Nuclear Forensics-by-Design for Material Security: Research Strategies and Outcomes
Explore the source record for details and available documents.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
SEARCH · Search NASA
Results for “Uranium oxides”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Moon base reactor system
The objective of our reactor design is to supply a lunar-based research facility with 20 MW(e). The fundamental layout of this lunar-based system includes the reactor, power conversion devices, and a radiator. The additional aim of this reactor is a longevity of 12 to 15 years. The reactor is a liquid metal fast breeder that has a breeding ratio very close to 1.0. The geometry of the core is cylindrical. The metallic fuel rods are of beryllium oxide enriched with varying degrees of uranium, with a beryllium core reflector. The liquid metal coolant chosen was natural lithium. After the liquid metal coolant leaves the reactor, it goes directly into the power conversion devices. The power conversion devices are Stirling engines. The heated coolant acts as a hot reservoir to the device. It then enters the radiator to be cooled and reenters the Stirling engine acting as a cold reservoir. The engines' operating fluid is helium, a highly conductive gas. These Stirling engines are hermetically sealed. Although natural lithium produces a lower breeding ratio, it does have a larger temperature range than sodium. It is also corrosive to steel. This is why the container material must be carefully chosen. One option is to use an expensive alloy of cerbium and zirconium. The radiator must be made of a highly conductive material whose melting point temperature is not exceeded in the reactor and whose structural strength can withstand meteor showers.
Chavez, H.↗
Oxide structure of air-passivated U-6Nb alloy thin films
U-6Nb is a uranium alloy containing 6 wt% niobium that possesses high corrosion resistance. The structure and composition of the passivating oxide layer formed on air-aged U-6Nb, which gives the material its corrosion resistant properties, was characterized using surface scattering techniques. Stable oxide layers formed on the surface of a set of U-6Nb alloy thin films under ambient conditions were investigated using neutron reflectometry (NR), x-ray reflectometry (XRR) and grazing incidence x-ray diffraction (GIXD). The passivating oxide was composed of approximately 27% U, 5% Nb, and 68% O, primarily consisting of a thin niobium oxide layer (5.5 ± 0.4 nm) separating a thicker UO2 layer (27.1 ± 2.3 nm) from the underlying U-6Nb alloy. A critical density of enriched niobium oxide at the metal-oxide interface is hypothesized to limit oxygen diffusion and confer high corrosion resistance to the alloy.
36 MATERIALS SCIENCE↗
ThO 2 and Th 1– x U x O 2 Nanoscale Materials and Thin Films for Nuclear Science Applications
This study investigates the dynamics and mechanisms of solution combustion synthesis (SCS) for the preparation of nanoscale ThO 2 and Th 1–x U x O 2 materials, utilizing metal nitrates (Th(NO 3 ) 4 and UO 2 (NO 3 ) 2 ) and acetylacetone (C 5 H 8 O 2 ) as reactants dissolved in a 2-methoxyethanol (C 3 H 8 O 2 ) solvent. By combining thermodynamic calculations, dynamic time–temperature profile measurements with differential scanning calorimetry (DSC) and thermogravimetric analysis (TGA), this research reveals how variations in acetylacetone concentration and uranium content influence the structural parameters of the synthesized oxides. The time–temperature measurements show that the heating rate and maximum combustion temperatures are sensitive to acetylacetone concentration. DSC-TGA results indicate shifts in exothermic peak temperatures as the uranium content changes. The complexation between thorium and acetylacetone emerges as a critical factor, impacting combustion parameters and the structural characteristics of the final products. The uniform distribution of Th and U in the Th 1–x U x O 2 solid solution and the formation of nanoscale particles with strained crystallites are considered essential for the low-temperature densification of these materials for nuclear fuel pellet applications. Additionally, high-quality ThO 2 and Th 1–x U x O 2 thin (100–150 nm) films are successfully synthesized via electrospray deposition of combustible solutions followed by a brief period of heat treatment. Furthermore, these films exhibit excellent structural and morphological uniformity, making them ideal candidates for nuclear measurements, irradiation damage studies, and investigations into the physical properties of both pure and mixed oxides.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
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↗
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↗
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↗
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↗
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↗
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↗
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↗
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 ↗
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.↗
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.↗
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↗
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↗
The diffusion coefficient of oxygen into alpha-uranium at 323 °K determined by grazing angle X-ray diffraction technique
Not Available
36 MATERIALS SCIENCE↗
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↗