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At TEPCO's Fukushima Daiichi Nuclear Power Station, it is estimated that considerable amounts of cesium still remain in the reactors from the analysis results using the severe accident analysis codes and the reverse analysis from contaminated water. Since cesium is known to form stable compounds with uranium and zirconium, chemisorption experiments with uranium dioxide pellets and iron-zirconium melts for cesium hydroxide vapor were carried out using the similar method as for stainless steel. As the results, formations of cesium uranate, Cs2UO4, and cesium zirconate, Cs2ZrO3, were confirmed, indicating that cesium was chemisorbed on both of the uranium dioxide pellets and the iron-zirconium melts in an Ar-H2-H2O flow and an Ar-H2 flow, respectively. Therefore, it was considered that cesium released from fuel might be trapped by chemisorption with fuels and/or iron-zirconium melts during light water reactor severe accidents. [ABSTRACT FROM AUTHOR]

Non-destructive analysis(NDA) techniques can be used to ensure that no sensitive nuclear materials(SNM) are being lost or diverted from their defined flows and containments. Rapid measurements of SNM in feed materials, process lines, finished products, scrap and waste, and holdup in the plant are all possible with NDA techniques. With the development of nuclear industry, it has also brought huge challenges to the supervision of nuclear security and nuclear material control. The NDA of pressurized water reactor fuel assemblies is one of the important technologies for nuclear security. The establishment of the NDA for pressurized water reactor requires the possession of corresponding nuclear material standard materials. Therefore, it is necessary to analyze the uranium content and 235U abundance accurately in the uranium dioxide pellets of standard pressurized water reactor components. China North Nuclear Fuel Corpotation prepared comparison samples and conducted a uniformity test on the samples. Through F test analysis, it was deemed that the samples had good homogeneity. Beijing Research Institute of Uranium Geology(BRIUG) organizes several laboratories to determine the content of uranium in uranium dioxide by potentiometric titration, and the abundance of 235U by TIMS or MC-ICP-MS methods. The samples are sent to different laboratories. Six laboratories conducted comparative analysis of uranium content and 235U abundance. After summarizing and comparing the results, through online SPSS calculations, the small sample data of uranium content and 235U abundance shows a p-value greater than 0.05, indicating no significant difference. This indicates that the parameter conforms to the normal distribution characteristics. The uncertainty of the comparison data results was evaluated, the uncertainty of sample measurement results refers to ISO guide 35 2017 reference materials: guidance for characterization and assessment of homogeneity and stability, which is mainly composed of three parts: stability uncertainty, homogeneity uncertainty and characterization uncertainty. The uranium content and abundance of 235U in the sample have good homogeneity, and its properties will not change with time. Moreover, the uncertainty caused by the inhomogeneity of the sample has been reflected in the comparison analysis, so in this calculation, the uncertainty components arising from sample inhomogeneities and instabilities are not considered. The average value of uranium mass fraction is 88.00%(n=6), the relative uncertainty is 0.057%, and the measured average value of 235U abundance is 3.2326%(n=6), the relative uncertainty is 0.062%. The uranium content and 235U abundance in standard pressurized water reactor components made of uranium dioxide pellets were obtained through comparative analysis, providing base data for subsequent research on non-destructive analysis methods for pressurized water reactors.

Copyright of Inorganic Chemicals Industry is the property of Editorial Office of Inorganic Chemicals Industry and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

HAXPES measurements were carried out using a Scienta Omicron HAXPES instrument to provide reference spectra for depleted uranium dioxide. High purity uranium dioxide, as confirmed by trace elemental analysis and x-ray diffraction, was synthesized via the integrated dry route from uranium hexafluoride. The material was fixed on double sided carbon tape for the analysis with charge control measures in place. The expanded energy range, using a Ga Kα x-ray source, presents core level photoelectrons not observed in traditional XPS. In addition, a region associated with the x-ray induced Auger transitions MNN is evident at binding energies only achievable with HAXPES. The reference spectra presented here act as the first in a line of proposed investigations into the comparison of XPS and HAXPES from surface to bulk as well as a fundamental understanding of the electronic structure of uranium materials. [ABSTRACT FROM AUTHOR]

