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DOE OSTI2018

Quantitative Prediction of Uranium Speciation and Amidoxime Binding in Seawater from Advanced Simulation Techniques

The goal of this project was to quantitatively predict stability of uranium species and thermochemistry of uranium binding with amidoxime ligands in realistic seawater conditions. The specific aims of the project were: (i) determine the stable species of uranium and its solvation structure in realistic seawater conditions; also speciation and solvation of uranium's major competing species in seawater, vanadium; (ii) accurately predict the free energies (and thereby stability constants) of binding between uranium and amidoxime; (iii) accurately predict the reaction enthalpies and entropies of binding between uranium and amidoxime as well as between vanadium and amidoxime. Main results and conclusions are as follows. First-principles molecular dynamic simulations found that the structure of the Ca 2 UO 2 (CO 3 ) 3 complex is very stable and that one Ca ion binds to the center UO 2 (CO 3 ) 3 4- anion stronger than the other Ca ion. This finding suggests that using time-resolved EXAFS spectra may confirm the asymmetry in binding of the two Ca ions in the aqueous Ca 2 UO 2 (CO 3 ) 3 complex. To consider the common ions, solvation of the Ca 2 UO 2 (CO 3 ) 3 complex in seawater was simulated by classical molecular dynamics simulation. It was found that the structure of the Ca 2 UO 2 (CO 3 ) 3 complex is very stable in the model seawater. A Na+ ion was found to be closely associated with the Ca 2 UO 2 (CO 3 ) 3 complex by indirectly interacting with one axial oxygen atom of the UO 2 group bridged by a water molecule. In addition, the Na + ion interacts closely with one Ca 2+ ion than the other. The present simulations revealed the key role of common ions such as Na + in impacting the solvation, structure, and apparent charge of the Ca 2 UO 2 (CO 3 ) 3 complex in seawater. The sequential processes of UO 2 2+ binding with the three CO 3 2- groups and the two Ca 2+ ions in pure water and in 0.1 M NaCl were simulated by classical molecular dynamics with both the non-constrained brute-force approach and umbrella sampling. The simulated free energies show excellent agreement with the experiment. The coordination structures of UO 2 (CO 3 ) 3 4- and CaUO 2 (CO 3 ) 3 2- were found to be significantly affected by the presence of Na + ions, leading to a monodentate binding of a carbonate group to U and a water molecule entering the first coordination shell of U. To shed light on the binding between uranyl and glutardiamidoxime (H 2 B), an important model and state-of-the-art ligand for seawater uranium extraction, a suite of computational methods was employed. From molecular dynamics (MD) simulations, it was found that the binding configurations of B 2- with uranyl favor the twofold distorted η 2 binding between the oximate ends (C=N-O-) and U, while HB - prefers the chelating mode for the oximate end with the neutral end being solvated by water. The free energies of sequential ligand binding to form UO 2 B, [UO 2 B 2 ] 2- , and [UO 2 (HB)B] - were simulated with umbrella sampling and very good agreement with the experimental values was achieved, which corroborates the structural insights into the binding mode. State-of-the-art polymeric sorbents employ both amidoximate and carboxylate groups on the side chains to achieve optimal U uptake and selectivity, so we simulated the binding of a model amidoximate–carboxylate bifunctional ligand with uranyl. Classical MD and free-energy simulations in 0.5 M NaCl showed that the carboxylate group binds first to uranyl, leading to a loose intermediate state, and then, the amidoximate group binds, resulting in a more stable and tight chelate state. Binding of the second bifunctional ligand follows a similar process, and the two ligands prefer a trans configuration around the uranyl group. The simulated free energies indicate that the two bifunctional ligands bind with uranyl 14 kcal/mol stronger than the two ligands with only amidoximate groups, confirming an important synergy between amidoximate and carboxylate groups in binding uranyl.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

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

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 OSTI2017

Recovery of uranium from an irradiated solid target after removal of molybdenum-99 produced from the irradiated target

