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At least 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Formation of uranium disulfide from a uranium thioamidate single-source precursor

A single-source-precursor approach was developed to synthesize uranium-based materials outside of the typically-studied oxides. This approach allows for shorter reaction times, milder reaction conditions, and control over the chemicals present in synthesis. To this end, the first homoleptic uranium thioamidate complex was synthesized as a precursor for US 2 materials. Pyrolysis of the thioamidate results in decomposition via an alkene elimination pathway and formation of γ-US 2 , which has historically been hard to access without the need for a secondary sulfur source. Despite the oxophilicity of uranium, the method successfully forms US 2 without the inclusion of oxygen in the bulk final product. These findings are supported by simultaneous thermal analysis, elemental analysis, powder X-ray diffraction, and uranium L 3 -edge X-ray absorption fine-structure spectroscopy. This work represents the first example of a single-source precursor approach to target and synthesize actinide materials other than the oxides.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Analysis of Using a Hybrid ZIRCEX Process to Generate High-Assay Low-Enriched Uranium (HALEU) - 20158

Idaho National Lab (INL) is investigating the feasibility of providing High-Assay Low Enriched Uranium (HALEU) by recovering Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) and downblending the HEU with natural, low-enriched, or depleted uranium (NU, LEU, or DU). That investigation identified that significant quantities of HALEU could be produced with a Hybrid ZIRCEX process to further advanced reactor research while reducing INL's environmental liability. Many new advanced reactor designs are being developed with improved safety, efficiency, and economics. Most require nuclear fuel with U-235 enrichment between 5% and 20%, which is defined as HALEU. Although technically feasible, there is no current domestic capability to make HALEU in the US. While it is anticipated that industry will provide HALEU through commercial enrichment once advanced reactors mature, until a market is established, an interim source of HALEU is needed to enable research and demonstration. An interim source of HALEU could be provided by recovering the HEU in some DOE-managed SNF and downblending it to between 5% and 20% using the Hybrid ZIRCEX Process. ZIRCEX is a dry head-end process to remove cladding (zirconium or aluminum) from SNF. Drying of the SNF is needed prior to introduction to ZIRCEX. After cladding removal, uranium and fission products in the bed material are oxidized and elutriated. In a Hybrid ZIRCEX process, the HEU process stream is sent to a very compact, modular solvent extraction process for uranium purification. The uranyl nitrate from solvent extraction is denitrated and calcined to produce a HALEU product ready for fuel fabrication. The long-lived fission products from solvent extraction are immobilized in glass using a small in-can melt. By removing the cladding before dissolution, the volume of glass waste is reduced by a factor of up to 300 times (when compared to not removing the cladding from the fuel). Other radioactive wastes from the process could be disposed of as Class A or Class B low-level waste. In addition to producing HALEU for advanced reactors, consumption of SNF would reduce DoE's environmental liability. Incorporating low-enriched spent fuel (versus exclusive use of LEU, DU, or NU) into the downblending scheme could reduce the SNF inventory even further. Several downblending and enrichment scenarios using INL fuels were investigated in this study. The results of this analysis show that the Hybrid ZIRCEX process could make a significant contribution to needed HALEU feedstock and reduce the environmental liability of managing SNF at INL. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Other Power Reactor Fuels

While oxide fuels have received the bulk of historic interest as power reactor fuels, a range of other uranium compounds and deployment architectures have also been considered and demonstrated to varying degrees. Metal alloys, technical ceramics, and oxycarbide or oxynitride particle fuels are candidates for various power reactor applications. Although in general these families of nuclear fuels have received more limited study than oxides, the benefits of uranium density, favorable thermophysical properties, and other factors have resulted in interest from both the traditional nuclear power industry as well as more recent entrepreneurial efforts. The dominant fuel chemistry issues for these fuel forms differ substantially from those of oxide systems. Synthesis routes, factors that affect fresh fuel properties, and critical aspects of fuel performance are unique to these systems and often dominated by the desired chemical behavior of uranium in each system. This chapter will briefly introduce metallic uranium alloys, specifically uranium-zirconium, non-traditional ceramic fuels such as carbides and nitrides, and oxycarbide/oxynitride particle fuels most familiar as the basis for TRISO fuel designs. A survey of major fuel chemistry issues is provided for each of these families of fuels, with a focus on contemporary challenges and active research avenues.

