) into a solution of phosphoric and hydrochloric acids, with quantitative dissolution occurring within 1 h. Total dissolution of uranium as U(VI) could then be obtained by nitric acid addition and heating for 0.5–1 h. Analysis of the resulting U(IV) and U(VI) solutions by ICP-OES or UV–Vis spectroscopy yielded uranium metal-to-UO 2 ratios with an accuracy of ± 10% or better. Accuracy of the ICP-OES method was slightly better than UV–Vis, but the simplicity of the latter makes it an attractive option for some applications.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Convenient Syntheses of Trivalent Uranium Halide Starting Materials without Uranium Metal

Low-valent uranium coordination chemistry continues to rely heavily on access to trivalent starting materials, but these reagents are typically prepared from uranium turnings, which are becoming increasingly difficult to acquire. Here we report convenient syntheses of UI 3 (THF) 4 (THF = tetrahydrofuran) and UBr 3 (THF) 4 from UCl 4 , a more accessible uranium starting material that can be prepared from commercially available uranium oxides. UCl 3 (THF) 2 (1), UBr 3 (THF) 4 (2), and UI 3 (THF) 4 (3) were prepared by single-pot reductions from UCl 4 using KH and KC 8 and converted to 2 or 3 by halide exchange with the corresponding Me 3 SiX (where X = Br or I). Reduction of UI 4 (Et 2 O) 2 (4; Et 2 O = diethyl ether) and UI 4 (1,4-dioxane) 2 (5) was also shown to cleanly yield 3. Complex 1 was also synthesized separately by the addition of anhydrous HCl to U(BH 4 ) 3 (THF) 2 , which was prepared by thermal reduction of U(BH 4 ) 4 . All three trivalent uranium halide complexes were isolated in high crystalline yields (typically 85–99%) and their formulations were confirmed by single-crystal X-ray diffraction, elemental analysis, and 1 H NMR and IR spectroscopy. Elemental analysis conducted on triplicate samples of 1–3 exposed to vacuum for different time intervals revealed significant THF loss for all three complexes in as little as 15 min. Altogether, these results offer expedient entry into low-valent uranium chemistry for researchers lacking access to uranium turnings.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Grain growth kinetics of the gamma phase metallic uranium

We report metallic uranium is a leading fuel form for sodium cooled fast reactors as an enabling technology of future nuclear energy systems. Mechanistic understanding of fuel behaviors and kinetics under thermodynamic equilibrium and highly non-equilibrium conditions are essential for evaluating fuel performance. It is important to understand and predict the grain and pore evolutions of metallic fuels under thermal and irradiation conditions. However, very limited data are available on the grain growth kinetics and mechanisms of pure gamma phase uranium. In this paper, the pure gamma uranium pellets with different grain structures were fabricated by combining high-energy ball milling and spark plasma sintering. Isothermal annealing tests were performed to investigate the grain growth behavior of the pure gamma phase uranium with different initial grain sizes. A parabolic relationship in grain growth with time was identified for the submicron-sized (374 nm) sample. In contrast, for the nano-sized (137 nm) sample, the grain growth shows a linear relationship with time. The activation energies of grain growth were determined as 199.5 KJ/mol and 80.6 KJ/mol for nano-sized and submicron-sized grain structures, respectively. For the nano-sized sample, the rate-control step of grain growth is dominated by the triple-junction migration, in which the grain boundary triple junction drags the grain growth, leading to a higher activation energy than the bulk diffusion. The dominating mechanism for the submicron-sized sample is grain boundary diffusion. The mechanistic understanding and critical data obtained on the kinetics of pure uranium phases will be useful to evaluate fuel behavior under thermodynamic equilibrium conditions and develop a high fidelity model to predict fuel performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Results from a controlled depleted uranium metal casting experiment designed to investigate nuclear Forensic Radiochronometry Signatures

The accurate interpretation of uranium metal 230 Th/ 234 U and 231 Pa/ 235 U radiochronometry model ages requires an understanding of how uranium parent nuclides and decay progeny ( 230 Th and 231 Pa) behave during uranium metal casting. In order to directly measure the spatial distribution of 230 Th and 231 Pa in uranium metal before and after vacuum induction melting (VIM), Los Alamos National Laboratory identified uranium metal feedstock, characterized the metal feedstock, conducted a controlled casting experiment of an approximately 120 kg uranium metal rod, and characterized the cast metal. Herein this study presents radiochronometry results and quantified 230 Th and 231 Pa VIM separation factors from bulk uranium.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Stabilization of Uranium Metal by Chloride Conversion-Economical, Safe, Scalable [Factsheet]

