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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Charged Defects in UO 2 Bulk and Surface: A First-Principles Study

Uranium dioxide (UO 2 ) is the primary fuel used in nuclear reactors. Under the extreme heat and radiation inside a reactor, this material inevitably develops defects in its crystal structure. To investigate the nature and behavior of these defects, DFT+U calculations were employed to investigate charged point defects in both bulk UO 2 and its most stable surface, the (111) plane. The formation of defects and their impact on the electronic structure were systematically examined. The results reveal that these defects introduce localized electronic states, alter magnetic behavior, and modify the structural properties. In general, such defects act as deep traps capable of capturing and retaining charge carriers. The stability of these defects depends strongly on the chemical environment and the position of the Fermi level. Surface defect calculations reveal that oxygen vacancies form more readily at the surface than in the bulk over a wide range of electron chemical potential, with subsurface oxygen vacancies being more stable than those in the top layer. Overall, the findings demonstrate how charged defects influence magnetism, transport, and stability in UO 2 , providing insights that may guide improvements in the safety and efficiency of nuclear fuel.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Hydration of α-UO 3 following storage under controlled conditions of temperature and relative humidity

Changes in chemical speciation of uranium oxides following storage under varied conditions of temperature and relative humidity are valuable for characterizing material provenance. In this study, subsamples of high purity α-UO 3 were stored under four sets of controlled conditions of temperature and relative humidity over several years, and then measured periodically for chemical speciation. Powder X-ray diffraction (XRD) analysis and extended X-ray absorption fine structure spectroscopy confirm hydration of α-UO 3 to a schoepite-like end product following storage under each of the varied storage conditions, but the species formed during exposure to the lower relative humidity and lower temperature condition follows different trends from those formed under the other three storage conditions (high relative humidity with high or low temperatures, and low relative humidity with a high temperature). Thermogravimetry coupled with XRD analysis was carried out to distinguish desorption pathways of water from the hydrated end products. Density functional theory calculations discern changes in the structure of α-UO 3 following incorporation of 1, 2 or 3 H 2 O molecules or 1, 2 or 3 OH groups into the orthorhombic lattice, revealing differences in lattice constants, U–O bond lengths, and U–U distances. The collective results from this analysis are in contrast to analogous studies that report that U 3 O 8 is oxidized and hydrated in air during storage under high relative humidity conditions.

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

Xenon–metal pair formation in UO 2 investigated using DFT + U

A recent experimental study on a spent uranium dioxide (UO 2 ) fuel sample from Belgium Reactor 3 identified a unique pair structure formed by the noble metal phase (NMP) and fission gas [xenon (Xe)] precipitate. However, the fundamental mechanism behind this structure remains unclear. The present study aims to provide an understanding of the interaction between five different metal precipitates [molybdenum (Mo), ruthenium (Ru), palladium (Pd), technetium (Tc), and rhodium (Rh)] and the Xe fission gas atoms in UO 2 , by using density functional theory (DFT) in combination with the Hubbard U correction to compute the formation energies involved. All DFT + U calculations were performed with occupation matrix control to ensure antiferromagnetic ordering of UO 2 . The calculated formation and binding energies of the Xe and solid fission products in the NMP reveal that these metal precipitates form stable pair structures with Xe. Notably, the formation energy of Xe–metal pairs is lower than that of the isolated single defects in all instances, with Pd and Mo showing the most favorable binding energy, likely accounting for the observed pair structure formation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Magnetoelastic standing waves induced in UO 2 by microsecond magnetic field pulses

Significance UO 2 is of particular interest due to its use as a nuclear fuel, its unusually poor thermal conductivity, and its uniquely strong magnetoelastic properties characterized by a linear coupling between strain and magnetization. We clearly observe mechanical standing waves and elastic resonances in a single crystal, induced by microsecond magnetic field pulses. The phase and amplitude of the mechanical resonances are modulated by the piezomagnetic switching behavior, offering a sensitive probe for otherwise hard to detect magnetic dynamics in UO 2 . We highlight the importance of magnetic domains in the attenuation of the mechanical resonances and the robustness of the antiferromagnetic state in UO 2 against externally applied magnetic fields to 150 T.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Submicrometer spectromicroscopy of UO 2 aged under high humidity conditions

