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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

A guided ion beam investigation of UO 2 + thermodynamics and f orbital participation: Reactions of U + + CO 2 , UO + + O 2 , and UO + + CO

A guided ion beam tandem mass spectrometer was employed to study the reactions of U + + CO 2 , UO + + O 2 , and the reverse of the former, UO + + CO. Reaction cross sections as a function of kinetic energy over about a three order of magnitude range were studied for all systems. The reaction of U + + CO 2 proceeds to form UO + + CO with an efficiency of 118% ± 24% as well as generating UO 2 + + C and UCO + + O. The reaction of UO + + O 2 forms UO 2 + in an exothermic, barrierless process and also results in the collision-induced dissociation of UO + to yield U + . In the UO + + CO reaction, the formation of UO 2 + in an endothermic process is the dominant reaction, but minor products of UCO + + O and U + + (O + CO) are also observed. Analysis of the kinetic energy dependences observed provides the bond energies, D 0 (U + –O) = 7.98 ± 0.22 and 8.05 ± 0.14 eV, D 0 (U + –CO) = 0.73 ± 0.13 eV, and D 0 (OU + –O) = 7.56 ± 0.12 eV. The values obtained for D 0 (U + –O) and D 0 (OU + –O) agree well with the previously reported literature values. To our knowledge, this is the first experimental measurement of D 0 (U + –CO). Furthermore, an analysis of the oxide bond energies shows that participation of 5f orbitals leads to a substantial increase in the thermodynamic stability of UO 2 + relative to ThO 2 + and especially transition metal dioxide cations.

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

Characterization of Uranyl Coordinated by Equatorial Oxygen: Oxo in UO 3 versus Oxyl in UO 3 +

Uranium trioxide, UO 3 , has a T-shaped structure with bent uranyl, UO 2 2+ , coordinated by an equatorial oxo, O 2– . The structure of cation UO 3 + is similar but with an equatorial oxyl, O •– . Neutral and cationic uranium trioxide coordinated by nitrates were characterized by collision induced dissociation (CID), infrared multiple-photon dissociation (IRMPD) spectroscopy, and density functional theory. CID of uranyl nitrate, [UO 2 (NO 3 ) 3 ] – (complex A1), eliminates NO 2 to produce nitrate-coordinated UO 3 + , [UO 2 (O • )(NO 3 ) 2 ]– (B1), which ejects NO 3 to yield UO 3 in [UO 2 (O)(NO 3 )] – (C1). Finally, C1 associates with H 2 O to afford uranyl hydroxide in [UO 2 (OH) 2 (NO 3 )] – (D1). IRMPD of B1, C1, and D1 confirms uranyl equatorially coordinated by nitrate(s) along with the following ligands: (B1) radical oxyl O •– ; (C1) oxo O 2– ; and (D1) two hydroxyls, OH – . As the nitrates are bidentate, the equatorial coordination is six in A1, five in B1, four in D1, and three in C1. Ligand congestion in low-coordinate C1 suggests orbital-directed bonding. Hydrolysis of the equatorial oxo in C1 epitomizes the inverse trans influence in UO 3 , which is uranyl with inert axial oxos and a reactive equatorial oxo. The uranyl ν 3 IR frequencies indicate the following donor ordering: O 2– [best donor] >> O •– > OH – > NO 3 – .

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

Equatorial Electronic Structure in the Uranyl Ion: Cs 2 UO 2 Cl 4 and Cs 2 UO 2 Br 4

Electric field gradient (EFG) tensors in the equatorial plane of the linear UO 2 2+ ion have been measured by nuclear magnetic resonance (NMR) and nuclear quadrupole resonance (NQR) experiments and computed by relativistic Kohn–Sham methods with and without environment embedding for Cs 2 UO 2 Cl 4 and Cs 2 UO 2 Br 4 . This approach expands the possibilities for probing the electronic structure in uranyl complexes beyond the strongly covalent U–O bonds. The combined analyses find that one of the two largest principal EFG tensor components at the halogen sites points along the U–X bond (X = Cl, Br), and the second is parallel to the UO 2 2+ ion; in Cs 2 UO 2 Cl 4 , the components are nearly equal in magnitude, whereas in Cs 2 UO 2 Br 4 , due to short-range bromide–cesium interactions, the equatorial component is dominant for one pair of Br sites and the axial component is larger for the second pair. Additionally, the directions and relative magnitudes of the field gradient principal axes are found to be sensitive to the σ and π electron donation by the ligands and the model of the environment. Chlorine-35 NQR spectra of 235 U-depleted and 235 U-enriched Cs 2 UO 2 Cl 4 exhibited no uranium-isotope-dependent shift, but the resonance of the depleted sample displayed a 58% broader line width.