Researchers at Los Alamos National Laboratory have developed SUCCESS, a process to convert uranium metal to uranium tetrachloride enabling production of these valuable compounds for actinide science, molten salt nuclear reactors, and energy research. Conventional methods are problematic due to the oxide reduction processes use rare reagents and generate significant amounts of waste and hydride conversion process using chlorine at high temperature is hazardous. This new method using inexpensive reagents chemically transform uranium metal into uranium halide compounds under mild conditions. The facile conversion to stable uranium tetrachloride could reduce the need to mine uranium for uses other than nuclear, by recycling the uranium without oxidation. The Laboratory is interested in supporting further scale-up efforts using a Cooperative Research and Development Agreement (CRADA) or by granting a license to a qualified entity

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Radiochronometric discordance in cast uranium metal: a multi-laboratory intercomparison exercise

Here, the model age of a nuclear material is crucial in nuclear forensic analysis. Uranium metals with complex production histories often exhibit discordant model ages from the 230 Th– 234 U and 231 Pa– 235 U chronometers. Recent studies involving targeted uranium metal castings have enhanced our understanding of decay product behavior during casting, aiding nuclear forensic interpretation. Building on this prior work, forensics laboratories at Atomic Weapons Establishment (AWE), Lawrence Livermore National Laboratory (LLNL), and Los Alamos National Laboratory (LANL) conducted an interlaboratory comparison to investigate spatial heterogeneity in uranium metal cast under controlled conditions. Each laboratory measured samples of a mixed feedstock and its corresponding cast product. This work furthers our understanding of discordant model ages and the use of discordance as a signature to enhance confidence in interpretations of radiochronometric data for nuclear forensics.

230Th/234U↗
DOE OSTI2021

Uranium Carbide and Uranium-Metal Thermal Scattering Law Evaluations and Cross Sections [Slides]

New Uranium Carbide and Uranium Metal evaluations have been developed using the FLASSH platform. Ab initio LD/MD models converged electronic, magnetic, and dynamical properties and demonstrate good agreement with experimental data. Thermal scattering cross sections generate inelastic and elastic thermal neutron scattering libraries in FLASSH . New TSL evaluations occurred in the production of ENDF File 7 data for Uranium Carbide and Uranium Metal submitted to NNDC for ENDF/B-VIII.1 release.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Effects of uranium metal carbon content on hydriding kinetics and corrosion blister number/area at sub-ambient pressures

Carbon is a common impurity in uranium metal, resulting in a number of uranium–carbon inclusion phases that contribute to an increase in metal defect density as carbon content increases. It is widely held that uranium hydride corrosion preferentially nucleates at these defect sites, and that an increase in carbon content will therefore represents an increase in uranium hydride corrosion sites on the metal surface. We hydrided six uranium sources with differing carbon contents to explore whether this assumption holds in a sub-ambient (~ 0.1 atm hydrogen), sealed environment, and report the resulting reaction kinetics and uranium hydride blister benchmarking data. We find that carbon content is not strongly correlated with reaction kinetics terms or the resulting hydride blister number and area, but that there is a tight relationship between corrosion blister number/area and kinetics as is expected. Further, we find that there is a strong trend of decreasing variance in the blister number, blister area, and induction time as carbon content increases (higher carbon content results in more reproducible blister populations). Additionally, we find a narrow band of uranium metal consumption at the end of the parabolic phase of reaction progress (beginning of linear growth phase) of 0.098 ± 0.011 w/w%, a fact that may be useful in assaying hydrogen corrosion of uranium metal within sealed environments generally.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Three Delayed Critical 15-Inch-Diameter Interacting Enriched (93.14) Uranium Metal Cylinders Without Moderator and Reflector

This report documents very accurately the configuration and the materials for experiments with three unmoderated, unreflected, interacting, coaxial, highly enriched, 15-in.-diameter, uranium metal cylinders performed at the Oak Ridge Critical Experiments Facility (ORCEF) in August and September 1963 and described in logbook E 20 associated with experiments in the East cell of ORCEF. The information is sufficiently accurate that it can be used as the basis for preparation of benchmarks for International Criticality Safety Benchmark Program (ICSBEP) at Idaho National laboratory. The thickness of the cylinders was varied and the spacing between them was adjusted to achieve a delayed critical configuration. The average enrichment of the uranium metal was 94.14 wt. % 235U. The heights of the 15-in.-diameter, equal-height cylinders varied from 1-1/8to 2.00 inches. All interacting cylinders were assembled coaxially with their flat faces parallel and their combined masses varied between 182 and 325 kilograms of HEU metal. The data from these six experiments described are judged to be acceptable for use as criticality safety benchmark experiments for the ICSBEP and EURATOM’s Nuclear Energy Agency nuclear criticality safety benchmark program, once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, it is expected that the uncertainties in measured keff could be as low as ± 0.0002.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Two Delayed Critical 15-Inch-Diameter Interacting Enriched (93.14) Uranium Metal Cylinders without Moderator and Reflector