Nelson, Andrew↗
DOE OSTI2021

Uranium In Situ Electrolytic Deposition with a Reusable Functional Graphene-Foam Electrode

Nuclear fission produces 400 GWe which represents 11% of the global electricity output. Uranium is the essential element as both fission fuel and radioactive waste. Therefore, the recovery of uranium is of great importance. In this study, an in situ electrolytic deposition method to extract uranium from aqueous solution is reported. A functionalized reduced graphene oxide foam (3D-FrGOF) is used as the working electrode, which acts as both a hydrogen evolution reaction catalyst and a uranium deposition substrate. The specific electrolytic deposition capacity for U(VI) ions with the 3D-FrGOF is 4560 mg g –1 without reaching saturation, and the Coulombic efficiency can reach 54%. Moreover, reduction of the uranium concentration in spiked seawater from 3 ppm to 19.9 ppb is achieved, which is lower than the US Environmental Protection Agency uranium limits for drinking water (30 ppb). Furthermore, the collection electrode can be efficiently regenerated and recycled at least nine times without much efficiency fading, by ejecting into 2000 ppm concentrated uranium solution in a second bath with reverse voltage bias. All these findings open new opportunities in using free-standing 3D-FrGOF electrode as an advanced separation technique for water treatment.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Modeling a generic TRISO-fueled heat pipe microreactor using SCALE: Depletion, transportation criticality, and shielding

This paper demonstrates the applicability of the SCALE code system to tristructural-isotropic (TRISO)-fueled heat pipe microreactors through depletion, transportation criticality, and shielding analyses of a generic reference design. The study conducted supports US Nuclear Regulatory Commission code readiness efforts for advanced non–light-water reactor technologies and is intended as a code capability demonstration rather than as an optimization of a specific microreactor design. The modeled reactor employs high-assay low-enriched uranium (HALEU) uranium oxycarbide (UCO) TRISO fuel and beryllium oxide (BeO) reflectors and operates at 7.5 MWth with a nominal lifetime of about 3 effective full power years. Representative cases for fresh and irradiated cores were selected to exercise SCALE methods relevant to reactor operation and post-irradiation transport. The discharged-core decay heat is approximately 6% of operating power immediately after shutdown. Transportation criticality calculations show that internal water ingress is the dominant reactivity effect, with fully flooded fresh core and irradiated core configurations remain above the subcriticality criterion, even with the available control mechanisms. Shielding calculations for a simplified transportation package indicate that normal-condition dose rates are governed mainly by shielding thickness and cooling time, whereas the breached hypothetical accident case is governed primarily by cooling time. Overall, the study shows that SCALE supports depletion, transportation criticality, and shielding evaluations efficiently for TRISO-fueled heat pipe microreactors within a single code system.

Criticality↗
DOE OSTI2020

Controlling the Reduction of Chelated Uranyl to Stable Tetravalent Uranium Coordination Complexes in Aqueous Solution

The solution-state interactions between octadentate hydroxypyridinone (HOPO) and catecholamide (CAM) chelating ligands and uranium were investigated and characterized by UV-visible spectrophotometry and X-ray absorption spectroscopy (XAS), as well as electrochemically via spectroelectrochemistry (SEC) and cyclic voltammetry (CV) measurements. Depending on the selected chelator, we demonstrate the controlled ability to bind and stabilize U IV , generating with 3,4,3-LI(1,2-HOPO), a tetravalent uranium complex that is practically inert toward oxidation or hydrolysis in acidic, aqueous solution. At physiological pH values, we are also able to bind and stabilize U IV to a lesser extent, as evidenced by the mix of U IV and U VI complexes observed via XAS. Finally, CV and SEC measurements confirmed that the U IV complex formed with 3,4,3-LI(1,2-HOPO) is redox inert in acidic media, and U VI ions can be reduced, likely proceeding via a two-electron reduction process.

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

Design of UO 2 -BeO Critical Experiment at Sandia [Poster]

The purpose of this proposal is to design a new integral critical experiment to investigate the effects of beryllium oxide and high assay low-enriched uranium fuels. this proposal considers using several existing resources at Sandia: (1) the Critical Experiments (SCX) facility and water tank, (2) spare UO 2 -BeO fuel for the Annular Core Research Reactor (ACRR), and 7uPCX fuel rods from previous benchmark experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Evaluating FRAM v.6.1 with LANL and Euratom data

The United States Department of Energy and the European Atomic Energy Community (EURATOM) collaborate to test, evaluate, and improve the performance of the FRAM 6.1 code with various nuclear materials, especially mixed oxide (MOX) containing an isotopic blend of uranium and very high burnup plutonium. This joint effort is intended to promote effective implementation of the FRAM 6.1 code into nuclear material verification methods.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1969

Study of actinide chemistry in saturated potassium fluoride solution

Study concerning the chemistry of actinides in saturated KF solution included work with neptunium, uranium, and americium. Solubilities, absorption spectra, oxidation-reduction reactions, and solid compounds which can be produced in KF solution were examined. The information is used for preparation of various materials from salts of the actinides.

Cohen, D.↗
NASA NTRS1972

Lunar zirkelite - A uranium-bearing phase.