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

Fabrication and thermophysical properties of UO 2 -UB 2 and UO 2 -UB 4 composites sintered via spark plasma sintering

Uranium dioxide (UO 2 ) composites with uranium diboride (UB 2 ) and uranium tetraboride (UB 4 ) have been proposed as advanced fuel candidates due to their high thermal conductivity, high melting point, high fissile density and their ability to incorporate a built-in burnable poison by tailoring the targeted 10 B/ 11 B ratio. As such, it is important to assess the fabrication, and thermal and micromechanical properties of such composites. In this work, UO 2 -UB 2 and UO 2 -UB 4 samples with boride phase fractions of 5, 15 and 30 wt% were fabricated to high densities (above 95 % theoretical density) via spark plasma sintering (SPS). This enabled sintering at relatively low temperatures and short timescales. SPS also aided in maintaining the target phase fractions of the samples as reactions between the constituent phases were suppressed due to the short timescales and reducing environment during sintering. Here, thermal diffusivity measurements from 299 to 1273 K were conducted through laser flash analysis (LFA). The diffusivity increased as a function of boride weight fraction, and UB 2 additions increased the thermal diffusivity of the composites more than UB 4 additions. Assessment of the LFA results indicated in-situ reactions between the UO 2 and boride phases that suppress the thermal diffusivity occur above 800 K for all samples. Oxidation of the boride phase was proposed as the underlying reaction. This was supported by thermodynamic assessments from the literature, as well as microstructural, crystallographic, and nanoindentation characterization performed on these samples.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Photodissociation and Infrared Spectroscopy of U + (CO 2 ) n , UO + (CO 2 ) n , and UO 2+ (CO 2 ) n Cation-Molecular Complexes

Laser vaporization of uranium in a pulsed supersonic expansion of carbon dioxide is used to produce complexes of the form U + (CO 2 ) n , UO + (CO 2 ) n , and UO 2+ (CO 2 ) n . These ions are selected in a reflectron time-offlight mass spectrometer and studied with visible laser photodissociation and tunable infrared laser photodissociation spectroscopy in the region of the CO 2 antisymmetric stretch. The dissociation patterns and spectroscopy of these ions indicate that CO 2 ligands are intact molecules. Although reaction products that form oxide-carbonyl or oxalate species are predicted to be stable, there is no direct evidence in the frequency range studied for the formation of these species. There is no clear indication for the coordination numbers for singly charged uranium and its oxide complexes with CO 2 . However, there is strong support in the vibrational patterns for an eight-coordinate complex of the doubly charged UO 2+ species, i.e., UO 2+ (CO 2 ) 8 .

Cluster chemistry↗
DOE OSTI2020

Structural and Optical Properties of Phase-Pure UO 2 , α-U 3 O 8 , and α-UO 3 Epitaxial Thin Films Grown by Pulsed Laser Deposition

Fundamental understanding of the electronic, chemical, and structural properties of uranium oxides requires the synthesis of high-crystalline-quality epitaxial films of different polymorphs of one material or different phases with various oxygen valence states. Here we report the growth of single-phase epitaxial UO 2 , α-U 3 O 8 , and α-UO 3 thin films using pulsed laser deposition. Both oxygen partial pressure and substrate temperature play critical roles in determining the crystal structure of the uranium oxide films. X-ray diffraction and Raman spectroscopy demonstrate that the films are single phase with excellent crystallinity and epitaxially grown on a variety of substrates. Chemical valance states and optical properties of epitaxial uranium oxide films are studied by X-ray photoelectron spectroscopy and UV–vis spectroscopy, which further confirm the high-quality stoichiometric phase-pure uranium oxide thin films. Epitaxial UO2 films show a direct band gap of 2.61 eV, while epitaxial α-UO 2 , α-U 3 O 8 and α-UO 3 films exhibit indirect band gaps of 1.89 and 2.26 eV, respectively. The ability to grow high-quality epitaxy actinide oxide thin films and to access their different phases and polymorphous will have significant benefits to the future applications in nuclear science and technology.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Heterometallic UO 2 2+ /Ag + Complexes: Structural Design and Luminescence Properties

