Hydroxide coprecipitation in nitric acid was used to synthesize U0.86Pu0.14O2±x solid solutions for the fabrication of model mixed nuclear fuels, one of the main objectives of this study being to evaluate the cationic homogeneity. Three precipitation methods were compared: simultaneous precipitation of U(IV) and Pu(IV), simultaneous precipitation of U(IV) and Pu(III), and precipitation of a premixed solution containing U(IV)-Pu(III). In order to obtain dense pellets, oxide precipitates were calcined and sintered at 1700 °C under Ar-4.3 vol H2 %. Despite an apparent similarity on a macroscopic scale, the pellets produced by the different methods reveal distinct microstructures, indicating differences in the sintering capability of the powders. These differences in microstructure and density have a low impact on the cationic homogeneity of the pellets after sintering. Regardless of the precipitation method used, the distribution of U and Pu elements remains uniform, with a higher level of homogeneity than that measured on MOX fuel obtained by other wet or dry chemistry routes. Finally, the most efficient coprecipitation route was applied to mixed oxides with 11 and 17 mol % Pu, in a range typical of PWR and FNR fuels, revealing a clear influence of Pu content on the physicochemical and sintering behavior of the powders.

An experimental study was conducted to investigate the effects of altering fresh (U,Pu)O2 MOX of high alpha-specific activity under alkaline conditions (pH = 12), in the presence of 2 mmol.L-1 silicate ions and in anoxic environment. This test was conducted in conjunction with analogous experiments using (U,Ce)O2 simulant materials of homogeneous and heterogeneous microstructure. The results obtained indicated that the oxidative dissolution induced by alpha-radiolysis of water in contact with (U,Pu)O2 MOX was suppressed under these alteration conditions. The uranium concentration determined at equilibrium was found to be fully comparable to those obtained with the (U,Ce)O2 model materials. This observation suggested that the alteration mechanism occurring at the solid/ solution interface was independent of the alpha-activity of the material and highlighted the analogy of the two types of materials in the tested conditions. Post-alteration characterizations of the material surface using Raman spectroscopy, TEM/EDS and SEM were conducted. The results obtained supported the assumption that adsorption of silicate ions onto reactive sites in the material was responsible for the inhibition of uranium oxidation. This study revealed the important role played by silicate ions in reducing uranium release in alkaline conditions and under alpha radiolysis of water. Considering that uranium is a tracer of the MOX matrix alteration, these results had positive implications regarding the long-term management of spent fuel in deep disposal especially in case of contact with cementitious water.

Nuclear energy is expected to play a key role in future low–carbon energy systems, but its long–term viability depends on advanced fuel cycles capable of recycling valuable materials from spent fuel. Fast reactors (FRs) operating with mixed uranium and plutonium oxides (MOx) are central to this strategy, but their safe deployment requires accurate knowledge of the thermophysical properties of the fuel under irradiation. However, such data remains extremely scarce, as irradiated MOx is difficult to obtain, highly radiotoxic and inaccessible for high–temperature measurements. This lack of experimental data is a significant source of uncertainty for the assessment of reactor safety analysis. Here we show that specially designed surrogate materials – SIMMOx – can reproduce the effect of dissolved and precipitated FPs on the thermal conductivity of irradiated FR–MOx fuel without any irradiation. The predominant role of dissolved FPs was demonstrated for the first time by the incorporation of these FPs into the (U,Pu)O2–x matrix during fabrication. These results clarify the effect of dissolved FPs and secondary phases on the thermal transport degradation in MOx fuel. More broadly, they provide a new experimental route to study fuel behaviour at high temperature without the constraints associated with irradiated materials.