Zirkelite (simplified CaZrTiO5) containing 7-17 wt % oxides of trivalent elements (largely yttrium and the rare earths) and minor amounts of U, Th, and Pb is described from an Apollo 12 feldspathic peridotite (12036,9) and two KREEP-type norite lithic fragments separated from Apollo 14 loose fines (14163,39 and 14257,3). Quantitative electron microprobe analyses indicate that this lunar phase conforms more closely to the zirkelite formula than the zirconolite formula.

Busche, F. D.↗
DOE OSTI2025

Ab-initio informed cluster dynamics simulation of self- and Xe diffusivity in uranium mononitride under irradiation

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuels considered for light water reactor and advanced reactor designs, as it presents significant advantages such as high uranium density (better economics) and high thermal conductivity and melting point (increased safety). Self- and fission gas diffusivities need to be better understood, given that they influence key fuel performance phenomena like swelling and fission gas release. Recently, radiation enhanced diffusivity was investigated in UN by means of cluster dynamics simulations relying on empirical potential-based parameterizations, the reliability of which highly depends on the interatomic potential accuracy. Here, in this work, we refine this approach by determining, using ab-initio calculations, the properties of defect clusters containing vacancies, self-interstitials and Xe impurities. We also consider larger clusters than previous studies. The obtained dataset (formation enthalpies, entropies, and migration barriers) is used to parameterize a cluster dynamics model of mobile clusters, and to calculate the defect cluster concentrations under irradiation. This gives us access to the radiation enhanced self- and fission gas diffusivities. Although the resulting diffusivities are close to the values reported in the literature, we find important qualitative differences in the diffusion mechanisms. Capturing the correct mechanisms is crucial to properly describe the chemistry and fission rate dependence of the model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Dinuclear Complexes of Uranyl, Neptunyl, and Plutonyl: Structures and Oxidation States Revealed by Experiment and Theory

Dinuclear perchlorate complexes of uranium, neptunium, and plutonium were characterized by reactivity and DFT, with results revealing structures containing pentavalent, hexavalent, and heptavalent actinyls, and actinyl-actinyl interactions (AAIs). Electrospray ionization produced native complexes [(AnO 2 ) 2 (ClO 4 ) 3 ] - for An:An = U:U, Np:Np, Pu:Pu, and Np:Pu, which are intuitively formulated as actinyl(V) perchlorates. However, DFT identified lower-energy structures [(AnO 2 )(AnO 3 )(ClO 4 ) 2 (ClO 3 )] - comprising a perchlorate fragmented to ClO 3 , actinyl(VI) cation An VI O 2 2+ , and neutral AnO 3 . For U:U and Np:Np, and Np in Np:Pu, the coordinated AnO 3 is calculated as actinyl(VI) with an equatorial oxo, [O yl =An VI =O yl ][=O eq ], whereas for Pu:Pu, it is plutonyl(V) oxyl, [O yl =Pu V =O yl ][-O eq • ]. The implied lower stability of Pu VI versus Np VI indicates weaker Pu=O eq versus Np=O eq bonding. Adsorption of O 2 by the U:U complex suggests oxidation of U V to U VI , corroborating the assignment of perchlorate [(An V O 2 ) 2 (ClO 4 ) 3 ] - . DFT predicts the O 2 adducts are [(An VI O 2 )(O 2 )(An VI O 2 )(ClO 4 ) 3 ] - with actinyls oxidized from +V to +VI by bridging peroxide, O 2 2- . In accordance with reactivity, O 2- addition is computed as substantially exothermic for U:U and least favorable for Pu:Pu. Collision-induced dissociation of native complexes eliminated ClO 2 to yield [(AnO 2 )(O) 2 (AnO 2 )(ClO 4 ) 2 ] - , in which fragmented O atoms bridge as oxyl O -• and oxo O 2- to yield uranyl(VI) and plutonyl(VI), or as oxos O 2- to yield neptunyl(VII), [O yl =Np VII =O yl ] 3+ .

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

Uranium Chemistry: Identifying the Next Frontiers

While uranium is the most extensively studied actinide in terms of chemical properties, there remains much to be explored about its fundamental chemistry. Organometallic and organoactinide chemistry first emerged in the 1950s with research that found inspiration from transition-metal chemistry with the synthesis and characterization of uranocene, expanding new opportunities for organoactinide chemistry. Since then, a significant amount of research has pursued many avenues characterizing the fundamental nature of the f orbitals and their modes of bonding as well as their potential in catalysis. Uranium(III/IV) arene complexes dominate much of uranium organometallic chemistry, with bonding interactions stabilized by δ-back-bonding. Recent additions to this area of chemistry include the first UI and new additions of U II organouranium compounds. Uranium–transition metal complexes are still rare and maintain U IV oxidation states, with variable bond lengths determining the transition-metal oxidation state. Resultant reactivities are discussed as synthetic complexes, and unique bonding and coordination motifs are highlighted. In conclusion, this Viewpoint will focus on significant developments in uranium chemistry from the last 15 years while considering key areas for future research.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Technical Basis for Extending Storage Life of DOE-STD-3013 Containers

