Results for “uranium” — Page 2

DOE OSTI2025

Rapid isotopic analysis of uranium microparticles via SP-ICP-TOF-MS

Inductively coupled plasma – time-of-flight – mass spectrometry (ICP-TOF-MS) was employed for the isotopic analysis of uranium particles of varying 235 U enrichment levels. Here, a single particle (SP)-based introduction scheme was employed such that individual particles, in a suspension, were analyzed. The uranium oxide microparticles were comprised of depleted uranium (DU, 235 U/ 238 U of 0.0017316(14)), natural uranium (NU, 235 U/ 238 U of 0.0072614(39)), and low enriched uranium (LEU, 235 U/ 238 U of 0.051025(15)). The percent relative difference of the SP-ICP-TOF-MS measured isotopic ratios compared to the expected values for the DU, NU, and LEU particle populations were 8.75, 0.12, and 1.23 %, respectively. After characterization, the DU and NU particles were doped within a complex sample matrix (Arizona Test Dust) containing Fe, Ti, Al, and Si particles, among others. Then, the suspension was analyzed via SP-ICP-TOF-MS and the detected particles were classified as DU or NU based on their measured 235 U/ 238 U ratio. In the same analysis, the matrix particles (i.e., Al, Fe, and Ti) were detected, demonstrating the simultaneous nuclide detection provided by the measurement platform. The presented SP-ICP-TOF-MS methodology for uranium particle characterization proved to be a high throughput method for detecting and isotopically discerning uranium particles with varying enrichment levels, in a complex matrix.

Stanberry, Jordan S. [Oak Ridge National Laborator↗

DOE OSTI2025

Demonstration of Electroreduction Technology to Convert GNF Uranium Oxide Powder to Metal

Global Nuclear Fuel – Americas LLC (GNF) has partnered with Argonne National Laboratory to demonstrate electroreduction of uranium oxides produced by deconversion of UF 6 to uranium metal through the Gateway for Accelerated Innovation in Nuclear (GAIN) program under the U.S. Department of Energy to accelerate the domestic production of metallic advanced reactor fuels. Electroreduction of uranium oxide was first demonstrated and patented by Argonne in the early 2000s as a technology to convert used oxide nuclear fuel from light water reactors to metal for further fuel reprocessing. More recently, electroreduction has been proposed as a front-end technology for metallization of uranium oxides produced by deconversion of UF 6 and for scrap recovery of oxide materials. During the electroreduction process, UO 2 powder is contained in a stainless-steel mesh basket with a cathode lead located in the center of the UO 2 bed. The basket is immersed in lithium chloride molten salt electrolyte containing 1 wt% lithium oxide along with a platinum anode and a nickel/nickel oxide (Ni/NiO) reference electrode. Current is applied to the cell between the cathode and anode to reduce the UO 2 to metallic uranium via a solid-state reduction reaction. Oxide ions released from the UO 2 during reduction are transported through the salt to the anode where oxygen gas is evolved. Once reduction is complete, the basket containing the metallicized uranium is removed from the salt and can be processed to remove the salt and consolidate the uranium into an ingot for use in metallic fuel fabrication. This project was performed to provide evidence of the electroreduction technology readiness level for metallization of UO 2 powder, identify and retire technical risks for industrialization of electroreduction, and accelerate the path to commercialization for metallic fast reactor fuel production. To that end, five electroreduction tests were performed with UO 2 provided by GNF and the resulting product was analyzed for the extent of conversion to metal and for impurity contents of the metal product to verify that electroreduction does not introduce impurities that would prevent use of the product in metallic fuel fabrication.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Progress Towards Synthesis of Uranium Chloride Fuel Salts Using Zinc Chloride