This report documents very accurately the configuration and the materials for experiments with two unmoderated, unreflected, interacting, coaxial, highly enriched, 15-in.-diameter, uranium metal cylinders performed at the Oak Ridge Critical Experiments Facility (ORCEF) in May to Aug 1963 and described in logbook E-19 and E -20 associated with experiments in the East cell of ORCEF. Measurements were also performed in April and May of 1965 and described in logbooks E-22 and E-23 The information is sufficiently accurate that it can be used as the basis for preparation of benchmarks for International Criticality Safety Benchmark Program (ICSBEP) at Idaho National laboratory. The thickness of the cylinders was varied and the spacing between them was adjusted to achieve a delayed critical configuration. The average enrichment of the uranium metal was 94.14 wt. % 235 U. The heights of the 15-in.-diameter, equal-height cylinders varied from 1-5/8 to 3.0-inches. All interacting cylinders were assembled coaxially with their flat faces parallel and their combined masses varied between 182 and 325 kilograms of HEU metal. The data from these 12 experiments described would be acceptable for use as criticality safety benchmark experiments for the ICSBEP and EURATOM’s Nuclear Energy Agency nuclear criticality safety benchmark program, once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, it is expected that the uncertainties in k eff could be as low as ± 0.0002.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Energy–structure–property relationships in uranium metal–organic frameworks

Energy–structure–property relationships in uranium metal–organic frameworks offer a deeper understanding of their behaviors, explaining why structure produces function. This understanding becomes crucial for the rational design of such materials with enhanced properties.

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

Model ages of three uranium metal CRMs and implications for radiochronometry data interpretation

In this study, we report 230 Th/ 234 U and 231 Pa/ 235 U model ages in three uranium metal certified reference materials with distinct production histories and isotopic compositions ranging from depleted to highly enriched (CRMs 112-A, 115, and 116-A). Here, our results provide the first model age values for these materials using multiple radiochronometers, which can be used as reference values for quality control of radiochronometry measurements performed during nuclear forensic investigations. Furthermore, we evaluate the model ages in the context of available production records for the materials, which provides insight into the processes that may impact radiochronometry systems during uranium metal production.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Conversion of Uranium Metal to Uranyl Sulfate Solution

Experiments involving conversion of uranium (U) metal to U 3 O 8 , followed by dissolution to form uranyl sulfate solution, were performed with batches of up to 300 g of U metal. The focus was to identify major U species formed during the oxidation process, determine the particle size distribution of the uranium oxide product, and identify the size of the dissolution vessel needed to accommodate the exothermic reaction during the conversion of U 3 O 8 to uranyl sulfate using 30% hydrogen peroxide and H 2 SO 4 . The information obtained in this study will be used for designing full-scale processes to convert kilogram quantities of U metal to uranyl sulfate solution with a concentration of 300 g U/L.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

Delayed Critical Enriched Uranium Metal, 7-in.-diam. Cylinder with Thin Stainless Steel Top and Bottom Reflectors

Three 7 in. diameter highly enriched uranium (HEU; 93.17 wt % 235 U) metal cylinders were assembled on the vertical assembly machine of the Oak Ridge Critical Experiments Facility (ORCEF) with thin stainless steel reflectors (0.5, 1, and 2 in. thick) on the top and bottom. These experiments were performed during January 13–25, 1965, and used seven operational days at ORCEF. Before these experiments, unreflected and unmoderated 7 in. diameter HEU metal cylinders had been assembled to delayed criticality at ORCEF, and the results were benchmarked in HEU-MET-FAST-051. In addition to the critical experiments, the prompt neutron decay constant was measured via the Rossi-α technique, and those results are also presented in this report. To achieve near-delayed-critical systems, 5 × 5 × 0.03125 in. rectangular uranium metal parts were used on top of 7 in. diameter cylinders inside the top reflector. The delayed critical configuration from the three experiments described herein are acceptable for use as criticality safety benchmark experiments for the International Criticality Safety Benchmark Evaluation Program (ICSBEP), coordinated by the Nuclear Energy Agency, once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, the uncertainty in $k_{eff}$ is expected to be as low as ±0.0002. The prompt neutron time decay data could be the basis of an International Reactor Physics Evaluation Program benchmark.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Delayed Critical Highly Enriched Uranium Metal Cylinders with Thin Graphite Top and Bottom Reflectors

Nine 7, 11, and 15 in. diam highly enriched uranium (HEU, 93.15 wt % 235 U) metal cylinders were assembled on the vertical assembly machine in the Oak Ridge Critical Experiments Facility (ORCEF) and had 1, 2, or 3 in. thick HLM graphite reflectors on the top and bottom. The experiments, which were performed between April 3, 1970, and February 18, 1971, used 23 operational days at ORCEF. Before those experiments were carried out, unreflected and unmoderated, graphite- and polyethylene-reflected, and polyethylene-moderated HEU metal cylinders had been assembled to obtain delayed criticality at ORCEF in the 1960s and reported by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) at the Nuclear Energy Agency (NEA)*. The data from the nine critical experiments are acceptable for use as criticality safety benchmark experiments for the NEA’s ICSBEP once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with HEU metal at ORCEF, the uncertainties in k eff are expected to be as low as ±0.0004.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