The distribution of neodymium dopant (as surrogate for americium) within the porous network of sintered (U,Nd)O 2 microspheres prepared via a sol-gel infiltration technique was comprehensively assessed. Powder X-ray diffraction (XRD) results indicated the formation of a neodymium oxyapatite secondary phase with Nd 4.69 (SiO 4) 3 O composition, alongside the principal U 1-y Nd y O 2 face-centered cubic phase for y = 20 and 30 mol%. The hexagonal oxyapatite phase was likely formed during sintering under reducing conditions, when volatile SiO from an external contamination reacted with the Nd3+ in the microspheres. Transmission electron microscopy (TEM) analyses on lamellae prepared via focused ion beam (FIB) further confirmed a distinct phase separation between the main (fluorite U 1-y Nd y O 2) phase and the oxyapatite phase, without mutual solubility of Si and U. To gain a deeper insight into the microstructure, energy dispersive X-ray spectroscopy (EDS) and electron probe microanalysis (EPMA) were performed at different locations in cross-sectioned U 1-y Nd y O 2 microspheres. The elemental mappings showed a higher intensity of combined Nd and Si content in the periphery region, where a higher filling of the porosity network is to be expected. XRD lattice parameter analysis was found to be the most reliable method to elucidate the Nd metal fraction y in U 1-y Nd y O 2 , as compared to local quantification methods using electron beam induced X-ray fluorescence. The sintered microspheres with a targeted 20 and 30 mol% Nd dopant concentration contain around 3 ± 2 wt% and 5 ± 2 wt% Nd-oxyapatite phase. The formation mechanism of the oxyapatite phase appears to involve capillary infiltration in the liquid state, during high temperature sintering, and it may be of interest for other applications. Despite the phase heterogeneity, the method was successful in infiltrating Nd throughout the porous microspheres at high loading levels, as intended for the fabrication of transmutation targets. The distribution of the Nd-dopant throughout the microsphere was achieved as desired, and this study presents a successful fabrication and detailed characterization of porous uranium oxide microspheres doped with Nd3+, contributing to the development of transmutation target research. [Display omitted] [ABSTRACT FROM AUTHOR]

Copyright of Uranium Mining & Metallurgy is the property of Uranium Mining & Metallurgy Editorial Office and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Protection against surface corrosion is essential for ensuring the reliability and long-term durability of uranium materials. Atomically thin two-dimensional (2D) nanomaterials, known for their unique chemical inertness, are particularly promising as anticorrosion coatings. The representative 2D nanomaterials from different classes (insulator, semimetal, semiconductor, and conductor), including BN, graphene, MoSe2, MoS2, and oxygen-passivated Ti2C layers (Ti2CO2 and Ti2CO), were selected to investigate their interactions with the UO2(111) surface using quantum mechanical calculations. Our results show that graphene and h-BN exhibit physical adsorption with the lowest binding energies, less than 1.0 J/m2. In contrast, MoS2 and MoSe2 demonstrate chemical adsorption in the range of 1.0 to 1.5 J/m2. The highest binding energy of 1.7 and 3.1 J/m2 was predicted for Ti2C-based MXenes (Ti2CO2 and Ti2CO, respectively). These results establish the MXene class as the most promising coating for UO2 among the 2D materials considered. It is worth noting that surface defects, whether induced by oxidation or reduction, can influence the strength of the coating, with the primary determinant being the nature of the 2D nanomaterial. [ABSTRACT FROM AUTHOR]

The possibilities of conducting a direct mass spectral determination of Li, Be, Mg, Sr, Zr, Mo, Ru, Rh, Pd, Ag, Cs, Ba, La, Ce, and Nd in the products of processing spent nuclear fuel based on metallic uranium or its oxides are discussed. The optimum method of sample preparation is selected to ensure a quantitative transition of all elements into a working solution. The influence of the uranium content on the analytical signal intensity and the detection limits of the elements to be detected is estimated. The presence of 1 g/L uranium in a solution is found to decrease the analytical signal intensity by 50 times and worsens the detection limits of individual analytes by two orders of magnitude compared to these indicators in uranium-free solutions. The possibility of using internal standards to correct matrix effects is shown. [ABSTRACT FROM AUTHOR]