Reliable, scalable methods for producing high-purity actinide chloride salts are needed to support molten salt reactor fuel development and deployment. This report describes the continued development and demonstration of a bench-scale chlorination and purification apparatus using a zinc chloride-based method for synthesizing uranium chloride fuel salts. In this approach, uranium metal is chlorinated by ZnCl2 to produce LiCl-KCl-UCl3. Reaction with three aliquots of added uranium metal was used to generate a target uranium concentration of 30 wt %. While this concentration was chosen for initial testing of the apparatus and method, the final uranium concentration is not limited to 30 wt %. The zinc metal generated in the reaction forms an immiscible layer that was removed by volatilization at moderately high temperatures. Electrochemical measurements confirmed the removal of zinc and applied sensing methods indicated the uranium concentration to be approximately 25 wt %. These initial results demonstrate that the bench-scale chlorination apparatus is an effective platform for the synthesis and purification of uranium chloride salts using ZnCl2. This method shows promise for application to industry-relevant salt systems such as NaCl-UCl3. Further development is recommended to optimize reagent loading, zinc removal, and avoid possible U-Zn alloy formation.

Dulovic, Stephanie↗

DOE OSTI2025

Distinguishing fissile uranium isotopes using an active well neutron coincidence counter

Proposed thorium-based nuclear fuel cycles are likely to require quantification and verification of 233 U within nuclear material. Because of their similar fission cross sections, active neutron nondestructive assay (NDA) systems may respond similarly to 233 U and 235 U. Traditional safeguards equipment has been optimized for 235 U and 238 U quantification associated with conventional uranium/plutonium fuel cycles and may not be directly applicable to 233 U quantification when mixed with other actinides. This work used models of the large volume active well coincidence counter (LV-AWCC) at Oak Ridge National Laboratory to evaluate the performance of this neutron NDA system to differentiate fissile uranium isotopes. The models were developed to simulate NDA system performance in response to a number of triangular radiation signature training device sources within the central cavity or well. This work predicted that the LV-AWCC can effectively differentiate 233 U from 235 U in certain modes of operation. In active mode, the LV-AWCC with the cadmium liner results in different doubles count rates between the fissile isotopes for a given fissile uranium mass. Without the cadmium liner, the uranium isotopes provide a statistically indistinguishable doubles count rate response for the fissile masses considered in this work (up to approximately 150 g). The cadmium liner serves to harden the neutron interrogation spectrum, which better exploits the notable difference in the 233 U and 235 U fission cross sections at approximately 1 eV. In passive mode, the two fissile isotopes exhibit different doubles and singles count rates regardless of liner presence because the passive source strength of 233 U is approximately 2 orders of magnitude stronger than that of 235 U due to the shorter half-life and correspondingly higher (α, n) yield. We conclude that using neutron interrogation in the LV-AWCC, two measurements are needed to quantify 233 U content in mixed uranium items. The first measurement is used to determine the total fissile uranium mass using a mode that cannot distinguish fissile isotopes (i.e., where a similar response is observed for both fissile uranium isotopes such as active doubles without cadmium or using a thermal neutron interrogation source). In conclusion, the second measurement is used to determine the 233 U content by using a differentiating technique (e.g., passive doubles, passive doubles to singles ratio, active doubles with cadmium).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2025

Geobacter sulfurreducens Immobilized Boron-Doped Diamond Electrodes for Uranium(VI) in Water Electrochemical Bioremediation

The proliferation of nuclear science and technology has resulted in an increase in nuclear waste containing uranium, posing significant risks to both human and environmental health. This study proposes the use of Geobacter sulfurreducens (G. sulfurreducens) modified boron-doped diamond electrodes to facilitate the reduction and removal of uranium(VI) from aqueous media. The bioremediation process involves electrochemically immobilizing the bacteria on a boron-doped diamond electrode (BDD). The immobilization process requires applying reduction potentials ranging from −0.40 to −0.70 V (vs Ag/AgCl (3 M NaCl)), with −0.60 V identified as the optimal potential for effective bacterial modification. The uranium source is provided by a 2.0 mM uranyl acetate solution in G. sulfurreducens growth medium. Scanning electron microscopy (SEM) reveals a highly uniform layer of uranium on the electrode surface. Energy-dispersive X-ray fluorescence spectroscopy (EDS) and cyclic voltammetry (CV) studies confirm the presence of uranium in the system. Raman spectroscopy and X-ray photoelectron spectroscopy (XPS) successfully elucidate the reduction process of U(VI) to predominantly U(IV) using a bacteria-electrode coupled system. Additionally, a comparison is made with the electrochemical removal of uranyl ions using the electrodeposition method on unmodified BDD. Results demonstrate the presence of three uranium oxide species (UO 2 , UO 3 , and U 3 O 8 ) on the BDD electrode after experimentation, in contrast to the G. sulfurreducens/BDD assembly, which achieves the predominant reduction of UO 2 2+ to UO 2 with a small quantity of UO 3 as the final species. This study highlights the efficient electrochemical removal of uranyl ions from aqueous media at the G. sulfurreducens/BDD interface through chronoamperometry, presenting a promising approach for remediating sites contaminated with radioactive materials. The findings contribute to the exploration of sustainable alternatives for managing nuclear waste, emphasizing the potential of this electrochemical bioremediation strategy.