In the context of spent fuel recycling and the valorization of plutonium, (U,Pu)O2 mixed oxides (MOX) have been developed for use in French Pressurized Water Reactors (PWR). They are also leading candidates for some GEN IV reactor concepts, such as sodium-cooled fast reactors (SFR). One of the critical challenges in the nuclear industry is the mastery of the nuclear fuel cycle, specifically plutonium multirecycling. In order to achieve this goal, it is crucial to identify the secondary phases created during irradiation. In this work, (U,Pu)O2 MOX have been doped with 11 stable fission products (FP) (Sr, Y, La, Nd, Ce, Zr, Mo, Pd, Rh, Ru, Ba) to reproduce FP-based precipitates existing in the real spent fuel. The structural and microstructural properties of these secondary phases were characterized by coupling scanning electron microscopy-energy dispersive X-ray spectroscopy (SEM-EDS), Electron Probe MicroAnalysis (EPMA), and synchrotron techniques such as X-ray Absorption Spectroscopy (XAS) and Synchrotron Powder X-ray Diffraction (SP-XRD). This analysis highlights the relationship between the partial segregation among metallic FP (Mo, Pd, Rh, Ru) and their crystallographic structures, as well as the speciation shift of several FP induced by the addition of Ba. The synthesized SIMMOX samples present a secondary phase representative of irradiated MOX and can be used as an effective model material to study spent nuclear fuel and its reprocessing.

The sintering of hydrothermally-prepared U0.8Ce0.2O2+delta powders was investigated as a model system for future MOX nuclear fuels. Dilatometric analyses evidenced a two-step process, with dehydration, redox reactions and partial removal of residual carbon at low temperature, followed by the onset of sintering around 800 degrees C. The powders can be directly sintered at 1700 degrees C, yielding pellets with relative densities up to similar to 98%. Nevertheless, maintaining temperature for several hours led to cracking and de-densification. A sintering map (1300-1700 degrees C, 1-15 h) revealed two regimes associated with densification and grain growth, separated near 91% theoretical density, with a unique densification trajectory. Ex situ analyses confirmed the persistence of a fluorite structure and progressive reduction toward oxygen-stoichiometric dioxide. Residual carbon, still detected at high temperature, is proposed to contribute to mechanical degradation. These results highlight the potential of hydrothermal conversion to produce homogeneous, directly sinterable mixed oxides, while emphasizing the need to optimize carbon removal.

Hf1-xYxO2-x/2 ceramics have been recently suggested as a solid electrolyte for potentiometric sensors monitoring oxygen in liquid sodium. High density ceramics can be produced by an innovative dropwise wet-process using hydroxide coprecipitation allowing a homogeneous mixture of hafnium and yttrium under specific stoichiometric conditions in the presence of ammonia. Different parameters such as the synthesis duration (2-26 min), concentration (2-10 M) and excess of ammonia (200-1000 %) used were varied to determine the optimized synthesis parameters for obtaining a very dense material. After synthesis, the precursor obtained was mainly characterized by BET and then calcined under air at 700 degrees C to obtain the corresponding oxide. The latter was further characterized by BET, SEM, EDS and TGA before being shaped and sintered at 1600 degrees C. The relative density of the densest ceramics reached 97.6 +/- 0.5 % of the theoretical density, rarely obtained by wet chemistry routes.

A multiparametric study was developed to optimize the hydrothermal synthesis of chernobylite solid solutions (Zr1-xUxSiO4), a phase first discovered in the Elephant's foot of the Chernobyl nuclear power plant following the 1986 nuclear accident. The goal was to obtain phase-pure samples for thermodynamic investigation. Optimal synthesis conditions were determined starting from uranium (IV) and zirconium (IV) chloride precursors, using a hydrothermal treatment for seven days at 250 °C with a reaction medium maintained within a pH range of 1.4 to 1.8, which minimized the formation of secondary phases Under these conditions, solid solutions of chernobylite were synthesized up to x = 0.80. However, the systematic presence of residual oxide phases in the samples required the development of a purification process involving alternating leaching steps in basic and acid media. This process removed amorphous silica and selectively dissolved residual oxide phases, respectively. This protocol yielded pure Zr1-xUxSiO4 samples over a wide range of compositions (x ≤ 0.6). The resulting powders consisted of spherical agglomerates of approximately 300 µm in diameter, with specific surface areas ranging from 19 to 22 m2 g-1. Finally, the thermal stability of the chernobylite solid solutions was evaluated using heat treatments between 1000 °C and 1300 °C. All samples, except for those most enriched in uranium, proved to be thermally stable even at 1300 °C.

