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The neutralization of acidic solutions containing U (IV) and Ce (III) at room temperature in glove box atmosphere and in the presence of dithionite results in coprecipitation of these elements as amorphous solid solutions Ce x U 1– x O 2± y . The solubilities of the precipitates with different mole fractions ( x ) of Ce(OH) 3 ( x = 0.01 or 0.1) were determined in 1  M NaClO 4 solutions between pH 2.2 and 12.8 under reducing conditions. The solids were investigated by a variety of methods (chemical analysis, SEM-EDX, XRD, XPS, XAS) to determine the nature of the solid solutions formed, their composition and the valence state of Ce and U. X-ray photoelectron spectroscopy confirmed the oxidation states of the solids both before and after the equilibration as Ce (III) and U (IV). The amorphous coprecipitates reached equilibrium relatively fast (∼1 week). The release of Ce from the coprecipitates was totally dominated by the release of uranium over the whole pH range. The Ce concentrations decrease slightly with the decrease of Ce content in the solid, suggesting that Ce x U 1– x O 2± y solids behave thermodynamically as solid solutions. The concentrations of U in equilibrium with the coprecipitate were in excellent agreement with the solubility of UO 2 (s) under reducing conditions reported in the literature. The conditional solubility product of Ce(OH) 3 from the coprecipitate was several orders of magnitude (∼4 in the near neutral pH range and ∼18 in the acidic range) lower than that of pure Ce(OH) 3 (s). The activities and activity coefficients of Ce(OH) 3 (s) in the coprecipitate were also estimated. Activity coefficients are much less than 1, indicating that the mixing of Ce(OH) 3 with UO 2 is highly favorable.

The neutralization of acidic solutions containing U (IV) and Ce (III) at room temperature in glove box atmosphere and in the presence of dithionite results in coprecipitation of these elements as amorphous solid solutions CexU1–xO2±y. The solubilities of the precipitates with different mole fractions (x) of Ce(OH)3 (x = 0.01 or 0.1) were determined in 1 M NaClO4 solutions between pH 2.2 and 12.8 under reducing conditions. The solids were investigated by a variety of methods (chemical analysis, SEM‐EDX, XRD, XPS, XAS) to determine the nature of the solid solutions formed, their composition and the valence state of Ce and U. X‐ray photoelectron spectroscopy confirmed the oxidation states of the solids both before and after the equilibration as Ce (III) and U (IV). The amorphous coprecipitates reached equilibrium relatively fast (∼1 week). The release of Ce from the coprecipitates was totally dominated by the release of uranium over the whole pH range. The Ce concentrations decrease slightly with the decrease of Ce content in the solid, suggesting that CexU1–xO2±y solids behave thermodynamically as solid solutions. The concentrations of U in equilibrium with the coprecipitate were in excellent agreement with the solubility of UO2(s) under reducing conditions reported in the literature. The conditional solubility product of Ce(OH)3 from the coprecipitate was several orders of magnitude (∼4 in the near neutral pH range and ∼18 in the acidic range) lower than that of pure Ce(OH)3(s). The activities and activity coefficients of Ce(OH)3(s) in the coprecipitate were also estimated. Activity coefficients are much less than 1, indicating that the mixing of Ce(OH)3 with UO2 is highly favorable.

To further the development of low-enriched uranium fuels, precedence has been placed on delivering the same amount of power while lowering the fuel temperature and radial temperature gradient. To address this, modeling efforts have resulted in a novel design featuring conductive fins of varying thermal conductivities and geometries inserted into the fuel matrix. These conductive inserts were not allowed to exceed 6% of the original fuel volume. This constraint was imposed due to other designs displacing 10% of fuel volume. A parametric study was performed that consisted of 2.56 million BISON simulations involving varying fin characteristics (i.e., fin thermal conductivity, number, and geometry) to determine the optimal geometric configuration for a desired amount of fuel volume displaced. The results from this study show that the thickness and length of each fin affect the fuel temperature and temperature gradient more than varying the number and thermal conductivity of the fins. The parametric study resulted in the development of an optimized combination to produce the lowest peak fuel temperature, lowest radial temperature gradient, and highest temperature reduction for the amount of original fuel volume displaced. The simulations presented in this work will eventually be compared with irradiation experiments of similar fuel designs at Idaho National Laboratory’s Advanced Test Reactor.

In this study, we investigated the behavior of xenon (Xe) bubbles in uranium dioxide (UO2) grain boundaries using molecular dynamics simulations and compared it to that in the UO2 bulk. The results show that the formation energy of Xe clusters at the Σ5 grain boundaries (GBs) is much lower than in the bulk. The diffusion activation energy of a single interstitial Xe atom at the GBs was approximately 1 eV lower than that in the bulk. Furthermore, the nucleation and growth of Xe bubbles in the Σ5 GBs at 1000 and 2000 K were simulated. The volume and pressure of bubbles with different numbers of Xe atoms were simulated. The bubble pressure dropped with increasing temperature at low Xe concentrations, whereas the volume increased. The radial distribution function was computed to explore the configuration evolution of Xe bubbles. The bubble structures in the GB and bulk material at the same temperature were also compared. Xe atoms were more regular in the bulk, whereas multiple Xe atoms formed a planar structure at the GBs.