Bacteria↗

DOE OSTI2026

Effects of Uranium Impurities in Downblended HEU on HTGR Performance

Many advanced reactor designs require fuel enriched between 5% and 20% 235 U. To assist in producing fuel at these enrichment levels, government-owned inventories of highly enriched uranium can be downblended. However, fuel produced from these inventories contain uranium impurities that are not often found when enriching natural uranium or accounted for when modeling reactor cores. To address this concern, this work models reactor designs like the X-energy Xe-100 and the Ultra Safe Nuclear Company’s Micro Modular Reactor, and compares their performance with fuel from enriching natural uranium to fuel from downblended highly enriched uranium. This paper evaluates the models based on the effective neutron multiplication factor, k eff , effective delayed neutron fraction, β eff , and energy- and spatially dependent neutron flux, ϕ, as well as the fuel, coolant, moderator, and total reactivity temperature feedback coefficients, α F , α C , α M , and α T . The results show that the fuel from downblended highly enriched uranium inventories leads to differences in each of the metrics, especially in the keff values. In the Xe-100–like and Micro Modular Reactor–like models, k eff changes by about 1400 pcm and up to 1200 pcm, respectively. Total reactivity feedback coefficients α T are negative with the impure fuels and the keff values remain above 1 for each core configuration and fuel composition. These results show that the impure fuel compositions do not necessarily prevent achieving key design parameters, such as cycle length, or from operating in a safe condition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Design Rules for Carboborothermic Reduction Synthesis of High Uranium Density UB 4 –UBC Composites