This study introduces a novel, safer, and more efficient method for the preparation of mixed oxide powders (ThO2/CeO2 as surrogates for UO2/PuO2) homogeneously distributed at a nanoscale using aqueous multicolloidal sols. The latter were obtained either by mixing salts in solution or by mixing monoelemental sols. Combining Small and Wide Angles X-ray Scattering with Pair Distribution Function techniques, size, shape and the interactions of the various colloids were studied as a function of pH and amount of complexing agent in solution. The data obtained allowed for the selection of sols that remained stable for several months. The latter were then freeze-dried to induce assembly of the colloids forming ordered structures (e.g., body-centered cubic or lamellar phases). Depending on the organization of the resulting colloid stacking, driven by their surface charge, Scanning Electron Miscroscopy and High-Resolution Transmission Electron Microscopy highlight that the materials obtained after thermal treatment exhibit an assembly of monodisperse ThO2 and CeO2 nanocrystallites of few nanometers with a homogeneous distribution at the microscale and nanoscale depending on the experimental conditions. This method addresses key challenges of nuclear Mixed Oxide nuclear fuel fabrication, such as radioactive dust or heterogeneous actinide distribution, thus opening promising prospects for the development of advanced Gen-IV nuclear fuels.

In this study, the influence of the working atmosphere on the sinterability and chemical durability of Nd-doped UO2 mixed oxides was investigated. To this end, the starting powder was first prepared by a hydroxide coprecipitation route, resulting in a nano-sized granulometry combined with a high specific surface area. The powders were then converted to oxides by heating and sintered in pellet form at 1600 degrees C under an argon or reducing (Ar-4 %H2) atmosphere. The use of argon or reducing atmosphere resulted in very different densification pathways and final microstructures. The reducing sintering atmosphere hindered the uranium (IV) oxidation that could occur at high temperature, leading to the formation of U3O8, as was the case when working under argon atmosphere. Regarding the microstructure of the sintered pellets, the use of an argon sintering atmosphere resulted in an average grain size ten times larger than that of a reducing sintering atmosphere, while macroscopic properties such as relative density, porosity and homogeneity of cation distribution at the pellet scale remained the same. Nevertheless, a slight local enrichment of Nd at the grain boundaries was observed for the pellet sintered under Ar-4 %H2. In a second step, the study of the chemical durability of these sintered samples showed a significant influence of the sintering atmosphere on the dissolution kinetics and mechanism. These differences could be related to the microstructural properties of the pellets, i.e. the average grain size and the occurrence of grain boundaries. The cation distribution in the pellets could also influence their chemical durability, such as local Nd enrichment, the formation of defects in the oxygen sublattice and the presence of a different fraction of U(V) depending on the sintering atmosphere, as shown by HERFD-XANES measurements. The use of reducing or argon sintering atmospheres could even direct the charge compensation mechanisms that occur into the solid, thereby simultaneously affecting the sinterability and chemical durability of the samples.

The effects of atmosphere and cerium content on the densification and the final microstructure of homogeneous U1-xCexO2+delta solid solutions (x = 0.10; 0.25; 0.50) were investigated. Dilatometric studies first revealed that while the cerium content only slightly modifies sintering under an Ar/H2 atmosphere, a change of the gas to argon dramatically modifies densification kinetics. As a result, samples prepared under Ar/H2 exhibit a dense microstructure with micrometric grains. Conversely, samples sintered under argon appeared to be less densified but with larger grains. These changes were correlated with the variation of the final O/M stoichiometry (M = U+Ce) and illustrated by the construction of sintering maps. Finally, grain growth was found to be driven by grain boundary motion when the O/M ratio remained close to 2.00, while usual power laws did not apply for most hyper-stoichiometric samples. The first values of activation energy for the sintering of U1-xCexO2+delta solid solutions under an Ar/H2 atmosphere were also determined.

