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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.

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At least 505 records · Page 28
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Synthesis of uranium mixed anion compounds synthesized using the Boron-Chalcogen Mixture method: Ba 6 Co 6 U 0·9 1S 13·5 O 0.5 and Ba 5·47 K 0·53 Zn 6 US 13·5 O 0.5

Mixed anion compounds have exhibited interesting structures and properties that differ from compounds only incorporating a single anion in their composition. Unfortunately, difficulties in the synthetic methods used to obtain these materials has slowed the evolution of this field, prompting investigations into alternate synthetic pathways to these materials. Our recently establish Boron-Chalcogen Mixture (BCM) method, which was originally developed for the synthesis of pure actinide chalcogenides from oxides, has been adapted to achieve the partial oxide to sulfide conversion of Ba 2 MUO 6 (M = Co, Zn) which resulted in two new uranium (IV/V) oxysulfide compounds, Ba 6 Co 6 U 0·91 S 13·5 O 0.5 and Ba 5·47 K 0·53 Zn 6 US 13·5 O 0.5 These compounds crystallize in the tetragonal crystal system adopting the space group I4/mcm. Finally, their syntheses, crystal structures, and trends observed in the pursuit of these new mixed anion compounds are reported.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

Determination of phosphorus and sulfur in uranium ore concentrates by triple quadrupole inductively coupled plasma mass spectrometry

The analysis of impurities in a uranium ore concentrate (UOC) could provide information regarding the source, production history, and potential intended use of the UOC. This study involves the analysis of UOC samples for phosphorus and sulfur. Here, concentrations were determined by triple quadrupole inductively coupled plasma – mass spectrometry and compared with results from a pyrohydrolysis method as well as previously reported results. The sulfur and phosphorus concentrations, determined by the mass spectrometer, were used to explore possible trends in a series of UOC material, and the uncertainties were calculated using GUM workbench software. The triple quadrupole inductively coupled plasma – mass spectrometer method allows for the removal of interferences in the analysis of species.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2020

The Stability of SiO{sub 2} Fe{sub 3}O{sub 4} Nanoparticles for Uranium Extraction in Acidic Media - 20189

This comprehensive study has shown that the proposed strategy of SiO{sub 2} coating of Fe{sub 3}O{sub 4} nanoparticles (NPs) for the use of uranium (U) extraction in acidic media was successful. The breakdown of Fe{sub 3}O{sub 4} in acidic media has proven to be a challenge when applying to a nuclear waste removal scenario, leading to the dissolution of Fe{sub 3}O{sub 4} and preventing uptake of U ions in the waste. To overcome this a SiO{sub 2} passivating layer was grafted to the Fe{sub 3}O{sub 4} NPs, to which an acid stability test was conducted between pH 3 and pH 4. It has been shown via Transmission Electron Microscopy (TEM) and zeta potential measurements that the uncoated Fe{sub 3}O{sub 4} NPs exhibit rapid coagulation when exposed to acidic conditions, moving from a stable zeta potential of -44 mV to an unstable zeta potential of +6 mV. SiO{sub 2} coated Fe{sub 3}O{sub 4} on the other hand can remain stable for several hours and continue to be stable beyond that. Throughout acid exposure the zeta potential of these NPs remained at ∼ -38 mV, showing great stability. As a functionality is needed on the surface of the NPs for separation and extraction, the phosphate complex used was also tested for its acid stability. It was determined that with a constant zeta potential of ∼ -20 mV, the phosphate functionality was able to remain stable in the acidic conditions simulating the environment of certain nuclear wastes. Overall SiO{sub 2} coated (PO)x-Fe{sub 3}O{sub 4} NPs show great promise for application in separation and removal of U in acid media. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Induced Ferromagnetism in Epitaxial Uranium Dioxide Thin Films

Abstract Actinide materials have various applications that range from nuclear energy to quantum computing. Most current efforts have focused on bulk actinide materials. Tuning functional properties by using strain engineering in epitaxial thin films is largely lacking. Using uranium dioxide (UO 2 ) as a model system, in this work, the authors explore strain engineering in actinide epitaxial thin films and investigate the origin of induced ferromagnetism in an antiferromagnet UO 2 . It is found that UO 2+ x thin films are hypostoichiometric ( x <0) with in‐plane tensile strain, while they are hyperstoichiometric ( x >0) with in‐plane compressive strain. Different from strain engineering in non‐actinide oxide thin films, the epitaxial strain in UO 2 is accommodated by point defects such as vacancies and interstitials due to the low formation energy. Both epitaxial strain and strain relaxation induced point defects such as oxygen/uranium vacancies and oxygen/uranium interstitials can distort magnetic structure and result in magnetic moments. This work reveals the correlation among strain, point defects and ferromagnetism in strain engineered UO 2+ x thin films and the results offer new opportunities to understand the influence of coupled order parameters on the emergent properties of many other actinide thin films.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Chlorination of uranium metal in molten NaCl-CaCl 2 via bubbling HCl