A process for minimizing waste and maximizing utilization of uranium involves recovering uranium from an irradiated solid target after separating the medical isotope product, molybdenum-99, produced from the irradiated target. The process includes irradiating a solid target comprising uranium to produce fission products comprising molybdenum-99, and thereafter dissolving the target and conditioning the solution to prepare an aqueous nitric acid solution containing irradiated uranium. The acidic solution is then contacted with a solid sorbent whereby molybdenum-99 remains adsorbed to the sorbent for subsequent recovery. The uranium passes through the sorbent. The concentrations of acid and uranium are then adjusted to concentrations suitable for crystallization of uranyl nitrate hydrates. After inducing the crystallization, the uranyl nitrate hydrates are separated from a supernatant. The process results in the purification of uranyl nitrate hydrates from fission products and other contaminants. The uranium is therefore available for reuse, storage, or disposal.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Integration of Real-Time Geophysical Monitoring with In Situ Uranium Remedy Implementation

An in situ injection of polyphosphate amendment was conducted to address a uranium source zone for a plume near the Columbia River at the U.S. Department of Energy Hanford Site. This treatment was applied as an enhanced attenuation approach, targeting uranium in a periodically wetted zone where groundwater inundates as the water table rises in response to the seasonal high river stage. Uranium within this zone is mobilized into the groundwater at the high river stage, resulting in a persistent groundwater plume. The polyphosphate amendment creates precipitates that decrease uranium mobility so that the amount of uranium flushed into groundwater when the zone is periodically wetted is significantly decreased. With a reduced flux of uranium into the plume, the plume concentrations are expected to decrease over time. During injection, real-time geophysical monitoring was applied to provide hourly images of injected-fluid distribution. The real-time autonomous data collection, processing, and reporting enabled the electrical resistivity tomography images to provide operational feedback during injection. These data were used in conjunction with operational information (e.g., flow rate, pressures) and subsurface data from 24 monitoring wells placed within and downgradient of the treatment area (e.g., water level and specific conductance probes, uranium concentration, phosphate concentration) to assess overall performance of the treatment.

Johnson, Timothy C.↗

DOE OSTI2021

Single-Use Destructive Assay for Uranium Hexafluoride Sampling

Sampling uranium hexafluoride (UF6) for the determination of enrichments by destructive analysis (DA) is a critical component in the International Atomic Energy Agency’s layered safeguards approach for uranium processing facilities. Typically, gram-quantity UF6 samples are collected during inspections and stored under tag-and-seal until transportation to an off-site analytical laboratory. The shipping times can be long, and evolving restrictions on radioactive/corrosive materials shipments may increasingly limit the IAEA’s ability to transport UF6 samples easily. Pacific Northwest National Laboratory has developed a low-cost UF6 sampling technology called Single-Use Destructive Assay (SUDA) that addresses these challenges, as well as provides DA sample geometries that can be tailored for different analytical methods, including potential on-site analyses. The SUDA samplers, along with a unique holder, are designed for direct attachment to existing taps at uranium processing facilities, allowing gaseous UF6 to come into direct contact with a zeolite film. The SUDA technology features the ability to capture uranium in a more easily shipped and handled form as the solid, more stable, and relatively less hazardous hydrated uranyl fluoride (UO2F2•nH2O), which is formed through the controlled hydrolysis of UF6. We have recently simulated uranium collection under enrichment plant sampling conditions to further improve our understanding of SUDA sampling. Presented here is our recent work on measuring the relationship between sampling conditions and uranium collection, which includes control of the uranium-mass-to-zeolite ratio and assessing variable UF6 gas and sampling parameters that can affect collection using the SUDA sampler.

Pope, Timothy R.↗

DOE OSTI2017

Promising Technology for Reprocessing Complex Uranium Ores from the Elkon Deposit

The results of research on reprocessing of complex uranium ores from the Elkon deposit with recovery of uranium, gold, silver, and vanadium by methods of radiometric sorting, combining of autoclave and atmospheric leaching of ore followed by sorption and extraction recovery of valuable components, are presented. The capacity was 90 mg/g in sorption of uranium on strongly basic anionite and the uranium concentration ≤0.003 g/dm{sup 3} in sorption filtrates. The choice of eluent is determined by the formulation of the problem of obtaining the finished uranium product of the requisite form and quality. A fundamentally new composition of the extracting agent is proposed for direct extraction of uranium from nitrate-containing solutions. An iron-vanadium concentrate with vanadium and iron content 2.58 and 22.8%, respectively, suitable for subsequent reprocessing, was obtained. The anionite PuroGold S-992 is recommended for recovery of gold and silver. The gold and silver concentrations in the commercial desorbate were 270 and 380 mg/dm{sup 3}, respectively; a gold-containing concentrate was obtained.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Solvothermal synthesis of uranium(VI) phases with aromatic carboxylate ligands: A dinuclear complex with 4-hydroxybenzoic acid and a 3D framework with terephthalic acid