Frustrated antiferromagnets offer a captivating platform to study the intricate relationship of magnetic interactions, geometric constraints, and emergent phenomena. By controlling spin orientations, these materials can be tailored for applications in spintronics and quantum information processing. The research focuses on the interplay of magnetic and exchange anisotropy effects in artificial heterostructures based on a canonical frustrated antiferromagnet, UO2. The potential to manipulate the spin directions in this material and switch between distinct antiferromagnetic (AFM) states is investigated using substrate‐induced strain. The phenomenon is probed using exchange bias effects in stoichiometric UO2/Fe3O4 bilayers. By employing many‐body first‐principles calculations magnetic configurations in the UO2 layers are identified. Even a minor tetragonal distortion triggers a transition between AFM states of different symmetries, driven by a robust alteration of single‐ion anisotropy due to the distortion. Consequently, this change influences the arrangement of magnetic moments at the UO2/Fe3O4 interface, affecting the magnitude of exchange bias. The findings showcase how epitaxial strain can effectively manipulate the AFM states in frustrated antiferromagnets by controlling single‐site anisotropy. [ABSTRACT FROM AUTHOR]

Morphological analysis of uranium materials has proven to be a key signature for nuclear forensic purposes. This study examines the morphological changes to magnesium diuranate (MDU) and sodium diuranate (SDU) during reduction in a 10 % hydrogen atmosphere with and without steam present. Impurity concentrations of the materials were also examined pre and post reduction using energy dispersive X-ray spectroscopy combined with scanning electron microscopy (SEM-EDX). The structures of the MDU, SDU, and UOx samples were analyzed using powder X-ray diffraction (p-XRD). Using this method, UOx from MDU was found to be a mixture of UO2, U4O9, and MgU2O6 while UOx from SDU were combinations of UO2, U4O9, U3O8, and UO3. By SEM, the MDU and UOx from MDU had identical morphologies comprised of large agglomerates of rounded particles in an irregular pattern. SEM-EDX revealed pockets of high U and high Mg content distributed throughout the materials. The SDU and UOx from SDU had slightly different morphologies. The SDU consisted of massive agglomerates of platy sheets with rough surfaces. The UOx from SDU was comprised of massive agglomerates of acicular and sub-rounded particles that appeared slightly sintered. Backscatter images of SDU and related UOx materials showed sub-rounded dark spots indicating areas of high Na content, especially in UOx materials created in the presence of steam. SEM-EDX confirmed the presence of high sodium concentration spots in the SDU and UOx from SDU. Elemental compositions were found to not change between pre and post reduction of MDU and SDU indicating that reduction with or without steam does not affect Mg or Na concentrations. The identification of Mg and Na impurities using SEM analysis presents a readily accessible tool in nuclear material analysis with high Mg and Na impurities likely indicating processing via MDU or SDU, respectively. Machine learning using convolutional neural networks (CNNs) found that the MDU and SDU had unique morphologies compared to previous publications and that there are distinguishing features between materials created with and without steam. [ABSTRACT FROM AUTHOR]

Oxidation of UO2 and UN by atmospheric oxygen and nitrogen hemioxide, which is a hard-to-localize greenhouse gas, was investigated by thermal analysis. For oxidation, mixtures of N2O–N2 and O2–N2 were used with a 20% volume fraction of the oxidizing agent. For UO2 and UN, the phase composition of the final oxidation product in air and in N2O is the same―U3O8. In both cases, N2O behaves as a milder oxidizing agent compared to atmospheric oxygen. Oxidation of UO2 and UN in an N2O flow starts at a temperature 180 and 70°C higher than in air, respectively. The oxidation of UN in an N2O flow proceeds in three stages. At the first stage, the reaction products are UO2 and U2N3; UO2 is the product of the second stage; and at the third stage U3O8 is produced. No pronounced staging is observed in the UO2 oxidation process. The possibility of utilizing nitrogen hemioxide when it is used in the course of voloxidation (oxidation) of spent nuclear fuel is shown. [ABSTRACT FROM AUTHOR]