Molten chloride salt fast reactors (MCFRs) will require UCl 3 dissolved in molten salt mixtures as fuel for nuclear fission. For infusing the salt with UCl 3 , bubbling HCl into NaCl-CaCl 2 in contact with U metal was investigated. The reaction was run up to 9 h and yielded U concentration up to 0.652 wt.%. Open circuit potential between a W electrode and Ag/AgCl reference electrode yielded a potential consistent with uranium existing as U(III) in the salt. Furthermore, this demonstrates that HCl can be a very effective chlorinating agent to infuse MCFR fuel with UCl 3 starting from U metal.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Updated uranium reference material 231 Pa/ 235 U consensus model ages for nuclear forensics

Radiochronometry provides the model age of nuclear materials, which is useful for understanding the production history of materials found outside of regulatory control. Certified reference materials (CRMs) are important for radiochronometry to increase confidence in measurement quality; however, the absence of 231 Pa/ 235 U CRMs necessitates that nuclear forensic laboratories measure the 231 Pa/ 235 U model ages of preexisting uranium reference materials. Here, in this work, new consensus 231 Pa/ 235 U model ages are reported from three nuclear forensic laboratories for three reference materials using current analytical methods. These updated consensus values can be used for quality control of 231 Pa/ 235 U model age measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Neutron capture signatures determined from gadolinium isotope compositions of uranium ore concentrates using MC-ICPMS

We present Gd isotope compositions on 25 well-characterized uranium ore concentrates (UOCs). About half the UOCs have isotope depletions in 157 Gd coupled with anticorrelated excesses in 158 Gd due to neutron capture effects. UOCs from older ore bodies and with higher U contents show larger neutron capture effects than UOCs from younger ore bodies with lower U contents. These Gd data are correlated with their previously measured Sm isotope compositions, and we use the Sm data to estimate the neutron fluence of these samples. In conclusion, this work demonstrates how Gd isotope signatures can be employed as a new tool for nuclear forensics.

Earth Sciences↗
DOE OSTI2025

Quantifying uncertainty in uranium concentration measurements via K-edge densitometry

This study quantifies the uncertainty in uranium concentration predictions of fluoride and chloride-based salts within a steel pipe using K-edge densitometry. Modeling and simulation was conducted with the Monte Carlo N-Particle Transport (MCNP) code. The quality of of this technique’s prediction in a pipe requires proper characterization of the pipe’s thickness, which is dependent on the source size and axial offset from the pipe centerline. The thickness was determined as either the center-line thickness seen by the X-ray source or an average value determined through random sampling. Generally, the predicted concentrations were slightly better at lower offset with the random sampling thickness and using the center-line thickness for the highest offsets. For a line-beam source and varying axial offsets, the relative error of concentration was within 1% of the true value but uncertainty increased by 2 orders of magnitude. Similarly, for no axial offset, the relative error was significantly less than 1% while no trend for uncertainty was found. However, at the largest possible offset for a given source size, the concentrations become erroneous and greater than the allowable 1% relative error. Furthermore, high offsets tended to increase the variance of the transmission spectra by 3 orders of magnitude.

Characterization and Analytical Technique↗
DOE OSTI2026

Characterizing microscale signatures in uranium ore concentrates using electron probe microanalyzer

Impurities in uranium ore concentrates (UOCs) serve as forensic signatures of processing history and origin. Here, this study utilizes Electron Probe Microanalyzer (EPMA) to characterize microscale compositional and textural features in individual UOC particles from both commercial and bench-scale production. At the particle scale, multiple internal phases with distinct morphologies, chemical signatures, and stoichiometries are documented. Our data shows that chemical impurities are heterogeneously distributed within single particles and among particles within a sample. These microscale heterogeneities correlate with known processing histories, indicating that microscale signatures of early fuel cycle materials can provide valuable information for nuclear forensic material analysis.