With the coordination of dimethylformamide (DMF), two new uranium(VI) complexes with either 4-hydroxybenzoic acid (H{sub 2}phb) or terephthalic acid (H{sub 2}tph) have been synthesized under solvothermal conditions and structurally characterized. [(UO{sub 2}){sub 2}(Hphb){sub 2}(phb)(DMF)(H{sub 2}O){sub 3}]·4H{sub 2}O (1) has a dinuclear structure constructed with both pentagonal and hexagonal bipyramidal uranium polyhedra linked through a µ{sub 2}-bridging ligand via both chelating carboxylate arm and alcohol oxygen bonding, first observation of such a coordination mode of 4-hydroxybenzoate for 5 f ions. [(UO{sub 2})(tph)(DMF)] (2) has a three-dimensional (3D) framework built with pentagonal bipyramidal uranium polyhedra linked with µ{sub 4}-terephthalate ligands. The 3D channeled structure is facilitated by the unique carboxylate bonding with nearly linear C–O–U angles and the coordination of DMF molecules. The presence of phb ligands in different coordination modes, uranyl ions in diverse environments and DMF in complex 1, and tph ligand, DMF and uranyl ion in complex 2 has been confirmed by Raman spectroscopy. In addition, their thermal stability and photoluminescence properties have been investigated. - Graphical abstract: With the coordination of dimethylformamide, two new uranyl complexes with either 4-hydroxybenzoate or terephthalate have been synthesized under solvothermal conditions and structurally characterized. - Highlights: • Solvent facilitates the synthesis of two new uranium(VI) complexes. • A dinuclear complex with both penta- and hexagonal bipyramidal uranium polyhedral. • A unique µ{sub 2}-bridging mode of 4-hydroxybenzoate via alcohol oxygen for 5 f ions. • A 3D framework with uranium polyhedra and µ{sub 4}-terephthalate ligands. • Vibration modes and photoluminescence properties are reported.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2023

Assessing the influence of microstructure on uranium hydride size distributions via small angle neutron scattering

Here, the effect of microstructure on the internal hydriding behavior of both cast (1 mm grain size) and rolled (25 μm grain size) uranium containing hydrogen concentrations between 0 and 1.8 wppm were evaluated via small angle neutron scattering (SANS). Increasing hydrogen content up to 1.8 wppm in the cast uranium only weakly affected the average uranium hydride (UH 3 ) precipitate size, calculated from the SANS data. Conversely, the UH 3 phase fraction was found to strongly depend on the hydrogen content in the same cast samples. A substantially reduced UH 3 particle size distribution was observed in the rolled uranium relative to cast uranium containing the same nominal hydrogen content. It is hypothesized that the suppression of UH 3 formation in the rolled uranium is driven by increased hydrogen trapping at grain boundaries, and theoretical calculations that account for trap density, potency, and hydrogen diffusion kinetics support this hypothesis.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Investigating the Effect of Sorbed Humic Acid on the Mobility of Uranium - 19461