We report the oxidation of uranium dioxide is a complicated process, depending on factors including humidity, temperature, and microstructure. To further determine the characteristics of this process, UO 2 particles were allowed to age and agglomerate under 98% relative humidity at room temperature for 378 days. A focused ion beam (FIB) section of this agglomeration was then measured at the O K-edge, U N 5 -edge, and C K-edge using the scanning transmission x-ray microscope (STXM) at the Advanced Light Source. O K-edge and U N 5 -edge x-ray absorption measurements allowed for the elemental and chemical species mapping of the agglomerates and indicated the formation of schoepite at the submicrometer scale in specific locations. Non-negative matrix factorization was employed to elucidate the main components at the O K-edge, which were uranyl (schoepite) formed primarily at the interface of the sample with controlled atmosphere, a UO 2 -like bulk component present in the majority of the sample, and an oxygen species present at the surface of the FIB section, which is likely adsorbed water. STXM spectromicroscopy measurements at the U N 5 -edge measurements also confirmed the location of oxidized uranium. This analysis is a valuable insight into the formation of schoepite on UO 2 and shows the sensitivity to and utility of STXM spectromicroscopy for uranium speciation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2021

Multiscale Modeling of Cr 2 O 3 -doped UO 2 Creep and Fracture

The pursuit of UO 2 fuel with various dopants (e.g., Cr 2 O 3 , Al 2 O 3 ) for improved accident tolerance due to larger grain sizes and suspected reductions in fission gas release and creep requires advanced modeling and simulation tools. These tools enable rapid multiscale development and understanding of material behavior in regimes where experimental measurements may be lacking. This work builds upon previous multiscale modeling efforts for fission gas behavior of Cr 2 O 3 -doped UO 2 to begin looking at the effects of dopants on mechanical properties with a particular focus on creep and fracture. The results indicate that dislocation climb is the most dominate creep mechanism at almost all temperatures with sensitivity analyses further providing evidence in support of that claim. Limited experimental data from the Advanced Fuels Campaign (AFC) program in the U. S. Department of Energy on tensile strength of Cr 2 O 3 -doped UO 2 indicates that fracture may be less severe in doped specimens. The results on creep and fracture are preliminary given the large uncertainty associated with the models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Compare Mechanistic Predictions for Doped UO 2 Mechanical Response and Other Properties with Empirical Models and Experimental Measurements

The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation program develops predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity, high resolution fuel performance tool at the engineering scale. As part of its development, additional modeling capabilities and improvements have been developed for relevant fuel forms. In this work, a fuel creep deformation model for Cr-doped fuel has been implemented into BISON, along with improvements to the empirical UO 2 fuel creep model based on experimental data and improvements to the radial power factor calculation for doped fuels. This work allows for more accurate simulation analyses for both UO 2 and doped-UO 2 fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Multiscale Modeling of Cr 2 O 3 -doped UO 2 Creep and Fracture (Rev.1)

The pursuit of UO 2 fuel with various dopants (e.g., Cr 2 O 3 and Al 2 O 3 ) for improved accident tolerance due to larger grain sizes and suspected reductions in fission gas release and creep requires advanced modeling and simulation tools. Such tools enable rapid multiscale development and enhanced understanding of material behavior in regimes for which experimental measurements may be lacking. This work builds on previous multiscale modeling efforts targeting the fission gas behavior of Cr 2 O 3 -doped UO 2 , in order to begin exploring the effects of dopants on mechanical properties, with a particular focus on creep and fracture. The results indicate that dislocation climb is the most dominate creep mechanism at almost all temperatures, and sensitivity analyses provide evidence in further support of this claim. The limited experimental data generated by the U.S. Department of Energy’s Advanced Fuels Campaign (AFC) on the tensile strength of Cr 2 O 3 -doped UO 2 indicate that fracture may be less severe in doped specimens. The results on creep and fracture are preliminary, given the large degree of uncertainty associated with the models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Charge-lattice coupling and the dynamic structure of the U–O distribution in UO 2+x

The different structures and behaviors of UO 2+x observed in crystallographic and local structure measurements were examined by extended X-ray absorption fine structure (EXAFS) measurements of pristine UO 2.0 , p + and He 2+ irradiated UO 2.0 , and, at multiple temperatures, bulk U 4 O 9 and U 3 O 7 and thin film U 4 O 9-δ on an epitaxial substrate. The disorder caused by irradiation is mostly limited to increased widths of the existing U–O/U pair distributions, with any new neighbor shells being minor. As has been previously reported, the disorder caused by oxidative addition to U 4 O 9 and U 3 O 7 is much more extensive, resulting in multisite U–O distributions and greater reduction of the U–U amplitude with different distributions in bulk and thin-film U 4 O 9 . This includes the significant spectral feature near R = 1.2 Å for all U 4 O 9 and U 3 O 7 samples fit with a U-oxo type moiety with a U–O distance around 1.7 Å. In addition to indicating that these anomalies only occur in mixed valence materials, this work confirms the continuous rearrangement of the U–O distributions from 10 to 250 K. Although these variations of the structure are not observed in crystallography, their prominence in the EXAFS indicates that the dynamic structure underlying these effects is an essential factor of these materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