Reported here are the synthesis, structural characterization, and luminescence properties of 11 novel UO 2 2+ /Ag + heterometallic complexes. Halogenated benzoic acids (2,6-dihalobenzoic acid (halo = F, Br), 3,5-dichlorobenzoic acid, and 3-halobenzoic acid (halo = Br, I)) and N-donor polycyclic ligands (2,2′-bipyridine, 2,2’;6′,2″-terpyridine, 1,10-phenanthroline, 2,2′-bipyrimidine) were employed to synthesize a set of compounds and induce structural diversity. The primary mode of coordination with the uranyl cation is hexagonal bipyramidal monomeric units with three halobenzoate ligands in the equatorial plane, though 1-D chains with pentagonal bipyramidal uranyl centers also form. The Ag + cations coordinate preferentially to the N-donor ligands and serve as counter-cations for the anionic uranyl motifs. The soft ligand character of the N-donor molecules is found to be a requirement for the inclusion of the Ag + cation into the structures. Anionic uranyl units and cationic silver units assemble via noncovalent interactions between π systems on adjacent rings and between halogens (when Br and I are present). Solid-state emission spectra display the usual uranyl band with superimposed vibronic fine structure, except for that of compound 1 , which shows emission from the 2,2′-bipyridine center. This family of compounds represents a substantial contribution to the already rich library of UO 2 2+ /Ag + compounds, and the synthetic parameters discussed within reveal a platform for the design of new heterometallic uranyl-containing materials.

anions↗
DOE OSTI2020

High density UO 2 and high thermal conductivity UO 2 composites by spark plasma sintering (SPS)

Embodiments of the invention are directed to a method for production of a nuclear fuel pellet by spark plasma sintering (SPS), wherein a fuel pellet with more than 80% TD or more than 90% TD is formed. The SPS can be performed with the imposition of a controlled uniaxial pressure applied at the maximum temperature of the processing to achieve a very high density, in excess of 95% TD, at temperatures of 850 to 1600° C. The formation of a fuel pellet can be carried out in one hour or less. In an embodiment of the invention, a nuclear fuel pellet comprises UO 2 and a highly thermally conductive material, such as SiC or diamond.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Demonstrate new plasticity models for doped UO 2 that capture dislocation mechanisms