The Savannah River Site (SRS) was one of the major nuclear facilities for the U.S Department of Energy. During the Cold War, these facilities produced large amounts of radioactive and hazardous waste. On site, the F-Area consisted of three unlined seepage basins that collected approximately 6.8 billion liters of low-level waste that contained nitric acid, radionuclides, and dissolved metals due to plutonium separation operations. Sufficient quantities of uranium isotopes, I-129, Tc-99, and tritium migrated into the groundwater, creating an acidic plume with a pH between 3 and 5.5. Although the basins functioned as designed, the acidic nature of the waste created a source of contamination for the groundwater and soil below the basins. With radionuclides above their maximum contaminant level (MCL) remediation efforts took place. Humic acid as a remediation technology has shown potential for controlling radionuclide mobility. Humic acid is a fraction of humic substance that are ubiquitous in the environment. Humic substances (HS) are complex heterogeneous mixtures of polydispersed materials formed by biochemical and chemical reactions during the decay and transformation of plant and microbial remains. Humic substances account for 50-80% of the organic carbon in the soil or sediment and are known for their excellent binding capacity for metals. Humic acid can be extracted with a potassium hydroxide base solution and then drying it. The high pH solubilizes the humic acid molecules and generates a dark-brown highly-concentrated solution, rich in humic acid, which can be diluted for use. Humic acid's strong sorption capacity and its ability to develop a bond with uranium at slightly acid pH constitute it as remediation technology to reduce uranium (VI) mobility. Column experiments were performed to identify the sorption and desorption of HS in flow-through porous media representative of the acidic aquifer at the F-Area Seepage Basins. Two commercially available HA were selected for comparison, Huma-K and KW15 modified humics (modified humic acid, mod-HA). By conducting a nonreactive rhenium tracer test, on columns filled with soil obtained from SRS's FAW-1 at a depth interval of 18 m -21 m, column parameters such as pore volume (PV), variance, dispersion and peclet number were determined. Columns were then used to estimate the influence of HA on uranium mobility. Data obtained from this study shows that using HA as an amendment for uranium remediation improves uranium removal significantly. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Uranium Mill Tailings Radiation Control Act Title II DOE Due Diligence and Lessons Learned from a Previous Site Transfer - 20352

Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) established that a government agency will provide perpetual care for closed uranium and thorium ore-processing sites that were operating under an NRC source material license in 1978 or were licensed thereafter. Commercial owners (licensees) operating under an NRC or agreement state specific license when UMTRCA was passed are responsible for conducting reclamation of any byproduct material remaining from uranium-ore processing operations in accordance with an NRC or agreement state approved reclamation plan. Reclamation includes both surface and groundwater remedies. Upon completion of reclamation and approval by NRC, the site is required to be transferred to either the host state or the DOE for long-term surveillance and maintenance. Since UMTRCA's enactment, six Title II sites have been transferred to DOE; an additional 24 Title II sites are anticipated to be transferred before 2050. DoE's role mandated under UMTRCA Title II as the long-term care custodian is to perform 'monitoring, maintenance, and emergency measures necessary to protect the public health and safety.' UMTRCA requires that the licensee pay a long-term surveillance charge 'sufficient to cover the annual costs of site surveillance.' However, at some sites such as the Bluewater, New Mexico, Disposal Site, this mandate has required additional effort and expense by DOE, beyond the originally anticipated and intended scope within UMTRCA, but within the authority of DOE under UMTRCA. In 1997, the Bluewater site became the second UMTRCA Title II site to be transferred to DOE. The site was the location of a uranium mill operated from 1953 until 1982. The specific licensee began site reclamation in 1991, and by 1995 all tailings and contaminated materials were encapsulated in two tailings disposal cells and other disposal areas. In addition to surface contamination, milling activities impacted groundwater in the two upper aquifers. In 1989, the specific licensee attempted active groundwater remediation; however, no significant reduction in contaminant concentrations was observed. As a result, the specific licensee applied to NRC for alternate concentration limits (ACLs) in 1990, which were approved in 1996 as being protective, after additional corrective actions were performed. Since transfer of the Bluewater site to DOE, unforeseen challenges have occurred, requiring additional actions. The first challenge is the occurrence of surface depressions located on the northern section of the main tailings disposal cell. The depressions were first observed during DoE's initial inspection in 1998; however, evidence of these can be observed on satellite images taken prior to transfer. Since being first observed, the depressions have continued to grow both in depth and areal extent. Due to the design of the main tailings disposal cell, the depressions impede storm water from being effectively shed off the 101- hectare (250-acre) top slope of the main tailings disposal cell. Instead, storm water accumulates in the depressions, forming a large ephemeral pond that has stored up to 16.3 x 10{sup 6} liters (4.3 million gallons) of stormwater. The ponding poses a potential risk to the integrity of the main tailings disposal cell in the case of a large storm event, with the potential to cause the pond to overtop and erode the cover material and underlying waste. DOE has taken a number of short-term actions to monitor, measure, and reduce the ponding and is currently working with the US Army Corps of Engineers to design and construct a repair. Additional challenges are associated with groundwater at the Bluewater site. Nine wells were present on the 1335-hectare (3300-acre) site upon transfer. Groundwater compliance was called into question after the State of New Mexico reduced its uranium groundwater standard from 5.0 to 0.03 milligrams per liter in 2004, and when an ACL for uranium was exceeded in a site monitoring well in 2010. Acquiring historical groundwater data and subsequent evaluations as well as additional DOE groundwater monitoring led to installing 10 new monitoring wells and performing additional site hydrogeology recharacterization. DOE continues to evaluate groundwater conditions at the site and works with the NRC to determine regulatory requirements and a path forward. As a result of lessons learned at the Bluewater site and other Title II sites, improved processes have been implemented at a programmatic level to increase due diligence before site transfer and prevent similar issues from occurring at other UMTRCA Title II sites under long-term management. DoE's due diligence process is documented in the Process for Transition of UMTRCA Title II Disposal Sites to DOE for Long-Term Surveillance and Maintenance and is designed to ensure that DOE has no technical or compliance concerns with regulatory decisions that might compromise protectiveness following site transfer to DOE. Implementation of the due diligence process has increased DoE's role prior to site transfer and has been effective in identifying potential issues. Actions by NRC and specific licensees, in response to enhanced due diligence efforts by DOE, are expected to minimize, if not totally prevent, the need for unanticipated actions by DOE pertaining to the surface and groundwater remedies after site transfer. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