In-situ ion irradiation of fission products in a spent UO 2 fuel

This study investigates the behavior of fission gas bubbles and five metal precipitates (5MPs) (Mo, Ru, Rh, Tc, Pd) in spent uranium dioxide (UO 2 ) fuel under various ion irradiation doses, temperatures, and flux conditions. Utilizing in-situ ion irradiation and advanced transmission electron microscopy, we analyzed the evolution of fission gas bubbles and 5MPs in UO 2 samples from the Belgium Reactor 3 (BR-3). Our findings reveal significant shrinkage of fission gas bubbles and 5MPs with increasing irradiation dose, accompanied by a decrease in pair density. We demonstrate that ion irradiation induces a homogeneous re-solution process where individual atoms are ejected from bubbles and precipitates, leading to their dissolution and subsequent re-precipitation in the matrix. In conclusion, this study provides critical insights into the dynamic behavior of fission products under irradiation, facilitating the development of predictive models and contributing to the optimization of nuclear fuel performance and safety.

Fission products↗
DOE OSTI2022

Modeling Heterogeneity in UO 2 Nanoparticles Using X‐ray Absorption Spectroscopy

Abstract EXAFS provides the capability to interrogate nanoparticle (NP) structure in atomistic detail without relying on long‐range crystallinity. There is a limitation in that EXAFS provides averaged structural information, making it difficult to separate a small amount of heterogeneous structure from bulk. In this work, models were developed to extract surface‐specific information from conventional EXAFS measurements collected on UO 2 NPs of varying size. Specifically, the surface terminating species of UO 2 NPs was determined from comparison of coordination numbers with geometric models while the origin of static disorder was interrogated from user‐defined simulations. Results show that the degree of oxygenation on the NP surface does not significantly deviate from bulk surface and that static disorder is highly enhanced in NP surface layers but cannot be attributed to surface relaxation effects alone. The approach described herein has the potential to be adapted to a range of inorganic NP systems to interrogate surface structure.

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

Quantification of uranium metal and total uranium in U-UO 2 mixtures using H 3 PO 4 -HCl

Here, the selective dissolution of uranium metal from its mixtures with UO 2 was studied to establish a simple and rapid quantitative method. Uranium metal was found to selectively dissolve as U(IV) into a solution of phosphoric and hydrochloric acids, with quantitative dissolution occurring within 1 h. Total dissolution of uranium as U(VI) could then be obtained by nitric acid addition and heating for 0.5–1 h. Analysis of the resulting U(IV) and U(VI) solutions by ICP-OES or UV–Vis spectroscopy yielded uranium metal-to-UO 2 ratios with an accuracy of ± 10% or better. Accuracy of the ICP-OES method was slightly better than UV–Vis, but the simplicity of the latter makes it an attractive option for some applications.

36 MATERIALS SCIENCE↗
DOE OSTI2026

High-gradient magnetic filtration of UO 2 colloids from aqueous suspensions

Environmental samples often contain innocuous components that make trace actinide analysis difficult and expensive. Therefore, it is expedient to separate these materials from actinide particles. One characteristic of many actinides is that they are paramagnetic. High-gradient magnetic filtration (HGMF) is a non-destructive technique that selectively captures magnetic constituents from a matrix containing non-magnetic species. Here, in this work, a novel HGMF device made of a permanent magnet array was used to effectively separate UO 2 particles from undesirable soil species. Up to 88% of the UO 2 particles present in suspensions were effectively captured by HGMF, resulting in their enrichment and a substantial reduction in the concentrations of other constituents.

Actinide separations↗
DOE OSTI2020

Grain Boundary and Lattice Fracture Toughness of UO 2 Measured Using Small-Scale Mechanics

Small-scale mechanical testing has been used to isolate the properties of individual grain boundaries and regions of the lattice in undoped UO 2 polycrystals. Notched microcantilever specimens were prepared by focused ion beam milling and tested in bending. Fracture toughness values were calculated from finite element analysis of the sample geometry at fracture load using the J-integral method. Lattice and grain boundary fracture toughness values of 1.17 ± 0.16 MPa m 1/2 and 1.08 ± 0.08 MPa m 1/2 , respectively, were obtained. The similarity between the lattice and grain boundary fracture toughness values is consistent with the preference for transgranular room-temperature fracture in UO 2 . The grain boundary and lattice fracture toughness values also agree well with the toughness of dense polycrystals, which suggests that microstructural-scale toughening mechanisms are limited in such samples.