In light water reactors, fuel vendors are investigating the use of dopants to modify the properties of UO 2 pellets, with the goal of improving pellet-cladding mechanical interactions during operation. Dopants are expected to ‘soften’ the pellets; that is, the doped pellets have higher plastic deformation than conventional UO 2 . This leads to a reduction in the severity of mechanical pellet-cladding interactions, helping to reduce the hoop strain on the cladding. By minimizing the strain exerted by the pellet on the cladding, it is anticipated that cladding performance under accident conditions can be enhanced (i.e., lowering the risk of burst during a LOCA). Dopants such as chromium (Cr) promote grain growth during pellet fabrication, leading to larger grains; therefore, understanding the link between chemistry, microstructure and mechanical deformation (enhanced creep rates) behavior of UO 2 is critical to helping operators further substantiate the benefits of doping UO 2 . Historically, the nuclear energy industry has relied on empirical models to make assessments of performance. Compared to empirical models, mechanistic physics-based models provide benefits, such as, fewer data points for validation and better extrapolation where experimental data is scarce or non-existent. In this report, Bayesian inference techniques have been applied to a previously developed lower length-scale-informed diffusional creep model. The objective is to i) infer lower-length-scale parameter distributions from available experiment and then ii) determine the uncertainties in the measurable quantity (in this case creep rates) after propagating the inferred lower length scale parameter uncertainties. The approach requires many evaluations of the model, which becomes computationally insurmountable; therefore, a neural-network model is trained to data obtained by sampling the full model over the most important parameters. This neural-network is then used in the Bayesian inference approach to determine probability distributions in the parameter values that represent the uncertainty in the model given what is known from the experiments (posterior). A significant reduction compared to conservative initial (prior) uncertainties is achieved through inference against the experimental data, demonstrating the efficacy of this approach. Furthermore, by accounting for uncertainties in the experimental conditions and sample non-stoichiometry, it is possible to resolve apparent discrepancies in experimental measurements within a self-consistent grain boundary (Coble) creep model that is sensitive to chemistry. This work has been written up and submitted to Nuclear Technology for a special issue on accelerated fuel qualification (AFQ). This uncertainty quantification (UQ) work not only improves the diffusional model, while accounting for uncertainty, but also establishes a framework which can readily be applied to the mechanistic models of dislocation deformation developed in this study. The most likely values from the Bayesian analysis are incorporated into our UO 2 diffusional creep model and a lower length scale-informed irradiation UO 2 creep mechanistic model to generate a dataset. This dataset has been provided to our INL collaborators for training an artificial neural network surrogate model, which will be implemented in the BISON fuel performance code to assess how the results differ from those currently obtained using a fully empirical model and that of using the nominal (uncalibrated) atomic scale parameters in our mechanistic model. Plastic deformation (creep and glide) in UO 2 is a complex phenomenon, governed by multiple underlying processes such as local defect concentrations, applied stresses, and microstructural characteristics. Consequently, there is a need for a meso-scale model with polycrystalline resolution capable of extrapolating to large grain sizes applicable to doped UO 2 , where data is limited and the model can help bridge the knowledge gap. By integrating atomistic data into the polycrystal LApx code, it becomes possible to predict dislocation climb and glide plasticity that simple analytical models cannot accurately represent. The application of atomic-scale data within LApx demonstrated the importance of climb and glide mechanisms in reproducing high-stress UO 2 behavior. Behaviors such as this are crucial to capture and implement in BISON, as parts of the fuel pellet can reach temperatures where glide can occur before pellet cracking. This model which captures dislocation based mechanisms for UO 2 is then used to stand up the doped model accounting for larger grain sizes. It was found that larger grain sizes can lead to enhanced deformation rates in the glide regime, and therefore can help with the pellet cladding mechanical interaction. Therefore if the fuel pellet reaches conditions (stress/temperature) where glide is active, the enhanced creep rates for larger grains in the glide regime (doped UO 2 ) can help with pellet cladding mechanical interactions. Plastic deformation in UO 2 involves multiple mechanisms, including diffusional creep, dislocation climb, and glide. This milestone contains two parts: (1) UQ of a pre-existing lower length scale informed mechanistic diffusional creep model, and (2) development of a new LApx based model for dislocation-mediated creep mechanisms in UO 2 , with application to large-grain doped UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Impact of uranium oxide (UO 2 ) fuel with molybdenum (Mo) inserts on pressurized water reactor performance and safety

This work investigates nuclear reactor performance and safety characteristics of UO 2 with high thermal conductivity Mo insert structures by using multiphysics modeling techniques. Additionally, the purpose of this study is to use scoping analyses to quantify the impact of using Mo inserts from neutronic and heat transfer standpoints. Attention is given to reactor performance parameters, such as cycle length, maximum fuel temperature, temperature gradients in the fuel, and stored energy in the fuel. The finite-element code BISON and the Monte Carlo particle transport code Serpent were used to perform sensitivity analyses on the Mo insert geometry to optimize the insert design and inform larger scale modeling that required the homogenization of the UO 2 and Mo. Although BISON is often used as a fuel performance analysis tool, it is used in this context for heat transfer analysis only. Fuel performance optimization is outside the scope of the current study, but would be important for future work focused on this concept. The results showed that the insert had little impact on neutronic performance and that homogenizing the UO 2 and Mo was acceptable for reactor physics calculations. Reactivity temperature coefficients calculated using homogeneous UO 2 -Mo were shown to be relatively similar to UO 2 , but higher Mo content and 235 U enrichment can reduce the worth of soluble boron and control rods. The effect of insert geometry on heat transfer was much greater, and an approximately 15–20% difference in maximum fuel temperature was predicted between the best and worst performing heat transfer geometries. Furthermore, thermal conductivity calibration based on the finite element analysis results was performed to improve the accuracy of temperature predictions in reactor analysis models that homogenized the UO 2 -Mo fuel. Compared with UO 2 in a pressurized water reactor (PWR), the optimized UO 2 -Mo design increased the margin to fuel melt by 13–32% across the fuel cycle, but it requires the 235U enrichment to exceed 5% to match the cycle length of conventional UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Statistical fracture behavior of doped UO 2 using a ball-on-ring equibiaxial flexure test method

