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At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Impact of Controlled Storage Conditions on the Hydrolysis and Surface Morphology of Amorphous-UO 3

The hydration and morphological effects of amorphous (A)-UO 3 following storage under varying temperature and relative humidity have been investigated. This study provides valuable insight into U-oxide speciation following aging, the U-oxide quantitative morphological data set, and, overall, the characterization of nuclear material provenance. A-UO 3 was synthesized via the washed uranyl peroxide synthetic route and aged based on a 3-factor circumscribed central composite design of experiment. Target aging times include 2.57, 7.00, 14.0, 21.0, and 25.4 days, temperatures of 5.51, 15.0, 30.0, 45.0, and 54.5 °C, and relative humidities of 14.2, 30.0, 55.0, 80.0, and 95.8% were examined. Following aging, crystallographic changes were quantified via powder X-ray diffraction and an internal standard Rietveld refinement method was used to confirm the hydration of A-UO 3 to crystalline schoepite phases. The particle morphology from scanning electron microscopy images was quantified using both the Morphological Analysis of MAterials software and machine learning. Results from the machine learning were processed via agglomerative hierarchical clustering analysis to distinguish trends in morphological attributes from the aging study. Significantly hydrated samples were found to have a much larger, plate-like morphology in comparison to the unaged controls. Predictive modeling via a response surface methodology determined that while aging time, temperature, and relative humidity all have a quantifiable effect on A-UO 3 crystallographic and morphological changes, relative humidity has the most significant impact.

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

Polymorphism and phase transitions in layered uranium(VI) hydroxides: Ab initio lattice dynamics simulations of UO 2 (OH) 2

The phase transitions and thermodynamics of stoichiometric α-, β-, and γ-UO 2 (OH) 2 polymorphs are investigated using density functional perturbation theory. The pressure-induced β(Pbca) → α(Cmca) phase transition is reproduced by calculations, with a volume reduction of ΔV/V = -14.7% similar to experiment. Consistent with observation, a temperature-driven γ(P2 1 /c) → β(Pbca) phase transition is predicted near 533 K. At 298.15K, the computed standard molar heat capacity of α-UO 2 (OH) 2 is C p 0 = 112.1 J mol -1 K -1 , only 1.6% smaller than the value of C p 0 = 113.96 ± 0.12 J mol -1 K -1 measured by calorimetry. C p 0 = 112.4 and 104.8 J mol -1 K -1 are predicted for the β- and γ-UO 2 (OH) 2 polymorphs, respectively. The calculated molar enthalpy and Gibbs energy functions of the α-, β-, and γ-UO 2 (OH) 2 polymorphs are also reported.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2022

Advanced Microscopy for Characterization of High Burnup Commercial UO 2 Fuel Before and After LOCA Testing

This report presents the results from the microstructural examination of high burnup UO 2 under pre- and post- loss of coolant accident (LOCA) conditions. The work in this report builds and expands upon work previously conducted on high burnup UO 2 fuel at Oak Ridge National Laboratory (ORNL). This work investigates the microstructural evolution that occurs during steady state irradiation of UO 2 , particularly in high burnup fuel. Several high burnup fuel samples were available from historic fuel shipments to ORNL. Microstructural characterization will aid in understanding why the fuel is susceptible to fuel fragmentation and support future work to investigate the formation mechanisms of certain features that form under irradiation. High burnup UO 2 microstructural data is also needed to improve constitutive models intended to predict high burnup fuel fragmentation (HBFF). This report summarizes the microstructural evaluation of these fuel samples to interpret and analyze high burnup fuel microstructures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Molecular dynamics simulations of fission gas xenon (Xe) diffusion at UO 2 grain-boundaries (Rev.1)

The diffusivity of fission gas xenon (Xe) at UO 2 grain-boundaries is one of the most important parameters in mechanistic modeling of fission gas diffusion in UO 2 based nuclear fuels. In this report, we use molecular dynamics simulations to investigate the Xe diffusivity in UO 2 grain-boundaries, employing the many-body potential developed by Cooper, Rushton and Grimes for UO 2 . Three different types of grain-boundaries are investigated, twist Σ5, tilt Σ5, and a random grain-boundary. Diffusion activation energies in the range of 0.39 – 1.46 eV are obtained for the Xe diffusivity. Comparison to results for the uranium vacancy diffusivity from MD simulations employing the same methodology suggests a weak to moderate attractive Xe-uranium vacancy binding energy depending on the grain-boundary type.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Diffusional creep in UO 2 informed by lower length scale simulations