The U.S. Department of Energy (DOE) Technical Standard for Stabilization, Packaging, and Storage of Plutonium-Bearing Materials, DOE-STD-3013 [USDOE 2018], provides requirements for packaging stabilized metals, alloys, and oxides containing at least 30 weight percent (wt%) plutonium (Pu) plus uranium for storage for up to 50 years. This document provides the technical basis for extending that storage duration to up to 100 years for metals and alloys, and oxides that do not present a corrosion risk to the integrity of the container.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Uranium-Mediated Peroxide Activation and a Precursor toward an Elusive Uranium cis -Dioxo Fleeting Intermediate

The activation of chalcogen–chalcogen bonds using organometallic uranium complexes has been well documented for S–S, Se–Se, and Te–Te bonds. In stark contrast, reports concerning the ability of a uranium complex to activate the O–O bond of an organic peroxide are exceedingly rare. Herein, we describe the peroxide O–O bond cleavage of 9,10-diphenylanthracene-9,10-endoperoxide in nonaqueous media, mediated by a uranium(III) precursor [(( Me,Ad ArO) 3 N)U III (dme)] to generate a stable uranium(V) bis-alkoxide complex, namely, [(( Me,Ad ArO) 3 N)U V (DPAP)]. This reaction proceeds via an isolable, alkoxide-bridged diuranium(IV/IV) species, implying that the oxidative addition occurs in two sequential, single-electron oxidations of the metal center, including rebound of a terminal oxygen radical. Furthermore, this uranium(V) bis-alkoxide can then be reduced with KC 8 to form a uranium(IV) complex, which upon exposure to UV light, in solution, releases 9,10-diphenylanthracene to generate a cyclic uranyl trimer through formal two-electron photooxidation. Analysis of the mechanism of this photochemical oxidation via density functional theory (DFT) calculations indicates that the formation of this uranyl trimer occurs through a fleeting uranium cis-dioxo intermediate. At room temperature, this cis-configured dioxo species rapidly isomerizes to a more stable trans configuration through the release of one of the alkoxide ligands from the complex, which then goes on to form the isolated uranyl trimer complex.

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

Rehydration of metastudtite in the alteration kinetics of α– and β–U 3 O 8 in dilute aqueous solutions of hydrogen peroxide

The formation of alteration phases on uranium ore concentrates and used nuclear fuels under oxidizing conditions is key to understanding the potential mobility of radionuclides in the environment and designing optimal storage conditions of materials. However, the time-dependent distribution of alteration phases on α– and β–U 3 O 8 under oxidizing conditions has yet to be explored. Here, in this study, crystalline powders of α– and β–U 3 O 8 were submerged in aqueous solutions of hydrogen peroxide (1.6 × 10 −1 to 5.4 × 10 −5 M) with aliquots of solution and solid removed for analysis at 1, 8, 15, 22, 29, 36, 46, 58, 71, and 83 days. Within one day there is significant alteration of U 3 O 8 to the uranyl peroxide metastudtite, [(UO 2 )(O 2 )(H 2 O) 2 ], that is replaced by studtite, [(UO 2 )(O 2 )(H 2 O) 2 ]·2H 2 O, within a week regardless of the polymorph of U 3 O 8 or the initial concentration of H 2 O 2 in solution, as determined by partial least squares regression (PLSR) of Raman spectra collected from the solids. A dissolution/reprecipitation mechanism is proposed for both the alteration of U 3 O 8 to metastudtite and the subsequent alteration of both U 3 O 8 and metastudtite to studtite. The two polymorphs of U 3 O 8 exhibit similar rates and extents of alteration over time. The rehydration of metastudtite to studtite has not been previously reported and highlights the need for future work to determine the mechanism by which metastudtite is converted to studtite and what other conditions facilitate this rehydration.

Alteration of U3O8↗
DOE OSTI2022

IER-523: Feasibility of Experiments Focused on Measuring the Effects of UO 2 BeO Material on Critical Configurations using 7uPCX [Slides]

Sandia National Labs has access to unused ACRR fuel, which is unique in its enrichment 35% and material composition BeO. ACRR fuel is available in quantities well above what is needed for experiments. Two experiment concepts have been investigated: UO 2 BeO fuel elements and pellets with 7uPCX fuel. The worth of UO 2 BeO is large enough to be well above the anticipated experiment uncertainties.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