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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Synthesis and characterization of a low-valent uranium complex supported by a redox-active molybdenum oxide metalloligand

Investigation of the reactivity of the redox-active molybdenum oxide cluster, Cp$^{*}_{2}$Mo 2 O 4 , toward low-valent uranium (Cp$^{*}_{2}$U(bpy), bpy = 2,2'-bupyridine) is reported. Here, structural and spectroscopic analyses provide insight into the interactions of the actinide and metalloligand in the product, [(Cp$^{*}_{2}$Mo 2 O 4 )Cp$^{*}_{2}$U]. Reactivity studies reveal the metalloligand functions as a redox reservoir for uranium.

Valerio, Leyla R. [University of Rochester, NY (Un↗
DOE OSTI2020

Materials Data on Rb6U2Mo4O21 by Materials Project

Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na6U2Mo4O21 by Materials Project

Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na2U(MoO5)2 by Materials Project

Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Rb6U(Mo2O9)2 by Materials Project

Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na6U(Mo2O9)2 by Materials Project

Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

36 MATERIALS SCIENCE↗
DOE OSTI2020

Materials Data on Na4U(MoO4)4 by Materials Project

Computed materials data using density functional theory calculations. These calculations determine the electronic structure of bulk materials by solving approximations to the Schrodinger equation. For more information, see https://materialsproject.org/docs/calculations

36 MATERIALS SCIENCE↗
DOE OSTI2025

Recovering high-purity uranyl nitrate from simulated used nuclear fuel dissolver solutions by crystallization: rejecting technetium

The separation of U from Tc and other problematic fission product elements like Mo and Ru, along with Sr, Zr, Cs, and Nd, has been achieved via the crystallization of uranyl nitrate hexahydrate (UNH). Rejection of technetium as pertechnetate anion ( 99 TcO 4 – ) is an especially important feature of this system, as it otherwise tends to follow U (VI) in extractive separations. It also raises the salient question regarding why this oxoanion cannot replace nitrate within the crystalline lattice of UNH. Results showed high-yield (>90 %), high-purity (>99 %) recovery of U as UNH from solutions containing 99 TcO 4 – by simple reduction of temperature from 60°C to 20°C. There was no observable interaction of 99 TcO 4 – with UO 2 2+ . The addition of other cations like, Sr 2+ , Zr 4+ , Cs + , and Nd 3+ , also did not form secondary, contaminant solid phases, leaving the > 99 % of the fission product elements in the mother liquor, while the U was recovered at > 90 %. Similarly, Mo and Ru, when added to the mixture, were shown to behave as the other fission-product elements, remaining in the mother liquor during crystallization. As a result, DFT calculations showed that, despite the higher binding strength of TcO 4 – , HMoO 4 – , and BiO 3 – with the UO 2 2+ cation compared to NO 3 – , the hydrogen-bonding network of the two coordinated ions and four waters of hydration in the UNH crystal structure is the driving force for the high specificity of this separation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Recovery and Recycle of Irradiated Low-Enriched Uranium from the Production of 99 Mo

Technetium-99m ( 99m Tc), the daughter (decay product) of molybdenum-99 ( 99 Mo), is the most commonly used medical radioisotope in the world. 99 Mo is produced primarily from the fissioning of 235 U. A potential future producer, Eden Radioisotopes, is planning to irradiate low-enriched-uranium metal-foil targets for production of 99 Mo. Their plan is to recover, purify, and recycle the irradiated LEU in a new set of targets inside their facility. This study assesses processes to (1) purify and recycle uranium following 99 Mo recovery and (2) convert the uranium to metal for subsequent foil production and target fabrication. A UREX (uranium recovery by extraction) liquid–liquid extraction flowsheet that utilizes centrifugal contactors was developed using the Argonne model for universal solvent extraction (AMUSE) to recover and purify the uranium. The calculated flowsheet predicted high decontamination from Pu and most fission products and >99.99% recovery of uranium. Suggestions for conversion of the UREX U-product (uranyl nitrate in dilute nitric acid) to U metal were provided based on literature studies and earlier laboratory studies performed at Argonne National Laboratory.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Covalent bond shortening and distortion induced by pressurization of thorium, uranium, and neptunium tetrakis aryloxides