Using molecular dynamics, we predict information at the atomistic scale used to develop a mechanistic UO 2 creep model for use in higher length-scale fuel performance codes. The ultimate objective of the model is to not only to capture the creep rates of UO 2 but to determine the dominant mechanism in the diffusional regime, which is still debated in the literature. It is important to have a model to capture the correct mechanisms for creep in UO 2 as this can be used as the foundation when applying to other fuels, such as doped UO 2 , and when irradiation is accounted. In last years NEAMS milestone (FY22), we developed a prelimnary model, however there were issues, for example, excessively high values of uranium vacancy concentrations at the grain boundary. This year we have addressed the issues with the previous version of the model, added a new term that accounts for the nucleation of dislocations at stress raisers (e.g., triple junctions) within the microstructure and discussed where there was disagreement in the literature about the underpinning physics (uranium self-diffusion at the grain boundary).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Cluster dynamics simulation of xenon diffusion during irradiation in UO 2

Diffusion of fission gas in UO 2 nuclear fuel impacts several important performance metrics, such as fission gas release, swelling, and thermal conductivity. Current empirical models of fission gas release have significant uncertainty, some of which derives from the bulk diffusion rate and its dependence on, for example, fuel chemistry and irradiation. In this work, we have applied the previously-developed Free Energy Cluster Dynamics (FECD) methodology in the code Centipede to calculate xenon cluster concentrations in UO 2 under intrinsic (high temperature) and irradiation-enhanced (intermediate temperature) conditions in order to develop a model of the xenon diffusion coefficient based on the atomic scale mechanisms responsible for transport. While the diffusion mechanism for xenon in UO 2 is adequately described by the Xe + U 2 O vacancy cluster for intrinsic conditions, a similar process is not capable of capturing measured in-pile fission gas diffusivity at intermediate temperatures. Therefore, a different diffusion mechanism must dominate under this regime. Using calculated atomistic data, we have shown that irradiation-enhanced diffusion at intermediate temperatures occurs via the larger Xe + U 4 O y vacancy clusters, which have lower migration barriers and increase in concentration by several orders of magnitude compared to intrinsic conditions. This mechanism is enabled by the increased uranium vacancy concentration under irradiation due to Frenkel pair production. In addition, the fast migration of uranium interstitials with two attached oxygen interstitials lowers the total uranium interstitial concentration through reactions with sinks. This allows the extended defects, such as Xe + U 4 O y vacancy clusters, to maintain high concentrations by limiting annihilation with attached vacancies. Furthermore, predictions using the Xe + U 4 O y diffusion mechanism are in good agreement with experiment, albeit with some differences in the Arrhenius slope, which we believe may be related to either experimental or model parameter uncertainty. Lastly, an analytical expression suitable for application in fuel performance simulations was derived to capture the predictions of the Centipede simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Application of BISON to UO 2 MiniFuel fission gas release analysis

There has been a recent push to accelerate fuel qualification by developing revolutionary capabilities to reduce irradiation periods, and thereby, reduce the time required to qualify a new fuel system. One such capability is the MiniFuel irradiation capsule designed to miniaturize fuel samples and irradiate “mini” fuel samples under isothermal temperature conditions. MiniFuel allows steady-state irradiations to decouple the traditionally coupled fission rate (i.e., power) and temperature parameters to understand and generate microstructures observed in fuel operated in a commercial reactor. Furthermore, this process offers the possibility to gather in situ data as well as postirradiation or transient data such as thermal conductivity, specific heat, fission gas diffusion and release, etc. However, accelerating fuel qualification is not solely reliant on generating large amounts of data but also on developing an informed test matrix designed to rapidly generate impactful data. Additionally, this process is reliant on fuel performance codes, such as BISON, to evaluate MiniFuel irradiations using existing material models. This process pinpoints model/data gaps, identifies desired irradiation conditions, and subsequently supports model validation and development. This work describes the use of BISON to perform a number of sensitivity studies designed to understand conditions that lead to fission gas release (FGR) under steady-state isothermal irradiation conditions and temperature transient conditions. The model is applied to a UO 2 MiniFuel example and shows an overall good qualitative agreement with experimental FGR annealing tests under different temperature conditions. It also accounts well for microstructural effects on FGR. When quantitatively compared with FGR data from previously irradiated 103 MWd/kgU UO 2 discs under thermal annealing, the model shows a less satisfactory agreement with the experimental data. Finally, a UO 2 MiniFuel test matrix is proposed to help to extend the model's operational range and validate the new FGR model capabilities to higher burnups and transient conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Synthesis of U 3 O 8 and UO 2 microspheres using microfluidics