Uranium borides are promising candidate fuel forms for use in advanced nuclear reactors due to their high thermal conductivity and potential for dual use as both fuel and burnable absorber. In this work, uranium tetraboride () and uranium monoboroncarbide (UBC) composite were synthesized by using industrially scalable carboborothermic reduction method. The final uranium boride phase composition is sensitive to the sample holding crucibles ( and graphite) such that graphite supply excess carbon, promoting the formation of a predominant UBC phase. The high‐temperature in situ synchrotron X‐ray diffraction of pristine –UBC show persistence , UBC, and phases while preoxidized –UBC leads to predominant and formation due to progressive oxidation and boron loss at high temperature. The oxidation behavior was further characterized using thermogravimetric analysis, allowing direct comparison with other potential accident tolerant fuels such as , , UC, and UN. The –UBC shows higher uranium loading than monolithic and demonstrates promising oxidation behavior at high temperature, pointing to its potential as an improved uranium boride‐based fuel form.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel Jana Howard1,3, Guang Yang3, Haiyan Zhao1*, Patrick Warren4, Tiankai Yao3, Steven Cavazos4 Elizabeth Sooby Wood4*, Ching-heng Shiau2 1 University of Idaho, Environmental Science, Idaho Falls, Idaho, USA 2 Boise State University, Microscopy and Characterization Suite, Idaho Falls, Idaho, USA 3 Idaho National Laboratory, Idaho Falls, Idaho, USA 4 University of Texas San Antonio, Department of Physics and Astronomy, San Antonio, Texas, USA *Corresponding author: haiyanz@uidaho.edu; elizabeth.soobywood@utsa.edu Tri-structural Isotropic (TRISO) particle fuel is the preferred choice for the newest and next-generation high-temperature nuclear reactors due to its robust construction [1]. The fuel design effectively contains fission products inside the particle at extremely high temperatures [2]. However, the effect of transition metal fission products on fuel performance and structural integrity remains largely unknown, creating uncertainties in predicting fuel performance and potential failure mechanisms [2]. For example, palladium (Pd) has a strong tendency to form intermetallic phases with uranium (U). These new phases tend to be hard and brittle which could lead to fracture of fuel kernel during irradiation [3,4]. Pd is not normally produced in high concentrations in normal fission reactions however, driving the production of Pd into the locality of a Uranium Oxycarbide (UCO) nuclear fuel provides an opportunity to closely observe the diffusion and chemical interactions. This study focuses on detailed transmission electron microscopy characterization of the intermetallic phases formed by the interaction between a UCO kernel and a Pd bar after annealing at 1100 °C for 100 hours. Figure 1 provides an overview of the UCO kernel in contact with the Pd bar. The UCO-Pd sample surface was examined using a Focused Ion Beam (FIB) Quanta 3D FEG FEI in VCD detector mode at 10kV, 83pA to reveal surface features. Several key surface features in the interaction region were observed including: (1) a small area with dendritic microstructure, (2) color contrast near the UCO kernel and Pd boundary, and (3) darker and lighter regions throughout the sample surface. To analyze diffusion behavior, FIB was used to prepare lamellae from five different areas of the UCO-Pd sample, as well as one for the as-received sample. These lamellae were examined using a Scanning Transmission Electron Microscope ThermoFisher Spectra 300 (STEM-Spectra). Energy Dispersive Spectroscopy results confirmed the diffusion between the UCO fuel and Pd bar. Figure 2 shows the identified phases of UC, UO2, and Pd dispersed throughout the interaction region. It was also found that the Pd concentration decreases as the radial distance from the UCO-Pd interface increases. The Pd concentration was 36.93% atomic fraction 11µm from the interaction region then decreased to 7.51% atomic fraction at 257 µm away from the interaction region. Figure 3 shows how concentrations of U and Pd vary across the interaction zone, highlighting the extent of diffusion. This study confirms that Pd interacts with surrounding material to form U-Pd phase and diffusion zones. These diffusion zones vary in composition depending on its radial distance from the point of contact with the Pd bar. These findings contribute to a deeper understanding of the fission product behavior in high temperature nuclear fuels, aiding in the prediction and mitigation of potential fuel degradation mechanisms. A B C Fig. 1. SEM micrographs of UCO fuel kernel in contact with solid Pd bar and the lift out locations. Figure 1A shows UCO fuel kernel in contact with solid Pd bar. Figure 1B shows a higher magnification image of the UCO fuel-Pd interaction zone. Figure 1C shows lift out sites for the lamellae. A B C Fig 2. EDS maps reveal the distribution of U, O, and Pd in location 3. Figure 2A shows the distribution of uranium. Figure 2B shows the distribution of oxygen. Figure 2C shows the distribution of palladium. A B C Fig 3. Palladium and uranium concentration across interaction zone in location 2. Figure 3A shows EDS map of lift out number 2. Figure 3B shows a zoomed in EDS map taken from the interaction zone. Figure 3C shows a line graph of uranium and palladium concentrations across the interaction zone. References: 1. B.E. Wells, N.R. Phillips, K.J. Geelhood. Pacific West Laboratory. (2021). TRISO Fuel: Properties and Failure Modes. https://www.nrc.gov/docs/ML2117/ML21175A152.pdf (Accessed January 16, 2025). 2. American Nuclear Society. TRISO Fuel Development Progresses in INL, ORNL. https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/nov13nn_fuel_reprint.pdf (Accessed January 16, 2025) 3. Clark R.A., M.A. Con