Covalency involving the 5f orbitals is regularly invoked to explain the reactivity, structure and spectroscopic properties of the actinides, but the ionic versus covalent nature of metal-ligand bonding in actinide complexes remains controversial. The tetrakis 2,6-di- tert -butylphenoxide complexes of Th, U and Np form an isostructural series of crystal structures containing approximately tetrahedral MO 4 cores. We show that up to 3 GPa the Th and U crystal structures show negative linear compressibility as the OMO angles distort. At 3 GPa the angles snap back to their original values, reverting to a tetrahedral geometry with an abrupt shortening of the M-O distances by up to 0.1 Å. The Np complex shows similar but smaller effects, transforming above 2.4 GPa. Electronic structure calculations associate the M-O bond shortening with a change in covalency resulting from increased contributions to the M-O bonding by the metal 6d and 5f orbitals, the combination promoting MO 4 flexibility at little cost in energy.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Mo-99 Concentration and processing by Solvent Extraction and Ion Exchange

Normally derived from its transient-equilibrium parent molybdenum-99 ( 99 Mo) (t½ = 66 h), the short-lived isotope technetium-99m ( 99m Tc) (t ½ = 6.0 h) is the most widely used medical isotope in diagnostic imaging today. The National Nuclear Security Administration’s (NNSA’s) Material Management and Minimization (M3) program—established under the auspices of the American Medical Isotope Production Act—has been tasked with facilitating the work of domestic 99 Mo suppliers that do not use highly enriched uranium (HEU). Superconducting electron linear accelerators that employ high-Z converter targets can generate bremsstrahlung photons and neutron fluxes that can induce photonuclear reactions and uranium (U) fission. Argonne, in collaboration with industrial partners, has been developing the process chemistry for superconducting linear accelerator (LINAC)-irradiated triuranium octoxide (U 3 O 8 ) targets to produce 99 Mo. In short, the process involves dissolving the irradiated targets in nitric acid, which simultaneously releases valuable fission products (xenon and iodine). After adjusting the acid concentration, producers process the feed using tri-n-butyl phosphate (TBP), then mix the raffinate derived from this extraction cycle with a phosphinic acid such as Cyanex 272 or di(2-ethylhexyl) phosphoric acid (HDEHP), which is selective for Mo (molybdenum liquid liquid extraction [MoLLE]). In the next step, a concentration column comprising an anion exchange platform is used to decontaminate the remaining fission products, generating a pure 99 Mo stream in sodium hydroxide (NaOH)/sodium chloride (NaCl).

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Removal of High Specific Activity Fission Products from Uranyl Sulfate Waste Solutions

The Savannah River National Laboratory (SRNL) is currently providing support to SHINE Medical Technologies (SHINE) which plans to deploy a low energy, accelerator-based neutron source to fission low enriched U in a uranyl sulfate target solution for 99 Mo production. The 99 Mo is initially separated from the fission products and target solution by an extraction column. Subsequent washing of the column will generate waste solutions containing residual U and fission product activity. A small number of high specific activity fission products (e.g., 90 Sr, 137 Cs, and 144 Ce) in these streams will likely control the classification of the low level waste (LLW). If a sufficient amount of the high specific activity isotopes are separated from the SHINE waste streams and concentrated into a waste form, it would be possible to treat a majority of the wash solutions from the column operations as a lower class of LLW (Class A versus Class B or C or Class B versus Class C). The high specific activity fission product elements could then be dispositioned as a much smaller volume of waste rather than requiring the disposal of the entire waste stream at the higher waste classification. The Savannah River Site (SRS) has experience with using monosodium titanate (MST) and crystalline silicotitanate (CST) to remove Cs and Sr from high salt content waste solutions generated during the reprocessing of nuclear fuels and targets. Both of these materials have worked very well for their intended purposes at the SRS where the fission product elements are removed from highly alkaline waste. On the other hand, SHINE waste streams from the extraction column contain H 2 SO 4 which makes the solution acidic. Additionally, the SRS waste streams do not contain other fission product elements such as transition metals or lanthanides because they precipitate upon neutralization of the SRS waste and are not present in the supernate which is dispositioned as LLW following treatment. As such, there are inherent differences between SHINE and SRS waste treatment strategies. Savannah River National Laboratory was tasked with performing scoping studies to see if MST and CST would remove Sr, Cs, and Ce from an acidic mixed metal simulant solution. Batch contact experiments were performed using MST and two CST type materials. The MST material is a 15 wt % powder in 0.15 M NaOH slurry. The MST showed low adsorption for elements of interest from acidic solution. Furthermore, the powder size makes MST non-ideal for column operations. A CST IE-911 ion exchange material had high Cs adsorption, moderate Sr, and marginal Ce adsorption. Based on adsorption of all species, the ion exchange capacity was found to be 0.032 meq/mL. A bench-top column experiment to measure elemental breakthrough curves was performed using CST IE-911 where chromatographic separations of the mixed simulant were expected to occur. While most elements behaved as expected, the lanthanide series, containing Ce, broke though the column earlier than expected. The second CST material, CST R9120, displayed high adsorption for all elements in the acidic mixed simulant solutions in a batch contact study, and had a calculated loading capacity of 0.091 meq/mL. Future studies to develop a waste treatment flowsheet should focus on CST R9120 to treat SHINE waste solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modifications to LEU-Modified Cintichem Process to Allow Processing of Fission-Produced 99 Mo from Up to 750 g of Irradiated Uranium Targets