organic↗
DOE OSTI2026

Ra-226 extraction and purification from simulated uranium tailings

Growing demand for Ac-225 has outstripped current production, increasing interest in alternative production routes. Furthermore, these often rely on irradiations of Ra-226, yet Ra-226 supplies are also limited. Uranium tailings provide an abundant source of Ra-226, though separation from these complex matrices remains challenging. A laboratory-scale process was developed to recover Ra-226 from simulated U-tailings. After acid leaching, Ra remained adsorbed on refractory oxides and silicates. An additional CaCl₂ leach of the residue raised recovery to ~ 64%. Subsequent purifications, including precipitation, ion-exchange, and extraction chromatography (TK101), achieved chemical yields up to 73 ± 7%, producing 7.68 ± 0.1 µCi of high-purity Ra-226.

and nuclear chemistry↗
DOE OSTI2020

Reticular exploration of uranium-based metal—organic frameworks with hexacarboxylate building units

We report that the rational reticular design of metal–organic frameworks (MOFs) from building units of known geometries is essential for enriching the diversity of MOF structures. Unexpected and intriguing structures, however, can also arise from subtle changes in the rigidity/length of organic linkers and/or synthetic conditions. Herein, we report three uranium-based MOF structures—i.e., NU-135X (X = 0, 1, 2)—synthesized from trigonal planar uranyl nodes and triptycene-based hexacarboxylate ligands with variable arm lengths. A new chiral 3,6-connected nuc net was observed in NU-1350, while the extended versions of the ligand led to 3-fold catenated MOFs (NU-1351 and NU-1352) with rare 3,6-connected cml-c3 nets. Lastly, the differences in the topology of NU-1350 and NU-1351/NU-1352 could be attributed to the slight distortions of the shorter linker in the former from the ideal trigonal prism geometry to an octahedral geometry when coordinated to the trigonal planar uranyl nodes.

3,6-connected net↗
DOE OSTI2020

Modelling uranium dioxide corrosion under repository conditions: A pore-scale study of the chemical and thermal processes

This paper investigates the corrosion behaviors of spent uranium dioxide (UO 2 ) fuel when placed in geological repositories. We performed pore-scale reactive transport simulations of the UO 2 corrosion process in a defective fuel rod with different orientations of breach or fracture on its clad. It is found that the corrosion rate has strong dependency on the breach orientation. The highest corrosion rate of UO 2 is calculated when the angle between the groundwater flow direction and the clad fracture reaches 180°. UO 2 corrosion simulations with different flow rates are compared. The results show the higher flow rate accelerates the corrosion of UO 2 fuel. The effect of pH on the corrosion process is also determined. It is indicated that higher pH can limit the corrosion of UO 2 by reducing the reaction rate. The dependence of reaction rate and time on reactive surface area is explored. Spent fuel with lower surface area demonstrates longer lifetime under corrosion conditions. A numerical model coupling thermal conduction and chemical reactions is developed to assess the impact of temperature on the process of UO 2 corrosion. The results show that higher temperature leads to larger corrosion rates for UO 2 . Here, the predicted reaction rates are higher in comparison with the isothermal results. Hence, our results can help improve the fundamental understanding of UO 2 corrosion in geological repositories for long-term storage of spent nuclear fuels and provide guidance for the safe operations and selection of appropriate repositories.

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

Modelling and analysis of salt-convection effect on oxide reduction process for uranium oxides using smoothed particle hydrodynamics

Oxygen ion transport via molten salt continuum is a key kinetic feature of electrochemical reduction process for uranium oxides in molten salt. During the electrolysis, the evolution of oxygen ion concentration field in molten salt continuum appears from porous metallic layer to open bulk salt. Despite the important role affecting process throughput, its effect coupled with controlled salt movement was not adequately investigated. To capture the reactive oxygen ion transport phenomena, a multidimensional reduction model based on smoothed particle hydrodynamics (SPH) was combined with the advection-diffusion model. A pelletized oxide fuel bed capturing interconnected salt continuum is considered as a reactant material geometry, which allows forced convective delivery of molten salt through the reactant materials with a relatively low pressure head. Here, in this study, reduction time of a pelletized-fuel bed was quantified with respect to key design parameters of the reactant material configuration. The simulation results demonstrated the advantage of salt convection which effectively removes oxygen ion products from the reactant materials and consequently accelerates the reduction process. Moreover, the reduction-time correlation of a single pelletized fuel was developed as a function of relevant dimensionless parameters and was extended to the application for a pelletized-fuel bed. Finally, several insights and implications in the use of a pelletized oxide fuel were presented. The developed SPH-based framework enabled multi-physics analysis of governing dynamics in appreciating and screening candidate reactant material configurations in a cost-effective manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Reactions of atomic thorium and uranium cations with CF 4 studied by guided ion beam tandem mass spectrometry