In operando SEM imaging of electrochemical oxidation of uranium oxide (UO2) will be presented in this contribution. Specifically the electrochemical version of the System for Analysis at the Liquid Vacuum Interface (SALVI) was used to study the interaction of dissolved oxygen, hydrogen peroxide and hydrogen with UO2. The Quanta 250 FEG SEM housed in the Radiological Processing Laboratory (RPL) was used to for experiments of radiological materials. In operando results and the application of in situ SEM imaging in studying nuclear materials will be presented.

Xe and Kr gases produced during the use of uranium dioxide (UO2)-fuelled reactors can easily form bubbles, resulting in fuel swelling or performance degradation. Therefore, it is important to understand the influence of point defects on the behaviour of Xe and Kr gases in UO2. In this work, the effects of point defects on the behavioural characteristics of Xe/Kr clusters in UO2 have been systematically studied using molecular dynamics. The results show that Xe and Kr clusters occupy vacancies as nucleation points by squeezing U atoms out of the lattice, and the existence of vacancies makes the clusters more stable. The diffusion of interstitial Xe/Kr atoms and clusters in UO2 is also investigated. It is found that the activation energy is ~2 eV and that the diffusion of the interstitial atoms is very difficult. Xe and Kr bubbles form at high temperatures. The more interstitial Xe/Kr atoms or vacancies in the system, the easier the clusters form.

The anodic reactivity of UO2 and UO2 doped with Gd2O3 was investigated by electrochemical methods in slightly alkaline conditions in the presence of silicate and calcium. At the end of the experiments, the electrodes were analysed by X-ray photoelectron spectroscopy to determine the oxidation state of the uranium on the surface. The experiments showed that the increase in gadolinia doping level led to a reduction in the reactivity of UO2, this effect being more marked at the highest doping level studied (10 wt.% Gd2O3). This behaviour could be attributed to the formation of dopant-vacancy clusters (GdIII-Ov), which could limit the accommodation of excess O2− into the UO2 lattice. In addition, the presence of Ca2+ and SiO32− decreased the anodic dissolution of UO2. In summary, the Gd2O3 doping in presence of silicate and calcium was found to strongly decrease the oxidative dissolution of UO2, which is a beneficial situation regarding the long-term management of spent nuclear fuel in a repository.

To ensure the safety and efficient operation of nuclear reactors, it is imperative to understand the effects of various dopants (Ti, Th, and Zr) on the solubility of the fission product Xe in UO2. In this study, Hubbard corrected density functional theory (DFT + U) and occupation matrix control were used to investigate the bulk and defect properties of UO2. The results show that the UO2-Ti system is more favorable for Xe dissolution in vacancies, whereas the UO2-Th system has little effect on the dissolution of Xe atoms. Th, Zr, and Ti inhibit the aggregation of Xe clusters, and Ti is the least favorable for the nucleation and growth of Xe clusters.

A nucleation method based on a composite of uranium dioxide (UO2) and graphene is presented by in situ synthesis, and the relevant mechanism and fuel properties are investigated. UO2–graphene composite fuel powders containing graphene volume (2%, 4%, 6%, and 8%) were prepared using a nucleation method through the reactive deposition of uranyl nitrate and aqueous ammonia on graphene by controlling the reaction parameters. The composite fuel pellets were prepared using spark plasma sintering (SPS). The results showed that the uniformity of UO2–graphene powder prepared by in situ synthesis reached up to 96.39%. An analysis on the relevant phase structure showed that only UO2 and graphene existed in the sintered pellets at 1723 K, graphene and UO2 were not destroyed during the reaction, and the pellet densities for the in-situ synthesis were 95.56%TD, 95.32%TD, 95.08%TD, and 94.76%TD for graphene contents of 2%, 4%, 6%, and 8%, respectively. The thermal conductivities of pellets at 293 K increased by 12.27%, 20.13%, 27.47%, and 34.13%, and by 18.36%, 35.00%, 47.07%, and 58.93% at 1273 K for 2%, 4%, 6%, and 8% graphene contents, respectively. The performance of graphene in the fuel was superior at high temperatures, which overcame shortcomings due to the low thermal conductivity of UO2 at high temperatures. SEM results showed that the grain sizes of the pellets prepared by synthesis in situ were 10–30 μm, and there was no obvious pore at the grain boundary because the grains were closely bound. The graphene was uniformly coated by UO2, and the thermal conductivity of the pellets improved upon the formation of a bridging heat conduction network.

Sintering process is the final stage of fuel kernel manufacturing prior to the coating process. This stage is very important part of the whole process, because it will determine the feasibility of UO2 kernel to comply with the specifications of HTR reactor fuel. The objective of this research was to obtain UO2 kernel with the density of ≥ 95% TD. The results showed that the highest density reached 92.56% TD or about 10.1441 g/cm3. This condition of sintering was gained at the temperature of 1400 °C with sintering time of 2 hours. The sintering product diameter gained was around 919 μm the specific surface area 4.4213 m2/g, and a total pore volume 4,751 x 10−3 cm3/g. The density of UO2 kernel produced from this research is the best compared to previous finding because of its density already approaches the HTR fuel specification requirements.