NGSI FY15 Final Report. Innovative Sample Preparation for in-Field Uranium Isotopic Determinations

Our FY14 Final Report included an introduction to the project, background, literature search of uranium dissolution methods, assessment of commercial off the shelf (COTS) automated sample preparation systems, as well as data and results for dissolution of bulk quantities of uranium oxides, and dissolution of uranium oxides from swipe filter materials using ammonium bifluoride (ABF). Also, discussed were reaction studies of solid ABF with uranium oxide that provided a basis for determining the ABF/uranium oxide dissolution mechanism. This report details the final experiments for optimizing dissolution of U 3 O 8 and UO 2 using ABF and steps leading to development of a Standard Operating Procedure (SOP) for dissolution of uranium oxides on swipe filters.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2018

Uranium exchange kinetics in a molten LiCl-KCl/Cd system at 500 °C

A dissolution experiment is performed in order to examine the kinetics of uranium isotope equilibration in a molten salt and cadmium system. The salt contains a solute concentration of UCl 3 in LiCl-KCl eutectic and the cadmium contained dissolved uranium at 500 °C. A small piece of high enriched uranium is introduced to the cadmium, which perturbs the equilibrium with respect to the uranium enrichment of the salt and cadmium phases. The series of salt and cadmium samples trace how the enrichment of the two phases reach equilibrium (in the absence of a chemical driving force) over the course of approximately 24 h. Here, a model linking dissolution and isotope exchange kinetics is developed to track the evolution of uranium isotope composition. With the assumption of unaccounted salt mass incurred from the dated sample reporting practice, the model explains the trend of the measured uranium isotope composition well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

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 OSTI2024

Laser-Induced Thermal Decomposition of Uranium Coordination Compounds with Non-oxidic Ligands to Produce Nitride and Carbide Materials

The production of ceramics from uranium coordination compounds can be achieved through thermal processing if an excess amount of the desired atoms (i.e., C or N), or reactive gaseous products (e.g., methane or nitrogen oxide) is made available to the reactive uranium metal core via decomposition/fragmentation of the surrounding ligand groups. Here, computational thermodynamic approaches were utilized to identify the temperatures necessary to produce uranium metal from some starting compounds—UI 4 (TMEDA) 2 , UCl 4 (TMEDA) 2 , UCl 3 (pyridine) x , and UI 3 (pyridine) 4 . Experimentally, precursors were irradiated by a laser under various gaseous environments (argon, nitrogen, and methane) creating extreme reaction conditions (i.e., fast heating, high temperature profile >2000 °C, and rapid cooling). Despite the fast dynamics associated with laser irradiation, the central uranium atom reacted with the thermal decomposition products of the ligands yielding uranium ceramics. Residual gas analysis identified vaporized products from the laser irradiation, and the final ceramic products were characterized by powder X-ray diffraction. The composition of the uranium precursor as well as the gaseous environment had a direct impact on the production of the final phases.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

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↗