The existing wet reconversion processes for the recovery of scraps generated in manufacturing of nuclear fuel are complex and require several unit operation steps. In this study, it is attempted to simplify the recovery process of high-quality fuel-grade UO2 powder. A novel wet reconversion process for uranyl nitrate hexahydrate solution is suggested by using a newly developed pulsed fluidized bed reactor, and the resultant chemical characteristics are evaluated for the intermediate ammonium uranate hydrate product and subsequently converted UO2 powder, as well as the compliance with nuclear fuel specifications and advantages over existing wet processes. The UO2 powder obtained by the suggested process improved fuel pellet properties compared to those derived from the existing wet conversion processes. Powder performance tests revealed that the produced UO2 powder satisfies all specifications required for fuel pellets, including the sintered density, increase in re-sintered density, and grain size. Therefore, the processes described herein can aid realizing a simplified manufacturing process for nuclear-grade UO2 powders that can be used for nuclear power generation.

The Mayer group expansion method is extended on the low-temperature solids where quantum effects cannot be ignored and applied to the prediction of polymorphic transition line between cubic Fm3m and orthorhombic Pnma phases in the highly compressed ionic crystal of uranium dioxide. The temperature dependence of the transition pressure and volume changes are evaluated and the results are compared with existing experimental data and ab initio predictions. [ABSTRACT FROM AUTHOR]

In this study, the effect of point defects on the thermal conductivity of UO2 is investigated. Especially, the effects of oxygen vacancy and interstitial are considered. Thermal conductivity of UO2, UO2+0.25 and UO2-0.25 is calculated by solving the phonon Boltzmann equation (BTE) under the relaxation time approximation (RTA). The results show that introducing any defects to the lattice structure of UO2 decreases thermal conductivity significantly. In addition, the results show that the variation of the thermal conductivity of UO2-0.25 is much lower than that of UO2+0.25 in the temperature interval of 300 to 1000 Kelvin.

In the present study, we investigate the effectiveness of existing methods for evaluating the heat capacity of uranium dioxide (UO2 ) by considering the fractal structure of materials. Two well known theoretical methods were investigated in this work which have the advantages of giving accurate and efficient calculation results for the heat capacities of uranium dioxide for a wide range of temperatures. To verify the effectiveness of these theoretical methods, we compare the calculation results with existing experimental data from the literature. Based on the concept of fractality, the results obtained in this work are useful for determining the best theoretical approaches, i.e., those that are more convenient and meaningful for the evaluation of the heat capacities of uranium dioxide. The approaches and results presented in this study can provide useful knowledge for accurate investigations of the physical properties of other nuclear fuels. [ABSTRACT FROM AUTHOR]

The dissolution step of uranium and mixed oxide uranium-plutonium spent nuclear fuel is a key initial step of a new alternative hydrometallurgical method—the CARBEX process. The study considers carbonate oxidizing systems NaHCO3/Na2CO3–H2O2/2Na2CO3·3H2O2/М2S2O8, where M = Na, K, or NH4+, for dissolving actinide oxide powders. The chemical and physical factors that determine the rate of oxidative dissolution of powdered individual oxides UO2, U3O8, PuO2 and NpO2 in carbonate media were found. The results obtained are important for the development of the oxidative and sonochemical options of dissolution of highly calcined crystalline samples of uranium, plutonium and neptunium oxides, as well of spent nuclear fuel in carbonate media. [ABSTRACT FROM AUTHOR]

Coupling High Energy Resolution Fluorescence Detection (HERFD) with electric dipole selection rules, the uranium 5f 5/2 and 5f 7/2 Unoccupied Densities of States (UDOS) have been determined in UF 4 , UO 2 and UCd 11 . Striking changes were observed between UF 4 (5f 2 ) to UCd 11 (5f 3 ), consistent with the Intermediate Coupling Model. Results for UO 2 were confirmed by Bremstrahlung Isochromat Spectroscopy (BIS). This new approach can be a powerful tool for shedding light on the hidden mysteries of Actinides 5f-electrons properties.