Eden Radioisotopes, LLC is developing a TRIGA* -sized, 2 MW all-target reactor system licensed from Sandia National Laboratories. This all-target reactor uses up to 60 Argonne-designed annular LEU-foil targets to produce fission-produced 99 Mo. The Eden low-enriched uranium (LEU) target contains ~137 g of uranium and will be irradiated at up to 30 kW, with a target burnup determined by the number of targets processed each week. The annular target utilizes metallic uranium foils 400 mm long by 72 mm wide by 250±25 -µm-thick covered by a thin (10 µm on each face) nickel fission barrier. The foil is placed between two 35 mil aluminum tubes that form the outer and inner walls of the cladding cylinder, which is hydroformed to provide high thermal conductivity within the target. Eden expects to process between three and five targets per process batch, to maintain the process batch 99 Mo inventory at or below 5000 curies to maintain high 99 Mo recovery, within a few hours of discharge from the reactor. Eden expects to meet full U.S. demand by processing 30 targets per week, with each target operated for two weeks at 30 kW. To meet less than full U.S. demand, both the target power and the number of targets processed weekly will be adjusted. With increased 99 Mo production and given the size of U targets processed in a single batch (up to ~750g-U), it is important to identify whether any modifications are needed to the LEU-modified Cintichem process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Materials Data on Na8ThU(MoO4)8 by Materials Project

Na8UTh(MoO4)8 crystallizes in the tetragonal I-4 space group. The structure is three-dimensional. there are two inequivalent Na1+ sites. In the first Na1+ site, Na1+ is bonded to six O2- atoms to form distorted NaO6 pentagonal pyramids that share corners with six MoO4 tetrahedra and edges with two equivalent NaO6 pentagonal pyramids. There are a spread of Na–O bond distances ranging from 2.39–2.60 Å. In the second Na1+ site, Na1+ is bonded to six O2- atoms to form distorted NaO6 pentagonal pyramids that share corners with six MoO4 tetrahedra and edges with two equivalent NaO6 pentagonal pyramids. There are a spread of Na–O bond distances ranging from 2.39–2.57 Å. U4+ is bonded in a 8-coordinate geometry to eight O2- atoms. There are four shorter (2.36 Å) and four longer (2.38 Å) U–O bond lengths. Th4+ is bonded in a 8-coordinate geometry to eight O2- atoms. There are four shorter (2.42 Å) and four longer (2.43 Å) Th–O bond lengths. There are two inequivalent Mo6+ sites. In the first Mo6+ site, Mo6+ is bonded to four O2- atoms to form MoO4 tetrahedra that share corners with six NaO6 pentagonal pyramids. There are a spread of Mo–O bond distances ranging from 1.77–1.85 Å. In the second Mo6+ site, Mo6+ is bonded to four O2- atoms to form MoO4 tetrahedra that share corners with six NaO6 pentagonal pyramids. There are a spread of Mo–O bond distances ranging from 1.76–1.86 Å. There are eight inequivalent O2- sites. In the first O2- site, O2- is bonded in a distorted trigonal planar geometry to one Na1+, one U4+, and one Mo6+ atom. In the second O2- site, O2- is bonded in a distorted trigonal planar geometry to one Na1+, one Th4+, and one Mo6+ atom. In the third O2- site, O2- is bonded in a 3-coordinate geometry to two Na1+ and one Mo6+ atom. In the fourth O2- site, O2- is bonded in a 3-coordinate geometry to two Na1+ and one Mo6+ atom. In the fifth O2- site, O2- is bonded in a distorted trigonal planar geometry to one Na1+, one Th4+, and one Mo6+ atom. In the sixth O2- site, O2- is bonded in a distorted trigonal planar geometry to one Na1+, one U4+, and one Mo6+ atom. In the seventh O2- site, O2- is bonded in a distorted trigonal planar geometry to two Na1+ and one Mo6+ atom. In the eighth O2- site, O2- is bonded in a distorted trigonal planar geometry to two Na1+ and one Mo6+ atom.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
NASA NTRS2005