Metal oxide dopants, such as titanium and chromium oxides, have garnered considerable attention for their potential to increase grain size (≥ 30 µm) in UO 2 fuel, purportedly enhancing fission gas retention during reactor operation. Fuel performance is significantly impacted by fuel fracture behavior, so it is important to understand the effects of enhanced grain size and dopant content on UO 2 fuel fracture. UO 2 pellets were doped with 0.1 wt% TiO 2 and 0.3 wt% Cr 2 O 3 to alter density and grain size. Inductively coupled plasma mass spectroscopy measured dopant levels pre- and post-sintering. X-ray diffraction revealed lattice changes and microstrain via Rietveld refinement. Field emission scanning electron microscopy determined grain sizes of approximately 30 µm for TiO 2 doping and 7 µm for Cr 2 O 3 doping. Transverse rupture strength tests were performed on over 30 samples per dataset to obtain characteristic strength and Weibull modulus. Results indicate no statistical difference in fracture strength between 0.1 wt% TiO 2 doped UO 2 and undoped UO 2 , while 0.3 wt% Cr 2 O 3 doped UO 2 exhibited a 20% decrease in fracture strength. Doped UO 2 samples also showed reduced Weibull modulus compared to undoped UO 2 , suggesting increased scatter in fracture strength. This study's findings suggest that titanium and chromium oxide doping in UO 2 , regardless of grain size, induce residual stresses, decreasing fracture strength and increasing variability in fracture behavior.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Benefits of using multiple Raman laser wavelengths for characterizing defects in a UO 2 matrix

Abstract Raman spectroscopy is one of the most useful techniques for studying the structure of UO 2 and changes due to specific defects caused by doping, changes in stoichiometry, irradiation, or heating under oxidizing conditions. In this paper, we illustrate several aspects of the application of Raman techniques to the study of UO 2 , including the use of wavelength‐dependent excitation (455, 532, and 785 nm) to assess the effects of doping (Nd, Th, and Zr), ion irradiation, and in situ heating and oxidation (UO 2 to U 3 O 8 ). Additionally, we show examples of how correlative microscopy is possible using electron backscatter diffraction combined with Raman maps of specific vibration bands or of laser‐induced luminescence generated by rare‐earth dopants in the matrix. For each of these applications, we suggest optimal excitation wavelengths that vary depending on the desired data. Blue (455 nm) excitation tends to promote oxidation even at low powers, but because Raman spectra change little with doping, irradiation‐induced changes are easier to observe. Green (532 nm) excitation is optimal for observing electron–phonon resonance effects in UO 2 and offers a good compromise for high‐temperature oxidation experiments, delivering high‐quality spectra for both UO 2 and U 3 O 8 . Infrared (785 nm) excitation is best for observing “defect” bands associated with doping in UO 2 , as changes with irradiation are small. Raman spectroscopy is particularly suited for studying the stability of UO 2 towards oxidation in the presence of dopants simulating fission products, where electron–phonon resonant effects, dopant ion luminescence, and mapping can be used together to investigate structural rearrangement as a function of temperature. These techniques can offer insight into microstructural changes in UO 2 fuels at higher burnups envisioned in future reactors.

42 ENGINEERING↗
DOE OSTI2020

UO 2 + 5 vol% ZrB 2 nano composite nuclear fuels with full boron retention and enhanced oxidation resistance

The boron isotope ( 10 B) can be used as a neutron absorber in UO 2 to control the reactivity of nuclear fuel pellets, however, the boron source can react with oxygen source in UO 2 to form B 2 O 3 that vaporize readily at temperatures above 1200 °C. Unfortunately, the sintering of UO 2 fuel requires hours holding at high temperature (>1700 °C), resulting in an inevitable B loss during sintering and unpredictable B concentration in final product. It is challenging to incorporate boron through a conventional sintering method. In this work, we demonstrated that spark plasma sintering (SPS), a field assisted sintering technology, can effectively densify UO 2 + 5 vol% ZrB 2 composite fuel pellets by rapid consolidation at 1600 °C for a short duration of 5 min under an applied pressure of 40 MPa. Further, thermogravimetric analysis (TGA) measurements confirm that ZrB 2 is fully retained inside the composite fuel pellets. Inside the composite fuel pellets, nano sized ZrB 2 particles are uniformly distributed along the grain boundaries of the UO 2 matrix. The ZrB 2 particle transforms to a glassy B 2 O 3 phase covering the sample surface and grain boundaries of UO 2 matrix after a simple post-sintering annealing at 1000 °C in flowing Argon gas for 4 h. The formed glassy B 2 O 3 slows down the diffusion of oxygen ions and postpones the onset temperature for oxidation of UO 2 from 400 °C to 550 °C. This study demonstrates the capability of SPS, an advanced fuel manufacturing technique, to achieve a full retention of ZrB 2 in UO 2 oxide fuel and increase oxidation resistance through a simple post-sintering annealing. The reported work holds great engineering potential for development of advanced oxide fuel for nuclear application.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Fission gas diffusion and release for Cr 2 O 3 -doped UO 2 : From the atomic to the engineering scale