The precipitation of U(VI) peroxide (UO2(O2)·4H2O, known as studtite) has been extensively studied in mildly acidic media (i.e. pH > 1). However, only a few studies have investigated the influence of highly acidic conditions (i.e. pH < 1) on the precipitation of studtite, particularly regarding the morphology of the final powder. Therefore, the influence of high acidity (0.1 M < C(H+)ini < 2.0 M) and the C(H2O2)ini/C(U)ini molar ratio (ranging from 10 to 70) on the precipitation kinetics, as well as the morphological and crystallographic properties of U(VI) peroxide, was investigated in this study. Decreasing the acidity and increasing the C(H2O2)ini/C(U)ini molar ratio have significantly enhanced both the final precipitation yield and the reaction kinetics. Furthermore, a correlation was found between the initial saturation index of the reaction mixture and the final morphology and crystallite size of the precipitated U(VI) peroxide, independent of the initial precipitation conditions. These results demonstrate that in highly acidic nitric media, it is possible to control the morphology of the precipitated U(VI) peroxide by simply adjusting the initial saturation index. This correlation opens up interesting perspectives for the synthesis of powders with specific morphologies, in particular uranium peroxide or corresponding oxide powders obtained after thermal treatment. This morphology control is of great interest in the field of nuclear fuel fabrication, as the characteristics of oxide powder, such as flowability or sintering reactivity, have a strong impact on the manufacturing process and the properties of the final ceramic.

The synthesis conditions of mixed uranium and thorium oxides by hydroxide precipitation (wet process) have been investigated and optimized. This mainly consists of controlling the order and the rate of addition of the reagents, which allows better reproducibility during the precipitation stage. Two synthesis methods were compared: Direct Droplet Adding (DDA, gradual addition of the ammonia solution) and Indirect Droplet Adding (IDA, gradual addition of the cation solution), with all other parameters held constant. The solids produced by these two methods were converted to oxide by calcination and then sintered at 1600 degrees C in a Ar-4 %H2 atmosphere to produce dense pellets. Although the pellets produced by the two synthesis methods did not differ at the macroscopic scale, differences in microstructure and chemical durability were observed. DDA induced cationic heterogeneity, which limits the densification of the material and leads to the formation of thorium-rich agglomerates. This resulted in significant open porosity at the end of the sintering stage. Conversely, IDA makes it possible to obtain a very homogeneous cationic distribution within the pellet as well as an increased density after sintering. These microstructural differences also influence the chemical durability of the pellets during dissolution tests. In fact, the presence of cationic heterogeneities in the samples prepared by DDA significantly affects the dissolution kinetics and requires an adaptation of the reprocessing conditions due to the refractory character of such heterogeneities. These results therefore highlight the importance of the synthesis protocol on the microstructure and chemical durability of sintered uranium-thorium mixed oxides.

A significant lack of experimental data concerning the thermophysical properties of U1-yPuyO2-x for high Pu contents (y > 0.45) is observed. To bridge this gap, dense, monophasic, stoichiometric and homogenous oxides U1-yPuyO2.00 samples (with y = 0.60-0.70) were initially manufactured and characterized. To study the impact of the Oxygen/Metal ratio on thermal properties, hypo-stoichiometric samples were obtained from the previous batches. The strategy was to optimize annealing to reach O/M = 1.98 and perform multi-scale characterizations (optical microscopy, X-ray diffraction, Raman spectroscopy, electron probe microanalysis and X-ray Absorption Near Edge Structure). These revealed monophasic and hypo-stoichiometric samples with a homogeneous chemical distribution of cations and oxygen and a complex redox cations speciation: U(IV)/U(V), Pu(III)/Pu(IV) and Am(III). The results demonstrate the potential of these samples for measuring thermophysical properties, and enable the extension of the Pu content validation range of some empirical laws from y = 0.45-0.69.

This study investigates the chemical durability of uranium oxide microparticles (UO2+x and U3O8), as potential reference materials for nuclear safeguards. To optimize long-term preservation, the particles were exposed to three different storage media: dilute nitric acid (10-2 mol L-1 HNO3), deionized water, and ethanol. Dissolution rates in nitric acid (∼5 × 10-4 g.m-2.d-1) were similar to those of bulk uranium oxides, but UO2+x particles experienced greater absolute leaching due to their higher specific surface area, resulting in noticeable morphological changes. In distilled water, rapid precipitation of secondary phases, such as schoepite and studtite, transformed the particle morphology into aggregated platelets. In contrast, ethanol preserved both particle shape and structural integrity over 8 months, ensuring stability for isotopic analysis using large geometry secondary ion mass spectrometry (LG-SIMS). These results suggest ethanol as the most effective storage medium, especially under anhydrous conditions. While both UO2+x and U3O8 exhibited comparable durability, their reactivity was primarily influenced by microstructural differences. This study highlights the importance of selecting appropriate storage conditions based on fabrication methods and particle characteristics to maintain uranium oxide reference materials for nuclear safeguards and ensure their long-term reliability in quality control applications.