It turns out that the ground states of some systems are symmetry-broken states in which some property is not symmetrically distributed. In the case of strongly correlated electron systems that were studied by the DFT+U method, researchers have shown that the total energy of the system is a multi-minima function of electron-configuration parameters and one has to single out the ground state out of the couples of minimum-energy states. However, the methods already introduced to determine these local minimum states, were not able to predict all such states, which may include the "true" ground state. In this work, we introduce a new simple and straight-forward method of SMC to find the GS as well as the meta-stable states of the 1k-order anti-ferromagnetic configuration for UO2. Using this method, it is shown that the ground state of the UO2 system is a spin-symmetry broken state of the electron spin magnetizations of oxygen atoms. Depending on the way we apply the SMC method, we obtain different numbers of meta-stable states, but the same ground states. The energetic properties, geometric properties, the electronic density distributions, and the electronic polarization density distributions of the ground state and the meta-stable states are shown to be different from each other. These properties also are shown to be sensitive to the magnitude of the initial opposite magnetizations of up-spin U-atoms (U1) and down-spin U-atoms (U2) in the 1k-order anti-ferromagnetic configuration, but the number of meta-stable states as well as the ground-state properties are insensitive to this magnitude. Using the PBEsol-GGA approximation for the exchange-correlation, we obtain the ground-state properties in excellent agreement with experiments.

The influence of technological processes and manufacturing of uranium dioxide fuel pellets for fuel elements for experimental fuel assembly (FA-X) which was designed as an alternative fuel for the nuclear research installation (NRI) "Neutron Source Controlled by Electron Accelerator" were investigated. Unlike standard production processes of UO2 pellets, the special feature fabrication process of this nuclear fuel type is production of uranium dioxide powder with enrichment of 4.4 %wt. of 235U achieved by mixing of two batches of powders with different uranium contents: 0.4 %wt. 235U and 19.7%wt. 235U, as well as ensuring the required tolerance of fuel pellets without the use of machining operations. A set of design and process documentation were developed in the R&D Center at NSC KIPT. Experimental stack of fuel pellets, fuel elements and a pilot fuel assembly FA-X were fabricated and designed to be compatible and interchangeable with VVR-M2 fuel assembly adopted as a standard assembly for the first fuel loading at the "Neutron Source Driven by an Electron Accelerator" FA. As opposition to the variant of VVR-M2 fuel assembly which consisted of three fuel rods of tubular shape with dispersion composition UO2‑Al, FA-X accommodates six fuel rods of pin-type with UO2 pellet which located in the zirconium cladding (E110) as the closest analogue of fuel rods of VVER-1000 power reactor. Inside cladding locate a 500 mm high fuel stack which is secured against displacement by a spacer. In the basic variant of FA-X the fuel pellets are made of UO2 with 235U enrichment near 4.4 %wt.

Oxidative dissolution of powdered UO2 and U3O8 samples in Na2CO3/NaHCO3–H2O2 solutions was studied. The U3O8 samples were prepared from UO2 at 480–1200°С. The rate and degree of the U3O8 dissolution depend on the characteristics of the oxide powder and on the dissolution temperature. The process rate constants under the conditions of the experiments were calculated. Based on the results obtained, the conditions of complete dissolution of UO2 and U3O8 in Na2CO3/NaHCO3–H2O2 solutions were determined. [ABSTRACT FROM AUTHOR]

After the Fukushima nuclear accident, the accident tolerant fuel (ATF) is developed to improve the ability of nuclear fuel components to resist the serious accidents and becomes the hot research topic in the nuclear industry. The enhanced thermal conductivity UO2 pellets represented by BeO and SiC doping, high uranium density and high thermal conductivity fuel pellets, and fully ceramic microencapsulated fuel pellets were reviewed in this paper. The advantage characteristics, thermal conductivity, preparation process, and research progress of the accident resistant fuel pellets were introduced. The problems and application prospects of the accident resistant fuel pellets were focused and prospected to provide the reference for the study of the accident tolerant fuel pellets.