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel

Chemical Interaction of Palladium in Uranium Oxycarbide Nuclear Fuel Kernel Jana Howard1,3, Guang Yang3, Haiyan Zhao1*, Patrick Warren4, Tiankai Yao3, Steven Cavazos4 Elizabeth Sooby Wood4*, Ching-heng Shiau2 1 University of Idaho, Environmental Science, Idaho Falls, Idaho, USA 2 Boise State University, Microscopy and Characterization Suite, Idaho Falls, Idaho, USA 3 Idaho National Laboratory, Idaho Falls, Idaho, USA 4 University of Texas San Antonio, Department of Physics and Astronomy, San Antonio, Texas, USA *Corresponding author: haiyanz@uidaho.edu; elizabeth.soobywood@utsa.edu Tri-structural Isotropic (TRISO) particle fuel is the preferred choice for the newest and next-generation high-temperature nuclear reactors due to its robust construction [1]. The fuel design effectively contains fission products inside the particle at extremely high temperatures [2]. However, the effect of transition metal fission products on fuel performance and structural integrity remains largely unknown, creating uncertainties in predicting fuel performance and potential failure mechanisms [2]. For example, palladium (Pd) has a strong tendency to form intermetallic phases with uranium (U). These new phases tend to be hard and brittle which could lead to fracture of fuel kernel during irradiation [3,4]. Pd is not normally produced in high concentrations in normal fission reactions however, driving the production of Pd into the locality of a Uranium Oxycarbide (UCO) nuclear fuel provides an opportunity to closely observe the diffusion and chemical interactions. This study focuses on detailed transmission electron microscopy characterization of the intermetallic phases formed by the interaction between a UCO kernel and a Pd bar after annealing at 1100 °C for 100 hours. Figure 1 provides an overview of the UCO kernel in contact with the Pd bar. The UCO-Pd sample surface was examined using a Focused Ion Beam (FIB) Quanta 3D FEG FEI in VCD detector mode at 10kV, 83pA to reveal surface features. Several key surface features in the interaction region were observed including: (1) a small area with dendritic microstructure, (2) color contrast near the UCO kernel and Pd boundary, and (3) darker and lighter regions throughout the sample surface. To analyze diffusion behavior, FIB was used to prepare lamellae from five different areas of the UCO-Pd sample, as well as one for the as-received sample. These lamellae were examined using a Scanning Transmission Electron Microscope ThermoFisher Spectra 300 (STEM-Spectra). Energy Dispersive Spectroscopy results confirmed the diffusion between the UCO fuel and Pd bar. Figure 2 shows the identified phases of UC, UO2, and Pd dispersed throughout the interaction region. It was also found that the Pd concentration decreases as the radial distance from the UCO-Pd interface increases. The Pd concentration was 36.93% atomic fraction 11µm from the interaction region then decreased to 7.51% atomic fraction at 257 µm away from the interaction region. Figure 3 shows how concentrations of U and Pd vary across the interaction zone, highlighting the extent of diffusion. This study confirms that Pd interacts with surrounding material to form U-Pd phase and diffusion zones. These diffusion zones vary in composition depending on its radial distance from the point of contact with the Pd bar. These findings contribute to a deeper understanding of the fission product behavior in high temperature nuclear fuels, aiding in the prediction and mitigation of potential fuel degradation mechanisms. A B C Fig. 1. SEM micrographs of UCO fuel kernel in contact with solid Pd bar and the lift out locations. Figure 1A shows UCO fuel kernel in contact with solid Pd bar. Figure 1B shows a higher magnification image of the UCO fuel-Pd interaction zone. Figure 1C shows lift out sites for the lamellae. A B C Fig 2. EDS maps reveal the distribution of U, O, and Pd in location 3. Figure 2A shows the distribution of uranium. Figure 2B shows the distribution of oxygen. Figure 2C shows the distribution of palladium. A B C Fig 3. Palladium and uranium concentration across interaction zone in location 2. Figure 3A shows EDS map of lift out number 2. Figure 3B shows a zoomed in EDS map taken from the interaction zone. Figure 3C shows a line graph of uranium and palladium concentrations across the interaction zone. References: 1. B.E. Wells, N.R. Phillips, K.J. Geelhood. Pacific West Laboratory. (2021). TRISO Fuel: Properties and Failure Modes. https://www.nrc.gov/docs/ML2117/ML21175A152.pdf (Accessed January 16, 2025). 2. American Nuclear Society. TRISO Fuel Development Progresses in INL, ORNL. https://www.gen-4.org/gif/upload/docs/application/pdf/2014-03/nov13nn_fuel_reprint.pdf (Accessed January 16, 2025) 3. Clark R.A., M.A. Con