Uranium-bearing microspheres below 50µm with a narrow size distribution allows for a wider variety of fuel forms. To accommodate the smaller size, gel microspheres with a composition of UO 3 ∙nH 2 O∙mNH 3 were synthesized using microfluidics and subsequently converted to U 3 O 8 and UO 2 . To accommodate the slower flow rates required by microfluidics, a more stable broth was established. Additionally, the gelation studies resulted in a broth that was stable for more than two days at 0°C and for close to 3 h at room temperature while still gelling within 25 s. Synthesis of gel microspheres with a narrow size distribution lasted for 5 h and produced ~0.5 g of air-dried material. The gelled microspheres were converted to U 3 O 8 and UO 2 and with sizes of 50 and 40 µm in diameter, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Diffusion in undoped and Cr-doped amorphous UO 2

UO 2 fuel pellets are often doped with chromium oxide to obtain favourable properties such as higher density, improved thermal stability, large grain sizes, improved pellet-clad interaction margins, and increased fission gas retention during transients. Chromium has a low solubility limit in UO 2 , with past experimental work reporting solubility limits ranging between 0.004 to 0.06 wt.% Cr. Due to its low solubility, segregation of Cr ions to the grain boundary may occur. Further, the complexity of these boundaries may be high as observed in other ceramics resulting in disordered or amorphous regions along the boundary, affecting a range of material and operational properties of the fuel pellet. To assess these disordered regions, in this work we study amorphous undoped and Cr doped UO 2 systems (containing 10–50 at.% Cr 3+ ) that have been modelled using classical molecular dynamics methods incorporating Cr 3+ into the well-used CRG potential library. Diffusion coefficients, pre-exponential factors, and activation energies for diffusion were computed for oxygen ions, assessing the impact of structure and extrinsic species on migration. Oxygen diffusion was observed to be much faster in the undoped amorphous system compared to its crystalline counterpart. Oxygen diffusion in doped systems decreased with increasing Cr concentration, highlighting the importance of additives to retain fission products and other migratory species.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

The influence of iron incorporation on the microstructure and crystal chemistry of UO 2 fuel

The incorporation of dopants into UO 2 is of interest for both doped advanced fuel and nuclear forensics research. Suitable forensics dopants, or taggants, for nuclear fuels are ones that do not adversely affect the performance of the parent material while still being traceable during key stages of the fuel’s life cycle. In this paper, we present our efforts to understand the effects of doping UO 2 with Fe 2 O 3 under multiple parameters, such as Fe concentration (0–3000 wppm), sintering temperature (1673 K, 1773 K, and 1973 K), sintering time (2–12 h), and atmosphere (reducing and inert). Here, we investigated densification behavior, microstructure evolution, and lattice defects based on the geometric densities, scanning electron microscopy (SEM), X-ray diffractograms, and Raman spectra. The effect of Fe on the microstructure and crystal chemistry of UO 2 is very complex, especially at the intermediate temperature (1773 K), where we found that densification kinetics are enhanced and that lattice contraction occurs for doping concentrations greater than 1000 wppm Fe.

Crystallography↗
DOE OSTI2022

Neutron capture of UO 2 targets prepared by spin-coating assisted combustion synthesis

Two uranium dioxide (UO 2 ) targets of (414 ± 23) nm and (1092 ± 93) nm thicknesses were prepared on 6061 aluminum alloy and puratronic grade aluminum backing materials. The targets were deposited with a novel method combining spin coating and solution combustion synthesis (SCS). The target layers consisted of small (3–7 nm) UO 2 grains and uniformly distributed ultra-small (1–3 nm) pores. The prepared targets were tested at the Los Alamos National Laboratory’s LANSCE facility for neutron irradiation damage and suitability for neutron capture experiments. The samples showed no signs of target material loss after the irradiation. However, irradiation caused a significant increase in the grain size (4–10 nm), as well as upward mass diffusion and coalescence of the pores due to the thermal spikes. The magnesium in the aluminum 6061 alloy backing also diffused into the UO 2 layer during neutron irradiation. The structural changes in the target after the irradiation do not affect the data from neutron capture. As a result, the new method can be used more broadly to prepare other actinide targets for nuclear physics experiments.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Grain subdivision and structural modifications by high-energy heavy ions in UO 2 with different initial grain size