[Display omitted] • Ethanol-TPD has been studied on a UO 2 thin film. • Two channels for ethanol desorption are identified. • The first is coverage dependent while the second is not. • This second desorption channel is accompanied by the desorption of CH 3 CHO + H2 and CH 2 = CH 2 + H2O. • The molar ratio of these two products (CH 3 CHO/CH 2 = CH 2) fits with the U-O bonding nature. The reaction of ethanol has been investigated on a UO 2 thin film by temperature programmed desorption. Two channels for ethanol desorption are identified. The first, in the 250–500 K region, is coverage dependent while the second with a maximum peak temperature (T p) at ca. 630 K is not. The desorption energy, E d , of the second channel is found to be equal to ca. 150 kJ/mol with a prefactor of 1012 s−1 and a desorption order n = 2. This is attributed to surface ethoxides re-combinative desorption. This second desorption channel is accompanied by the desorption of acetaldehyde (and hydrogen) and ethylene (and water). Acetaldehyde (CH 3 CHO) desorption, produced by the dehydrogenation of ethoxides, was sensitive to surface coverage. Its T p changed from 640 K at θ = 0.06 to 610 K at θ = 1, while ethylene (CH 2 CH 2) desorption T p , produced by the dehydration of ethoxides did not shift. The molar ratio of these two products (CH 3 CHO/CH 2 CH 2) of 0.8 is similar to that previously found on a UO 2 (1 1 1) single crystal and fits with the U O bonding nature that contains a non-negligible fraction of covalency. [ABSTRACT FROM AUTHOR]

Uranium dioxide (UO 2) is the standard fuel used in light water reactors (LWRs). However, it has a low thermal conductivity that ultimately limits its performance both during normal operation and in accident conditions. Adding a material with high thermal conductivity is a potential approach to enhance the thermal conductivity of UO 2. Forming an interconnected structure of high-conductivity material can significantly enhance the overall thermal conductivity of the composite. Molybdenum (Mo) has been used as an additive material in UO 2 composites previously. A new method for the fabrication of interconnected UO 2− Mo composites using pre-sintered UO 2 granules to improve the continuity of Mo channels was investigated in this study. UO 2 –10 wt% Mo composites were produced using UO 2 granules and 1073 K and 1473 K pre-sintered UO 2 granules, followed by spark plasma sintering (SPS) of the mixtures at 1473 K for 5 min. The composites were characterised using scanning electron microscopy and X-ray diffractometry and their thermal conductivities were measured by the laser flash method and compared with a reference UO 2 pellet. At a maximum measurement temperature of 1073 K, a 52 % increase in thermal conductivity was observed in the composites containing UO 2 without pre-sintering, and UO 2 pre-sintered at 1073 K. The increase was 31 % for composites manufactured from UO 2 pre-sintered at 1473 K. These results suggest that higher temperature pre-sintering may be detrimental to forming interconnected Mo structures. [ABSTRACT FROM AUTHOR]

The epsilon particles that result from nuclear fission of UO 2 fuel possess both advantages and disadvantages from a spent nuclear fuel (SNF) management perspective. In this study, the effect of epsilon particles, namely Ru, Mo, and Pd, inherent in simulated UO 2 pellets is examined. Various analytical methods have been used to explore the changes in the structural, surface, and electrochemical properties of which contain epsilon particles. A notable finding is that the epsilon particles are not evenly distributed and tend to clump together, transforming into a metallic state after sintering, as detailed in the X-ray diffraction analyses. Energy-dispersive X-ray spectroscopy analyses highlight interesting aspects of the distribution of elements, especially the disappearance of Pd during sintering, which is likely due to its high vapor pressure. Although the lattice structure of UO 2 remains unchanged, the sizes of the grains and pores visibly change, which may influence the tendency of UO 2 pellet-cracking. Despite the addition of the epsilon particles, the electrical conductivity analyses show no significant changes, suggesting that they act as minor impurities without affecting the structural lattice. However, their possible role as catalysts in electrochemical reactions opens new and interesting areas that require thorough investigation. Moreover, examining the anodic dissolution under various conditions provides detailed insights into UO 2 dissolution and oxidation, revealing how epsilon particles subtly influence the oxidative dissolution process. This study clarifies the basic interactions and effects of epsilon particles in UO 2 pellets and broadens the path for a deeper understanding and improvement of nuclear fuel matrices and steering advancements in the safe and effective use of nuclear energy. [ABSTRACT FROM AUTHOR]