36 MATERIALS SCIENCE↗
DOE OSTI2020

The predicted shapes of voids and Xe bubbles in UO 2

Morphology is a fundamental attribute when investigating voids and bubbles in UO 2 . Here we use molecular dynamics and Monte Carlo simulations to predict the lowest energy shapes for voids and bubbles in UO 2 . The energies of the {100}, {110} and {111} surfaces have been calculated and used to predict the equilibrium void shape from Wulff construction. This equilibrium shape is compared to low energy faceted voids exhibiting different relative proportions of each family of terminating surfaces. It is found that the equilibrium Wulff shape does not represent the lowest energy morphology for nm void sizes at temperatures between 300 K and 1200 K. Furthermore, the lowest energy faceted voids are slightly more energetically favourable than spherical voids, and as Xe is added, and bubble pressure increases, the faceted morphology becomes even more favourable than the spherical shape.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Small uranium and oxygen interstitial clusters in UO 2 : An empirical potential study

The aggregation of defect clusters in UO 2 under irradiations creates dislocation loops and cavities, which in turn contribute to swelling of the fuels, leading to detrimental pellet-clad mechanical interactions. The kinetics of the defect clusters, however, have received less attention and are not well understood. As such, using two interatomic potentials, the Morelon potential and a many-body model by Cooper, Rushton, and Grimes (CRG), the energetics and kinetics of small uranium interstitial clusters, with and without bound oxygen interstitials, are investigated with molecular dynamics and accelerated molecular dynamics simulations. For the single bound anti-Schottky interstitial cluster (U i 2O i ), both potentials find the interstitialcy migration mechanism to be a low barrier migration mechanism. However, for the CRG potential, an oxygen Frenkel pair assisted interstitial cluster migration mechanism with a lower migration barrier is identified during accelerated molecular dynamics simulations. For the Morelon potential, the interstitial clusters have geometrically compact structures, and small interstitial clusters (two and three anti-Schottky clusters) are more mobile than the single bound anti-Schottky interstitial cluster; for the CRG potential, the interstitial clusters have geometrically non-compact structures, yielding a more complex energy landscape. The rapid migration of interstitials and interstitial clusters are important for the irradiation response of UO 2 and correlates well with the very low barriers reported in a few historic experimental studies.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Micro-cantilever beam experiments and modeling in porous polycrystalline UO 2

Understanding the impact of microstructure on the thermo-mechanical behavior of oxide nuclear fuels is vital to predicting their performance through multiscale models. Evaluating the mechanical properties at the sub-grain length scale is key to developing these multiscale models. In this work, 3D finite element (FE) models were constructed to simulate the micrometer-scale bending of micro-cantilever beams fabricated using porous polycrystalline uranium dioxide (UO 2 ) and tested at room temperature. Here, the results showed that the porosity and elastic anisotropy of individual grains can play a significant role in determining the effective mechanical properties of the material deduced from the tests. Specifically, the porosity had a non-negligible effect, given that the pore size was of the same order of magnitude as the dimensions of the micro-beams. Correlations between load-deflection data, pore location, and elastic properties (effective Young's modulus) were investigated using UO 2 micro-beam FE models, where pore clusters were included and placed at different locations along the length of the beam. Results indicated that the presence of pore clusters near the substrate, i.e., the clamp of the micro-cantilever beam, has the strongest effect on the load-deflection behavior, with the porosity leading to a reduction of stiffness that is the largest for any location of the pore clusters. Furthermore, it was also found that pore clusters located towards the middle of the span and close to the end of the beam have a comparatively small effect on the load-deflection behavior. Therefore, it is concluded that accurate estimates of Young's modulus can be obtained from micro-cantilever experiments after accounting for porosity on the one third of the beam length close to the clamp. This, in turn, provides an avenue to improve microscale experiments and their analysis in porous, anisotropic elastic materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Characterization of the radial microstructural evolution in LWR UO 2 using electron backscatter diffraction

Studies on high burnup UO 2 subjected to loss-of-coolant accident conditions have shown that restructured regions of the fuel are susceptible to pulverization and eventual dispersal. Due to a lack of pre-test characterization, the distinct microstructural features rendering the fuel prone to fragmentation remain ambiguous. Four samples of commercially irradiated light-water reactor UO 2 have been characterized utilizing electron backscatter diffraction to assess the susceptible microstructure. The microscopy focused on determining the burnup and temperature conditions responsible for the formation of the different microstructural regions where the regions were denoted as the high-burnup structure (HBS), HBS transition, mid-radial, restructured central, and central region. Previous works have outlined the specific conditions required for the restructuring of the microstructure into the HBS, but the conditions responsible for the restructuring in the central region of the fuel are not well understood. The four analyzed samples confirm a burnup threshold of 61 GWd/tU, and an unknown temperature range is needed to facilitate the formation of the restructured central region. In conclusion, additional fuel performance evaluations are needed to quantify the temperature range promoting restructuring in the central region.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