Nuclear power plays a pivotal role in ensuring a scalable, affordable, and reliable low-carbon electricity supply. Along with other low-carbon energy technologies, nuclear energy is essential for reducing our reliance on fossil fuels, addressing climate change and air pollution, and achieving a sustainable economy. Whilst significant progress has been made in reducing the volume of final radioactive waste, its management remains one of the most important challenges when considering the continued use and expansion of nuclear energy. This recently published collection highlights the latest technological and scientific advances aimed to improve the safe, long-term, and sustainable management of wastes produced from nuclear power generation.

Hydrothermal conversion of uranyl oxalate into UO 2+ x unravelled by in situ XANES, opening a new avenue for nuclear fuel fabrication.

Nd 1−2 x Ca x U x PO 4 monazite–cheralite with x < 0.1 is prepared by wet chemistry route. Uranium incorporated in the monazite–cheralite structure is 100% U( iv ). Phosphates protect U against oxidation during the thermal treatment of the precursor in air.

The defect chemistry and thermal oxidation of lanthanide (Ln) incorporated-UO2 are critical for understanding and predicting their behavior as enhanced fuels, mixed oxide (MOX) fuels, spent nuclear fuels (SNF), and particles for safeguard purposes. In this study, we independently controlled the Ln type (Ce4+, Nd3+, and Gd3+) and the preparation condition (reduced and nonreduced) to investigate their correlations to the generated non-equilibrated defects correspondingly. From early to late lanthanides: Ce and U formed close-to-ideal solid solutions in Fm-3m and oxidized to (Ce, U)4O9, Nd and U mixing under the reducing condition formed solid solutions with oxygen vacancies aggregating near Nd, and the mixing of smaller Gd with U resulted in short-range subnano-domain segregations with Ia-3 region embedded in the global Fm-3m matrix. Both trivalent Ln-incorporated UO2 oxidized to a mixture of (Ln, U)4O9 and (Ln, U)3O8. From these signature defect structures resulting from both Ln type and preparation condition, we proposed kinetic model and thermodynamic hypothesis for explaining the oxidation resistance of (Ln, U)O2. Although originated from f-block oxides, the discovery of long-range disorder short-range ordering may be not uncommon in other metal oxide systems, which can strongly influence their functionalities and properties.

A known issue for future nuclear reactors is helium accumulation inside the steel structure materials, responsible for structural issues such as embrittlement and cracking. One possible solution is using new types of reinforced steel, such as oxide dispersion strengthened (ODS) steel. It consists of adding oxide nanoparticles to the Fe-based material, especially yttrium oxide (yttria, Y$_2$O$_3$), improving its properties. Therefore, one first step is understanding the helium diffusion inside this system. Very little is known about helium inside yttria, with most studies being theoretical ones. Based on this context, this work proposes a combined theoretical and experimental multiscale approach to investigate helium diffusion inside yttria. The theoretical approach starts with the density functional theory, used to model the atomic yttria cell and determine helium insertion sites. The transitions between the sites were described using the NEB method. Kinetic Monte Carlo was then employed to obtain the interstitial diffusion coefficient expression. It showed a limited diffusion at temperatures below 600 K, which may indicate a tendency for He to be blocked in the oxide. Then, the charged vacancies were explored. It showed that the vacancy further reduces helium diffusion. Finally, it was demonstrated that helium spreads across different vacancies and interstitial sites. The experimental part involved implanting helium ions at 50 keV in samples with nanometric or micrometric grains. Then, the specimens were characterised with transmission electron microscopy (TEM) and thermo-desorption spectroscopy (TDS) techniques. TEM did not evidence detectable bubbles even at the highest studied fluence (1\time 10^{16}$ cm$^{-2}$). The TDS highlighted different mechanisms for helium diffusion and the grain size's role, providing a model for diffusion coefficient calculation based on interstitial diffusion.