In this work, we propose a modelling approach for the intra-granular fission gas behaviour in UO 2 under restructuring process. Leveraging the definition of restructured volume fraction, we consider the fuel matrix transition from the non-restructured to the restructured phase, together with the evolution of the corresponding fission gas concentrations retained in the fuel matrix. Firstly, we derive a sweeping term that exchanges fission gas atoms from the non-restructured to the restructured fuel region. The sweeping term is then included in the conventional intra-granular fission gas diffusion problem. Secondly, the spectral diffusion algorithm is employed to solve two spatially-dimensionless problems, properly representing the non-restructured region with micrometric grains and the restructured region with sub-micrometric grains. The model developed is implemented in SCIANTIX, a 0D meso-scale code for physics-based modelling of fission gas behaviour in nuclear oxide fuel and compared with experimental data and semi-empirical models. • Fission gas behaviour in forming high-burnup structure. • Model tailored for application in grain-scale fission gas behaviour modules. • Assessment of the model with available open-literature experimental data. • Comparison of the model results with state-of-the-art approaches. [ABSTRACT FROM AUTHOR]

A study of irradiation-induced creep and re-sintering of large grain sized fuel of the VVER type was carried out using the developed procedures and experimental facilities under irradiation with fission rates of 1.0 × 1013 and 1.2 × 1013 1/(cm3 s). Strain diagrams generated during creep studies within the strain range of 10−40 MPa were obtained at effective sample temperatures of 943, 1020, 1200 and 1301 K. It was demonstrated that strain is controlled by irradiation-induced creep within the temperature range of up to 1123 K. This was confirmed by its linear dependency on stress. At higher temperatures, the effect of radiation-thermal creep grows, and its contribution to the total strain increases with the growth of stress. Creep rate values obtained at temperatures of 1200 and 1301 K are higher than those of uranium dioxide with standard grain size. Strain diagram produced during the studies of re-sintering of an 88.32 mm high fuel column was obtained at the effective sample temperature of 960 K. In terms of height, re-sintering amounted to 0.070−0.095%. It was demonstrated that irradiation-induced re-sintering did not affect the irradiation-induced creep rate within the experimental accuracy range. [ABSTRACT FROM AUTHOR]

In this study, the possibility of sintering industrial pressed uranium dioxide pellets using microwave radiation for the production of nuclear fuel is shown. As a result, the conditions for sintering pellets in an experimental microwave oven (power 2.9 kW, frequency 2.45 GHz) were chosen to ensure that the characteristics of the resulting fuel pellets meet the regulatory requirements for ceramic nuclear fuel, including the following: a density of about 10.44 g/cm3; a volume fraction of open pores of tablets of about 0.1%; an oxygen coefficient of no more than 2.002; hydrogen content of about 0.30 ppm; and the change in density after re-sintering on average no more than 1.16%. [ABSTRACT FROM AUTHOR]

An analysis of sintered uranium dioxide has been conducted using a hyperspectral camera sensitive to short-wave infrared wavelengths in the range 949-2472 nm. Three groups of sintered UO 2 nuclear fuel pellets were prepared and analysed, with stable sub-group surrogates introduced at the preparation stage to emulate the presence of fission product elements. Results show a clear, consistent, and reproducible spectral response across the pellet groups for pure UO 2 . Furthermore, the addition of fission product elements is observed to affect the shortwave infrared response, causing an overall flattening of the spectra. We have shown that this spectral change is correlated significantly with the presence of lanthanides in the fuel matrix. This result could have important potential in post-irradiation examination for quantifying nuclear fuel burn-up and radiotoxicity at discharge, as the hyperspectral imaging setup allows multiple (> 20) samples to be analysed in a single image, captured in under 30 s., Competing Interests: Declarations. Competing interests: The authors declare no competing interests., (© 2025. The Author(s).)

Actinide materials have various applications that range from nuclear energy to quantum computing. Most current efforts have focused on bulk actinide materials. Tuning functional properties by using strain engineering in epitaxial thin films is largely lacking. Using uranium dioxide (UO2) as a model system, in this work, the authors explore strain engineering in actinide epitaxial thin films and investigate the origin of induced ferromagnetism in an antiferromagnet UO2. It is found that UO2+x thin films are hypostoichiometric (x<0) with in‐plane tensile strain, while they are hyperstoichiometric (x>0) with in‐plane compressive strain. Different from strain engineering in non‐actinide oxide thin films, the epitaxial strain in UO2 is accommodated by point defects such as vacancies and interstitials due to the low formation energy. Both epitaxial strain and strain relaxation induced point defects such as oxygen/uranium vacancies and oxygen/uranium interstitials can distort magnetic structure and result in magnetic moments. This work reveals the correlation among strain, point defects and ferromagnetism in strain engineered UO2+x thin films and the results offer new opportunities to understand the influence of coupled order parameters on the emergent properties of many other actinide thin films. [ABSTRACT FROM AUTHOR]