36 - MATERIALS SCIENCE↗

DOE OSTI2025

Accurate determination of uranium isotope abundances by wavelength modulation spectroscopy in atomic beams

The design and demonstration of an optical analysis system based on wavelength modulation spectroscopy in an atomic beam for uranium isotope abundance determinations is presented. This system probes the uranium 5f 3 6d7s 2 ( 5 L 6 ) → 5f 2 6d 2 7s 2 ( 5 K 5 ) transition at 861.031 nm, which is considered to be the most suitable transition for uranium isotopic analysis. A new laser characterization strategy was developed for the conditions where optimum laser wavelength modulation depth was small compared to the free spectral range (FSR) of etalons. Two capabilities enabled the higher-precision determination of isotope abundances of atomic beams: (1) reduction of low-frequency additive noise, especially the noise caused by black-body radiation and (2) suppression of non-absorption transmission losses. The performance of this system was validated with uranium samples of various isotopic compositions. Further, by comparing the measurements using natural uranium samples between the direct absorption and the wavelength modulation approaches, a 21-fold decrease in uncertainty of the integrated absorbance and a 6.8-fold improvement in the 1-σ precision of the number density were achieved. In addition, by comparing the results using uranium oxide samples, a 6.1-fold decrease in the uncertainty of inferred isotope abundance was obtained. These results demonstrate that the 1f-normalized 2f wavelength modulation spectroscopy (WMS-2f/1f) technique enables higher-precision analysis of atomic beams.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

DOE OSTI2026

Fingerprinting Uranium Oxides with Electron Energy Loss Spectroscopy Supported by Theoretical Computations

Uranium oxides occur in a variety of phases that differ in crystal structure and uranium oxidation states. Electron energy loss spectroscopy (EELS) is one of the few techniques that has sufficient spatial resolution and sensitivity to electronic structure to distinguish amongst phases at the nanoscale. However, beam-sensitive materials such as uranium oxides are subject to spectral modification due to interactions with the electron beam. Therefore, theory support is essential to reliably exclude the impact of beam damage and generate true reference datasets. Here we use a comparison of theoretical and experimental spectra to probe the impact of beam damage on O K-edge and U N-edge (N6,7 and N4,5) EELS spectra of various single-valent and mixed-valence uranium oxide bulk phases. Using a low-dose experimental set-up, we show that the O K-edge theoretical spectra are in excellent agreement with experiment for both peak positions and relative intensities of respective peaks. In contrast, U N-edge features are less distinguishing due to the partially localized nature of the U 5f orbitals and overlapping multiplet and spin–orbit coupling effects. This work demonstrates that O K-edge EELS is sufficiently diagnostic to distinguish a wide range of uranium oxides and that the experimental approach used here minimizes beam damage and allows valence state discrimination across the U(IV), U(V) and U(VI) series. When combined with imaging modes available in electron mi-croscopy, the work enables detailed investigation and characterization of uranium redox transformations at the nanoscale.

Carbone, Jacopo↗

DOE OSTI2024

p K a of alcohols dictates their reactivity with reduced uranium-substituted thiomolybdate clusters

The uranium-substituted thiomolybdate cluster, (Cp* 3 Mo 3 S 4 )UCp*, has been demonstrated as a model for water reduction by single uranium atoms supported on a molybdenum sulfide surface (U@MoS 2 ). In this study, the scope of O–H bond activation is expanded through the investigation of the reactivity of various alcohols with differing pKa values for the –OH proton. The reaction of (Cp* 3 Mo 3 S 4 )UCp* with stoichiometric amounts of methanol, phenol, 2,6-dichlorophenol, and nonafluoro-tert-butyl alcohol affords the corresponding mono-alkoxide species, (Cp* 3 Mo 3 S 4 )Cp*U(OR), via a uranium-metalloligand cooperative activation of the O–H bond. This observed reactivity is analogous to the O–H bond activation reported by (Cp* 3 Mo 3 S 4 )UCp* in the presence of water. However, addition of tert-butanol induces protonolysis of the Cp* ligand on uranium, resulting in the formation of a uranium tris-tert-butoxide cluster, (Cp* 3 Mo 3 S 4 )U(O t Bu) 3 . Independent synthesis of (Cp* 3 Mo 3 S 4 )Cp*U(O t Bu) was possible via an alternative pathway, eliminating sterics as a justification for the observed discrepancy in reactivity. Furthermore, these results offer insight into the role the –OH proton pKa plays in dictating the mechanism of O–H bond activation of alcohols by the uranium-substituted thiomolybdate cluster.