Here, the anticipated benefits of large grains in Cr 2 O 3 -doped UO 2 pellets include improved mechanical and fission gas retention properties. To support the assessment of fission gas release (FGR) from doped pellets, the impact of doping on fission gas diffusivity for in-reactor conditions must be understood. In this work, we tackle this issue by informing the fission gas model within the BISON fuel performance code using material models developed at the atomic scale. The investigation of intra-granular fission gas diffusivity in Cr 2 O 3 -doped UO 2 is carried out by adapting a cluster dynamics model that, accounting for UO 2 thermochemistry, is capable of describing Xe diffusion under irradiation in undoped UO 2 as the starting point. Using a thermodynamic analysis, it is shown that in stoichiometric UO 2 with additions of Cr 2 O 3 the oxygen potential is defined by the Cr-Cr 2 O 3 two-phase equilibrium. Using the cluster dynamics model, the predicted Xe diffusivity in doped UO 2 was significantly increased in both the intrinsic and irradiation-enhanced regimes compared to undoped UO 2 as a result of higher concentrations of uranium and oxygen vacancies, respectively. This is a consequence of the more oxidizing conditions at high temperature, and more reducing conditions at low temperature, as a result of doping. Arrhenius functions have been fitted to the cluster dynamics results to enable implementation of the new diffusivities in the BISON fission gas behavior model. BISON simulations were carried out, showing the competing effects of the enlarged grains and the new fission gas diffusivity model, which act to suppress and enhance fission gas release, respectively. The new physics-informed model was validated against in-reactor experimental measurements under normal operation. Additionally, benchmarking was carried out for power ramp conditions. The predicted fission gas release agreed well with the experimental data, showing noticeable improvements over the standard UO 2 model.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Optical vibrational spectra and proposed crystal structure of ε-UO 3

ε-UO 3 is an exotic polymorph in the uranium trioxide system with an undetermined crystal structure and limited optical vibrational spectroscopic data. To improve understanding of this compound, we synthesize and investigate the crystal structure and optical vibrational spectra of ε-UO 3 . Infrared spectra collected for ε-UO 3 are in good agreement with previously published results, and our studies extend the available data into the low-energy (600–100 cm –1 ) regime. For the first time, Raman spectra are presented for ε-UO 3 using both 785 and 532 nm excitation wavelengths. Previous reports suggest an impurity phase may be present in ε-UO 3 produced by calcination of U 3 O 8 ; however, spectral center-of-mass calculations, principal component analyses, and Raman spectroscopic mapping employed to investigate this possibility indicate that the product of U 3 O 8 calcined in O 3 (g) in this work is likely phase-pure. A possible novel structure solution for ε-UO 3 is determined via Rietveld refinement of powder X-ray diffraction data and is triclinic, P-1, with a = 4.01 Å, b = 3.85 Å, c = 4.18 Å, and α = 98.26°, β = 90.41°, γ = 120.46° (R wp = 8.30%). The asymmetric unit of ε-UO 3 consists of U(VI) in hexagonal bipyramidal coordination with displaced equatorial oxygen. Further analysis reveals that the structure of ε-UO3 is best described by a 2 × 1 × 2 supercell structure in P-1 with a = 8.03 Å, b = 3.86 Å, c = 8.37 Å with α = 98.26°, β = 90.41°, and γ = 120.46°, although a higher-symmetry structure is possible. Optical vibrational spectroscopic and structural measurements of ε-UO 3 presented here furthers our understanding of this complex uranium oxide and clarifies the origin of reported structural similarity to U 3 O 8 .