The combination of electronic and nuclear energy deposition can have an effect on the defect production and evolution that is not simply the sum of both independent contributions. In nuclear reactors, the fuel is exposed to the irradiation of several particles such as the fission fragments (FF) whose electronic to nuclear energy loss ratio evolves along their path. To understand the impact of this energy-loss coupling on the fuel microstructure, single and dual-beam ion irradiations of uranium dioxide (UO 2) were carried out. The damage evolution was investigated by Raman spectroscopy analysis correlated with Transmission Electronic Microscopy observations. It was found a significant acceleration effect of the electronic energy dissipation on the formation and evolution of defects induced by the nuclear energy deposition. The magnitude of this acceleration was shown to depend on both the electronic energy loss level and the microstructure at the moment when the coupling occurs. [Display omitted] [ABSTRACT FROM AUTHOR]

The safe use of nuclear energy is crucial, especially after the Fukushima nuclear disaster. A novel uranium dioxide (UO 2)-graphene nanosheet (GN) ceramic matrix composite fuel has been proposed as an accident tolerant fuel (ATF) with excellent thermal-mechanical properties. The UO 2 -GN pellet system is developed using mechanically mixed and hot pressing (HP) sintering, providing a new idea for the industrial production of UO 2 -GN pellets. Raman spectroscopy and scanning electron microscopy (SEM) show that most GN survives in composite pellets by HP sintering. Furthermore, we discovered various toughening mechanisms in the composite pellets for the first time. With the further increase of graphene content (1, 3, 5 wt%), the behavior of GN in the matrix changed from anchoring to wrapping. Among different sintering methods, HP sintering has excellent thermal performance. The thermal diffusivity of the composite pellets prepared by HP sintering (HPed) doped with 3 and 5 wt% is 1.300 and 1.423 mm2 s−1 at 973 K, increased by 5.43% and 15.41% compared with the pure pellets, respectively. Furthermore, the thermal conductivities of the HPed doped with 1, 3, and 5 wt% are 4.31, 5.00, and 6.06 W m−1 K−1 at 973 K, which were 3.61%, 20.20%, 45.67% higher than those of the pure pellets, respectively. These results highlight the immense possibilities of HP sintering in fabricating UO 2 -GN pellets to improve fuel performance. [ABSTRACT FROM AUTHOR]

• An in situ Raman monitoring of isotopically-labeled studtite growth was performed. • The link between studtite formation mechanism and radiolysis effects was described. • The preservation of the peroxide bond between H 2 O 2 and studtite was demonstrated. • The oxidation mechanism of uranium dioxide is detailed. Studtite [(UO 2)(O 2)(H 2 O) 2 ]·2(H 2 O) is a secondary phase precipitating during the alteration of uranium-bearing materials. The mechanistic link between the formation of uranyl and peroxide bonds in the solid and the nature of the oxidizing species produced by water radiolysis remains to be elucidated. In order to improve our understanding of these mechanisms, an original experimental methodology in the presence of 18O isotopes and Raman spectroscopy has been developed. It appears that there is a direct chemical relationship between the peroxo ligands inside studtite and the peroxide entities of H 2 O 2 molecule into the solution. The link between H 2 O 2 formation mechanism in solution by the radiolysis of water, the nature of the radiation and the isotopy of the peroxo ligands inside studtite has been described thanks to a coherent set of experimental data. For the uranyl UO 2 2+ ions, the characterization of its isotopy allows to specify the mechanism of oxidation at the UO 2 /water interface. The isotopic configurations observed for the uranyl ion inside studtite, can be explained by assuming an oxidation mechanism of UO 2 involving both a simple transfer of electrons by interaction with H 2 O 2 and the incorporation of oxygen atoms from the solution into the fluorite structure via OH° radicals. [Display omitted] [ABSTRACT FROM AUTHOR]