Irradiation with 84 MeV Xe ions at peak fluence above 7 × 10 17 ions/cm 2 has been carried out on two UO 2 samples with different initial grain sizes at 300 °C and 500 °C. Characterization of both grain structure and bubble morphology has been performed using transmission electron microscopy, image analysis and precession electron diffraction. The bubble size in the large-grained sample was smaller than that in the standard-grain sample, confirming fission gas precipitation resistance in the former. The ion irradiation in the high electronic energy deposition range has caused formation of a large density of dislocations which resulted in grain polygonization for the sample with standard grain size, which was observed in multiple lamellas extracted next to each other at peak dose. The sample with large grains showed inconsistent behavior. In one case, dislocation pile-up occurred, but no sub-grain formation was observed. An adjacent location showed sub-grain formation, even if to less extent compared to the standard grain-size sample. Furthermore, these findings suggest a limited polygonization resistance in large-grained UO 2 compared to standard-grain UO 2 . Recrystallization of both sample surfaces was present, leaving unclear if this is a surface effect or a consequence of the highest energy deposition rate in this area.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

UO 2 -liquid metal suspension fuel concept for enhanced passive safety of LWRs: A heat pipe case study

This work proposes a novel fuel concept for use in light-water type reactors (LWRs): UO 2 particles suspended in a low-melting-point liquid metal (LM). This new fuel form offers lower energy density and higher thermal conductivity, thus enhancing the fuel’s thermal performance compared to sintered UO 2 pellets. Two candidate alloys for the LM are proposed based on previous studies into accident-tolerant fuels: Bi-Pb-Sn and Pb-Sn. Ten research questions are provided to guide future development of this fuel. A simple heat pipe model was constructed by coupling a reactor physics code to a heat transport code and other physics modules. Scoping calculations were performed on this model core as a preliminary investigation into five of the research questions. A fuel with 13 wt% UO 2 was shown to maintain a steady-state peak coolant channel wall temperature below 400 °C, and maintain criticality via self-regulating reactivity feedback. Shortcomings of the present model are discussed in this work, along with proposed future model improvements and fuel development. The appendix provides an example of a reactor concept that could be developed to utilize this fuel, but more modeling and experimental work are needed before a full reactor concept can be developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Pyrrophens: Pyrrole-Based Hexadentate Ligands Tailor-Made for Uranyl (UO 2 2+ ) Coordination and Molecular Recognition

Derivatives of a novel pyrrole-containing Schiff base ligand system (called “pyrrophen”) are presented which feature substituted phenylene linkers (R 1 = R 2 = H (H 2 L 1 ); R 1 = R 2 = CH 3 (H 2 L 2 )) and a binding pocket modeled after macrocyclic species. These ligands bind neutral CH 3 OH in the solid state through pyrrolic hydrogen-bonding. Here, the interaction of the uranyl cation (UO 2 2+ ) and H 2 L 1–2 yields planar hexagonal bipyramdial uranyl complexes, while the Cu 2+ and Zn 2+ complexes were found to self-assemble as dinuclear helicate complexes (M 2 L 2 ) with H 2 L 1 under identical conditions. The favorable binding of UO 2 2+ over Zn 2+ provides insight into the molecular recognition of uranyl over other metal species. Structural features of these complexes are examined with special attention to features of the UO 2 2+ coordination environment which distinguish them from other related salophen and porphyrinoid complexes.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Computationally Guided Investigation of the Optical Spectra of Pure β-UO 3

Single-phase β-UO 3 is synthesized by flash heating UO 2 (NO 3 )·6H 2 O in air to 450 °C and annealing for 60 h under the same conditions. For the first time, we report the Raman spectra of pure β-UO 3 . To facilitate the assignment of Raman and infrared vibrational modes, we use density functional theory with density functional perturbation theory. By employing a novel analysis scheme that includes the mode frequencies as well as a quantitative analysis of the mode eigenvectors, we assign the observed spectral features to individual chemical modes. In particular, the density functional theory optimized structure, observed Raman spectrum, and eigenvector analysis suggest the presence of four crystallographically distinct uranyl ions, one more than has previously been suggested.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Dehydration of UO 2 Cl 2 ·3H 2 O and Nd(NO 3 ) 3 ·6H 2 O with a Soft Donor Ligand and Comparison of Their Interactions through X-ray Diffraction and Theoretical Investigation [plus Supplemental Information]