The fundamental chemistry of the actinides continues to be an area of interest because of their use in nuclear fuel cycles. Here, the kinetic energy dependences of the reactions of thorium and uranium cations with CF 4 were studied using guided ion beam tandem mass spectrometry. The products observed include AnF x + (An = Th or U, x = 1-3) and CF y + (y = 1-3). Although most reactions are exothermic according to literature thermochemistry, all cross sections display strong endothermic features indicating a restriction in the reactivity at low energies. For both actinides, only the AnF + + CF 3 product channel is observed at low energies; however, its formation proceeds very inefficiently. Furthermore, these results indicate that the reactions of Th + and U + with CF 4 are not thermodynamically driven and suggest that the low-energy reactivity results from precise alignment of the reactants.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Evolving radon diffusion through earthen barriers at uranium waste disposal sites

Field measurements of Rn-222 fluxes from the tops and bottoms of compacted clay radon barriers were used to calculate effective Rn diffusion coefficients (D Rn ) at four uranium waste disposal sites in the western United States to assess cover performance after more than 20 years of service. Values of D Rn ranged from 7.4 × 10 -7 to 6.0 × 10 -9 m 2 /s, averaging 1.42 × 10 -7 . Water saturation (S W ) from soil cores indicated that there was relatively little control of D Rn by S W , especially at higher moisture levels, in contrast to estimates from most steady-state diffusion models. Further, this is attributed to preferential pathways intrinsic to construction of the barriers or to natural process that have developed over time including desiccation cracks, root channels, and insect burrows in the engineered earthen barriers. A modification to some models in which fast and slow pathway D Rn values are partitioned appears to give a good representation of the data; 4% of the fast pathway was needed to fit the data regression. For locations with high S w and highest D Rn (and fluxes) at each site, the proportion of fast pathway ranged from 1.7% to 34%, but for many locations with lower fluxes, little if any fast pathway was needed.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Controlled current-rate AC flash sintering of uranium dioxide

Uranium dioxide (UO2) pellets with controlled microstructures were densified up to 93.4% of their theoretical density in less than 25 minutes at a furnace temperature of 873 K, utilizing controlled current-rate alternating current (AC) flash sintering (FS). Using this AC-FS method it was possible to control the sintering rate and thermal gradients, resulting in dense pellets with no appreciable hourglassing and good mechanical integrity. Moreover, the apparent sintering activation energy for FS and for conventionally sintered samples was estimated to be 108 kJ mol -1 and 380 kJ mol -1 , respectively using the master sintering curve method. The apparent activation energy for FS was remarkably close to those reported in the literature for spark plasma sintering of UO 2 . Furthermore, both these field assisted sintering methods utilize fast heating rates and electrical effects that are likely enhancing the grain boundary diffusion mechanism. Finally controlled current-rate AC-FS has been demonstrated as a technological advancement, capable of producing ceramic nuclear fuels in a fraction of the conventional processing time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

In situ monitoring of microstructure evolution during thermal processing of uranium-zirconium alloys using laser-generated ultrasound

In this study, laser-generated ultrasound was used to monitor microstructure evolution during thermal processing of as-cast, polycrystalline binary uranium-zirconium metallic fuel alloys with compositions of U-20wt.%Zr (U-20Zr), U-50wt.%Zr (U-50Zr) and U-80wt.%Zr (U-80Zr). Ultrasonic waveforms were recorded during heating and cooling the samples from room temperature to >973 K and back. A phase transition temperature for all three compositions was estimated from the temperature at which an abrupt and rapid reduction in ultrasonic velocities was observed. Microstructural features on the length scale of tens of micrometers were inferred from the observation of scattering of ultrasonic waves by elastic heterogeneities above ~823 K in U-20Zr, while a hysteresis in the ultrasonic velocities of U-80Zr upon cooling was attributed to a partial retention of the high temperature phase following thermal annealing. The U-50Zr alloy exhibited a reversible viscoelastic response above 933 K, as evidenced by the observation of high frequency attenuation of the shear component of the waveforms at high temperature. Ultrasonic measurements were supplemented by in situ transmission electron microscopy (TEM). The TEM images revealed that the δ-U-Zr matrix in the three compositions underwent a spinodal decomposition above ~823 K into nanoscale regions. The ultrasonic measurements revealed larger, micron-scale structure evolution in the U-20Zr alloy at the same temperature. This large-scale structure is associated with heterogeneous regions having different Zr content. Our findings show the potential heating rate dependence of microstructural evolution in U-Zr alloys and highlight differences in the thermomechanical response and associated length scales during thermal annealing between single- and dual-phase compositions. These results demonstrate the utility of laser ultrasonics to rapidly and efficiently scan phase boundaries and monitor micrometer-scale structure evolution in metallic fuel alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