Patra, Kamaless↗

DOE OSTI2025

Multi-Objective Optimization of Uranium Target Assembly–3: A Comparison of Genetic and Traditional Methods

Commonly produced as a byproduct of uranium fission, 99 Mo is a key medical isotope that is in high demand in the United States. An international goal is to switch from medical isotope production technologies that require highly enriched uranium to medical isotope production technologies that require only low-enriched uranium. Niowave Inc. is contributing to this goal by developing an accelerator-driven subcritical assembly called the Uranium Target Assembly (UTA). This work compares the performance of Dakota’s Multi-Objective Genetic Algorithm (MOGA) against traditional sensitivity analysis in the neutronic optimization of the UTA-3 system. The design objectives are k-eigenvalue (k eff ) and natural uranium fission power, which are directly correlated with the amount of 99 Mo produced. Dakota:MOGA did not perform as well as human engineering ingenuity in optimization studies with high numbers of input parameters, such as fuel rod type selection and fuel rod placement. However, Dakota:MOGA did outperform traditional sensitivity analysis in optimization studies with fewer than 20 parameters and revealed the degree to which each parameter influences the optimal design space for k eff and natural uranium fission power (to a lesser extent). As the design model became more complex in the final stage of design, the computational resources required to calculate the design objective values in the Monte Carlo N-Particle transport code from selected input parameter combinations limited Dakota:MOGA’s performance, and, unfortunately, human intervention was required to discern the optimal design space. In conclusion, future work will attempt to reduce computational resource constraints by incorporating areduced-order neutronics model into the optimization cycle.

Accelerator-driven systems↗

NASA NTRS1966

Stabilization of Tungsten-Uranium Dioxide Composites Under Thermal Cycling Conditions

Uranium losses during thermal cycling of tungsten - uranium dioxide composites to 2500° C in flowing purified hydrogen for specimens initially containing 35 volume percent uranium dioxide were reduced by the use of thorium dioxide, calcium oxide, or yttrium oxide as additives in solid solution with the uranium dioxide. The effectiveness of the additives decreased in the order yttrium oxide, calcium oxide, and thorium dioxide. Stabilization of uranium dioxide by calcium oxide or yttrium oxide can be explained in part on the basis of the introduction of oxygen vacancies into the fluorite lattice and the associated lowered partial molar free energy of oxygen. The difference in the effectiveness of calcium oxide and yttrium oxide is discussed, and a possible explanation of the observed behavior is proposed. Of the concentrations of yttrium oxide tested (i.e., 2. 5, 5, and 10 mole percent), 10 mole percent of yttrium oxide was the most effective in pre venting loss of uranium from composities.

Gluyas, Richard E.↗

NASA NTRS1973

Uranium concentration and distribution in six peridotite inclusions of probable mantle origin

Fission-track activation was used to investigate uranium concentration and distribution in peridotite inclusions in alkali basalt from six localities. Whole-rock uranium concentrations range from 24 to 82 ng/g. Most of the uranium is uniformly distributed in the major silicate phases - olivine, orthopyroxene, and clinopyroxene. Chromian spinels may be classified into two groups on the basis of their uranium content - those which have less than 10 ng/g and those which have 100 to 150 ng/g U. In one sample accessory hydrous phases, phlogopite and hornblende, contain 130 and 300 ng/g U, respectively. The contact between the inclusion and the host basalt is usually quite sharp. Glassy or microcrystalline veinlets found in some samples contain more than 1 microgram/g. Very little uranium is associated with microcrystals of apatite. These results agree with some earlier investigators, who have concluded that suboceanic peridotites contain too little uranium to account for normal oceanic heat flow by conduction alone.