Herein we investigated whether the relatively Lewis basic imidazole-2-thiones could be used to substitute water ligands bound to f-element cations and generate f-element soft donor complexes. Reactions of 1,3-diethylimidazole-2-thione (C 2 C 2 ImT) with Nd(NO 3 ) 3 ·6H 2 O and UO 2 Cl 2 ·3H 2 O led to the isolation of the anhydrous thione complexes Nd(NO 3 ) 3 (C 2 C 2 ImT) 3 and UO 2 Cl 2 (C 2 C 2 ImT) 2 , characterized by single crystal X-ray diffraction. Differences in the strength of metal–thione interactions have been examined by means of the crystal structure analysis and density functional theory (DFT) calculations. The C 2 C 2 ImT ligands were found to be affected by both coordination and noncovalent interactions, making it impossible to deconvolute the effects of one from the other. Calculated partial atomic charges indicated greater ligand-to-metal charge transfer in the [UO 2 ] 2+ complex, indicative of a stronger interaction. The reactivity of C 2 C 2 ImT demonstrates its usefulness in the preparation of f-element soft donor complexes from readily available hydrates that could be useful intermediates for promoting the coordination and studying the effects of soft donor anions.

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

Coupled Cluster Study of the Heats of Formation of UF 6 and the Uranium Oxyhalides, UO 2 X 2 (X = F, Cl, Br, I, and At)

The atomization enthalpies of the U(VI) species UF6 and the uranium oxyhalides UO 2 X 2 (X=F, Cl, Br, I, At) were calculated using a composite relativistic Feller-Peterson-Dixon (FPD) approach based on scalar relativistic DKH3-CCSD(T) with extrapolations to the CBS limit. The inherent multideterminantal nature of the U atom was mitigated by utilizing the singly charged atomic cation in all calculations with correction back to the neutral asymptote via the accurate ionization energy of the U atom. The effects of SO coupling were recovered using full 4-component CCSD(T) with contributions due to the Gaunt Hamiltonian calculated using Dirac-Hartree-Fock. The final atomization enthalpy for UF 6 (752.2 kcal/mol) was within 2.5 kcal/mol of the experimental value, but unfortunately the latter carries a ±2.4 kcal/mol uncertainty that is predominantly due to the experimental uncertainty in the formation enthalpy of U atom. The analogous value for UO 2 F 2 (607.6 kcal/mol) was in nearly exact agreement with experiment, but the latter has a stated experimental uncertainty of ±4.3 kcal/mol. The FPD atomization enthalpy for UO 2 Cl 2 (540.4 kcal/mol) was within the experimental error limits of ±5.5 kcal/mol. FPD atomization energies for the non-U-containing molecules (used for reaction enthalpies) H 2 O and HX (X=F, Cl, Br, I, At) were within at most 0.3 kcal/mol of their experimental values where available. The FPD atomization enthalpies, together with FPD reaction enthalpies for two different reactions, were used to determine heats of formation for all species of this work with estimated uncertainties of ±4 kcal/mol. The calculated heat of formation for UF 6 (-511.0 kcal/mol) is within 2.5 kcal/mol of the accurately-known (±0.45 kcal/mol) experimental value.

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

Cation-π Bonding in Actinides: UO x + (Benzene) ( x = 0, 1, 2) Complexes Studied with Threshold Photodissociation Spectroscopy and Theory

Cation-π complexes of the form UO x + (benzene) (x = 0, 1, 2) are produced by laser vaporization and cooled in a supersonic molecular beam. These ions are mass selected and studied with UV–visible laser photodissociation spectroscopy. Each of these complexes photodissociates by elimination of the benzene ligand. Above an energetic threshold, the absorption and photodissociation are continuous, indicating a high density of strongly coupled electronic states. The thresholds for the dissociation of each of these three complexes are measured and assigned as their respective bond dissociation energies. The bond energies determined [U + –(benzene): 42.5 ± 0.3 kcal/mol; UO + –(benzene): 41.0 ± 0.3 kcal/mol; UO 2 + –(benzene): 39.7 ± 0.3 kcal/mol] are comparable to those of transition metal ion-benzene complexes. Computational studies at the DFT/B3LYP level complement the experiments, predicting dissociation energies in reasonably good agreement with the experiments. Experiments and theory agree that the U+(benzene) complex is more strongly bound than its corresponding oxide ions. This new thermochemistry on actinide cation-π bonding should stimulate higher-level computational studies on these systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