-UO3↗
DOE OSTI2024

Fracture behavior of irradiation induced nanocrystalline UO 2 studied by in-situ mechanical testing in transmission electron microscopy

Uranium Dioxide (UO 2 ) is widely used as a fuel in current light water reactors (LWRs). Upon accumulation of radiation damage, LWR UO 2 fuel pellets start to develop a different microstructure at the pellet periphery when fuel burnup exceeds 45–50 GWd/tHM. The resulting porous, nanocrystalline microstructure is one of the most prominent microstructural changes occurring in such fuel. Its fracture mechanisms, which causes fuel fine fragmentation, could impact safety limits when the cladding breaches. Direct measurements of these properties are challenging, therefore a surrogate obtained via ion irradiation can be used. In this study, multiple microcantilevers were fabricated by focused ion beam from both fresh UO 2 and UO 2 irradiated with 84 MeV Xe 26+ ions to a peak dose of 1357 displacements per atom (dpa). Further, the irradiation produced a pseudo high burnup structure approximately 2 µm below the surface. In-situ nano-mechanical bending tests were conducted to investigate the fracture behavior and the effect of the surrogate UO 2 high burnup structure on local fracture properties. Fresh UO 2 fuel was observed to fracture in transgranular mode without nucleation or movement of dislocations. However, the Xe-irradiated nanocrystalline microcantilevers fractured along the grain boundaries, with no influence from the pre-existing micro-cracks in the microcantilever. Fracture toughness for this type of surrogate high burnup UO 2 structure is reported for the first time in literature. Both the fracture stress and toughness show degradation for UO 2 as a result of Xe-irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

NOx mediated formation of ε – UO 3 during thermal decomposition of uranyl nitrate hexahydrate under static air conditions

The thermal decomposition of uranyl nitrate hexahydrate in air at temperatures of 650 – 800 °C is expected to yield α – U 3 O 8 through intermediate UO 3 phases. Here, n this study, Raman spectroscopy complimented by powder X-ray diffraction revealed the unexpected formation of ε – UO 3 as an accompanying phase during thermal decomposition of UNH at 700 °C under static air conditions. Experiments demonstrated that ε – UO 3 does not form during the initial ramp up stage nor during the high temperature heating period; but instead forms during the cooling stage between 250 – 400 °C. Additional experiments revealed that UNH initially decomposes through an amorphous UO 3 intermediate prior to α – U 3 O 8 formation. Attempts to bypass the U 3 O 8 precursor in the formation of ε – UO 3 were unsuccessful, supporting the necessity of U 3 O 8 (α – U 3 O 8 in this study) in the formation of ε – UO 3 . The observed phase evolution is proposed to result from the NOx species generated during UNH thermal decomposition, which create localized oxidizing conditions within the furnace under static air conditions. Furthermore, a predominantly phase pure ε – UO 3 was synthesized under static air conditions through the addition of an extra plateau at 250 °C during the cooling step. These findings demonstrate the importance of gas-phase chemistry and cooling conditions in uranium oxide phase evolution and provide additional insight into ε – UO 3 formation pathways.

Epsilon uranium trioxide↗
DOE OSTI2026

Hygroscopic growth of UO 2 F 2 nanoparticles

Hygroscopicity is an important physicochemical property of aerosol that describes the ability of a particle to uptake water. The hygroscopic properties of uranyl fluoride (UO 2 F 2 ) aerosol generated from a UF 6 hydrolysis reactor was investigated for the first time using a custom-built Humidified Tandem Differential Mobility Analyzer (HTDMA). The HTDMA is capable of measuring UO 2 F 2 nanoparticle growth determined by mobility size over a wide range of atmospheric humidity from dry conditions at <10% relative humidity (RH) to 85% RH. The hygroscopic properties were determined for nanoparticles as small as 3.5 nm in this study. Although the largest size of UO 2 F 2 nanoparticles was 80 nm, monodisperse aerosol with a mobility diameter of up to approximately 500 nm can be investigated using the HTDMA. Anhydrous UO 2 F 2 nanoparticles with a mobility diameter of 3.5 nm were shown to be highly hygroscopic with a deliquescence relative humidity (DRH) of 10%. Hydrates with a larger mobility diameter from 10 to 80 nm were non-hygroscopic with no observable DRH and limited water uptake up to 85% RH. Here, these results demonstrate the hygroscopic properties of UO 2 F 2 nanoparticles are highly variable and based on both the mobility size and hydration state. Hygroscopicity affects the physicochemical properties of UO 2 F 2 nanoparticles, including the aerosol phase state and viscosity, with impacts on aerosol growth, coagulation, and deposition that is critical for understanding the fate and transport of UO 2 F 2 nanoparticles in the atmosphere.

Hygroscopicity↗