The Palladium Isotopic Composition in Iron Meteorites

Ru, Mo and Pd are very useful indicators for the identification of nucleosynthetic components. We have developed techniques for Pd isotopes, in an effort to check the extent of isotopic effects in this mass region and for a Pt-group element which is less refractory than Ru. Stable Pd isotopes are produced by the process only (102Pd), the s-process only (104Pd), the process only (Pd-110) and by both the r- and s-processes (Pd-105, Pd-106, Pd-108). Kelly and Wasserburg reported a hint of a shift in 102Pd (approx. 25(epsilon)u; 1(epsilon)u (triple bonds) 0.01%) in Santa Clara. Earlier searches for Mo and Ru isotopic anomalies were either positive or negative.

Chen, J. H.↗
DOE OSTI2022

Impact of Irradiation and Decay Time on Chemical Speciation in Uranyl Sulfate Solution

A pH-1, 200-g U/L, highly enriched uranyl sulfate solution was neutron-irradiated using an accelerator-driven neutron source (the Universal Neutron Irradiator, UNI) to fission U-235 under conditions similar to those used by SHINE to produce Mo-99 in a subcritical assembly. We completed a series of five uranyl sulfate irradiations to assess the impact of irradiation and decay time on uranium precipitation and iodine speciation in solution. The duration of each irradiation was increased in order, with the shortest irradiation taking 2.9 hours and the longest, 64.5 hours. A total of 9.5E16 U-235 nuclei were fissioned in all irradiations, and a maximum fission power density of 0.26 W/mL was reached in the final irradiation. We observed no evidence of uranyl peroxide or other precipitates, demonstrating that the addition of Fe 2+ to a final Fe concentration of 200 ppm in the target solution was successful at preventing precipitation under representative irradiation conditions. Nonvolatile I - was found to be the dominant (>70%) iodine species in solution, both immediately after irradiation and after days to weeks of decay time.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Enhancing isotope mixing in U-10Mo downblend castings with electromagnetic stirring

Uranium alloyed with 10 wt percent molybdenum (U-10Mo) is the proposed fuel for use in the United States’ high-performance research reactors. The U-10Mo fuel is fabricated by a two-step casting process that downblends highly enriched uranium with depleted uranium (DU) and/or natural uranium (NU) and subsequently alloys the resulting high-assay low-enriched uranium (HALEU) with Mo. Currently, a two-step casting process is needed to meet the ingots’ U-235 enrichment homogeneity specifications. This work demonstrates the ability to provide more homogeneous U-235 distributions during downblend casting by using zirconia for crucibles rather than graphite. When graphite is used as a crucible, the electromagnetic field produced by the induction heater couples directly with the graphite. Zirconia is nonconducting in this study, and therefore does not couple with the field. The induction field couples directly with the metal, and causes electromagnetic stirring (EMS) in the molten metal pool. Eight downblend casting experiments were carried out in this work. Four were performed with zirconia crucibles and four with graphite crucibles. The U-235 enrichment was measured at nine discrete points in each casting using laser ablation multi-collector inductively coupled mass spectrometry (LA-MC-ICP-MS). The enrichment homogeneity was approximately eight times better in zirconia crucible castings than in graphite crucible castings as measured by the plate enrichment range and enrichment coefficient of variation. The results show a single casting step could be used to meet USHPRR enrichment specifications if electromagnetic stirring is present during casting. A related investigation was carried out to determine whether it is appropriate to measure enrichment on as-cast ingots or whether heat treated specimens should be used to accurately characterize enrichment. The results show that the as-cast microstructure plays a significant role in enrichment homogeneity. It is recommended that all enrichment measurements be done after the homogenization heat treatment typical of U-10Mo processing.

36 MATERIALS SCIENCE↗