Haines, E. L.↗

NASA NTRS1976

Laboratory-scale uranium RF plasma confinement experiments

An experimental investigation was conducted using 80 kW and 1.2 MW RF induction heater facilities to aid in developing the technology necessary for designing a self-critical fissioning uranium plasma core reactor. Pure uranium hexafluoride (UF6) was injected into argon-confined, steady-state, RF-heated plasmas in different uranium plasma confinement tests to investigate the characteristics of plamas core nuclear reactors. The objectives were: (1) to confine as high a density of uranium vapor as possible within the plasma while simultaneously minimizing the uranium compound wall deposition; (2) to develop and test materials and handling techniques suitable for use with high-temperature, high-pressure gaseous UF6; and (3) to develop complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma and residue deposited on the test chamber components. In all tests, the plasma was a fluid-mechanically-confined vortex-type contained within a fused-silica cylindrical test chamber. The test chamber peripheral wall was 5.7 cm ID by 10 cm long.

Roman, W. C.↗

NASA NTRS1976

Measurements of uranium mass confined in high density plasmas

An X-ray absorption method for measuring the amount of uranium confined in high density, rf-heated uranium plasmas is described. A comparison of measured absorption of 8 keV X-rays with absorption calculated using Beer Law indicated that the method could be used to measure uranium densities from 3 times 10 to the 16th power atoms/cu cm to 5 times 10 to the 18th power atoms/cu cm. Tests were conducted to measure the density of uranium in an rf-heated argon plasma with UF6 infection and with the power to maintain the discharge supplied by a 1.2 MW rf induction heater facility. The uranium density was measured as the flow rate through the test chamber was varied. A maximum uranium density of 3.85 times 10 to the 17th power atoms/cu cm was measured.

Stoeffler, R. C.↗

DOE OSTI2025

Preliminary Insights Into the Feasibility of Determining the Purification Date of Enriched Uranium by Direct Measurement of the 230 Th/ 234 U Ratio Using an All-Faraday Detector Configuration on the Neoma MC-ICP-MS

Rationale: Mass spectrometric measurement of the 230 Th/ 234 U ratio to calculate the purification age of enriched uranium is typically conducted via a combination of ion counters and faraday detectors, thus requiring an inter-detector calibration scheme. Here, our aim is to understand whether the pursuit of a simplified measurement scheme involving only faraday detectors is feasible. Methods: We investigate the possibility of determining U-Th model ages for two enriched uranium standards (NBL U630 and U850) by direct measurement of the 230 Th/ 234 U ratio (without chromatographic separation or isotope dilution) on a ThermoFisher Scientific Neoma MC-ICP-MS utilizing both solution and laser ablation (LA)-based sampling techniques and an all-faraday detector configuration. Results: For the solution mode analyses conducted on aliquots containing sub μg/mL total U, we produce composite average 230 Th/ 234 U model dates of May 19, 1988 (± 351 days), and March 26, 1961 (± 2.5 years) using the directly measured 230 Th/ 234 U ratios for the NBL U630 and U850 uranium standards, which have certified purification dates of June 6, 1988 (± 190 days), and December 31, 1957 (± 36.5 days), respectively. The ages produced by LA-based sampling of dried residues of the same standards deposited onto cotton TexWipes are less accurate and of poorer precision (June 23, 2004 ± 8.7 years for U630 and December 21, 1965 ± 7.9 years for U850) but still yield meaningful information in regards to the purification date. Conclusions: We believe that further refinement of the all faraday detector measurement approach to include development of a more robust Th/U relative sensitivity factor determination, signal cutoff selection, and data processing protocols will allow for this approach to be confidently applied to enriched uranium materials with unknown purification histories. Potential advantages of the method include the reduced sample handling and infrastructure requirements as well as the ability to simultaneously generate a broad picture of the uranium isotopic composition in tandem with the U-Th age determination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