Littérature scientifique sur le sujet « UO2 Cr doped »

AbstractCr-doped UO2-based model materials were prepared at SCK CEN, mimicking modern LWR fuels, to understand the influence of Cr doping on the spent fuel dissolution behaviour in geological repository conditions. Tests were carried out with four model materials: depleted UO2, Cr-doped depleted UO2, Pu-doped UO2 and Pu-Cr-doped UO2. Static dissolution experiments have been performed up to 4 months in autoclaves under 10 bar H2 pressure with a Pt/Pd catalyst in media at pH 13.5 and at pH 9. The Cr-doping appeared to reduce the U concentrations by a factor 6 at pH 13.5, but it had no or not much effect at pH 9. Graphic abstract

ABSTRACTThe addition of Cr2O3 to modern UO2 fuel modifies the microstructure so that, through the generation of larger grains during fission, a higher proportion of fission gases can be accommodated. This reduces the pellet-cladding mechanical interaction of the fuel rods, allowing the fuels to be “burned” for longer than traditional UO2 fuel, thus maximising the energy obtained. We here describe the preparation of UO2 and Cr-doped UO2 using Hot Isostatic Pressing (HIP), as a potential method for fuel fabrication, and for development of analogue materials for spent nuclear fuel research. Characterization of the synthesised materials confirmed that high density UO2 was successfully formed, and that Cr was present as particles at grain boundaries and also within the UO2 matrix, possibly in a reduced form due to the processing conditions. In contrast to studies of Cr-doped UO2 synthesised by other methods, no significant changes to the grain size were observed in the presence of Cr.

Cr-doped UO2 as a modern nuclear fuel type has been demonstrated to increase the in-reactor fuel performance compared to conventional nuclear fuels. Little is known about the long-term stability of spent Cr-doped UO2 nuclear fuels in a deep geological disposal facility. The investigation of suitable model materials in a step wise bottom-up approach can provide insights into the corrosion behavior of spent Cr-doped nuclear fuels. Here, we present new wet chemical approaches providing the basis for such model systems, namely co-precipitation and wet coating. Both were successfully tested and optimized, based on detailed analyses of all synthesis steps and parameters: Cr-doping method, thermal treatment, reduction of U3O8 to UO2, green body production, and pellet sintering. Both methods enable the production of suitable model systems with a similar microstructure and density as a reference sample from AREVA. In comparison with results from the classical powder route, similar trends upon grain size and lattice parameter were determined. The results of this investigation highlight the significance of subtly different synthesis routes on the properties of Cr-doped UO2 ceramics. They enable a reproducible tailor-made well-defined microstructure, a homogeneous doping, for example, with lanthanides or alpha sources, the introduction of metallic particles, and a dust-free preparation.

Abstract. In Germany it is planned to directly dispose spent nuclear fuel (SNF) from nuclear power plants together with other high-level radioactive wastes (HLW) from former SNF reprocessing (e.g., vitrified waste), in a deep geological repository for heat-generating wastes – the siting process for this repository was started in 2017 and is ongoing. Based on several decades of research, development, and demonstration (RD&D) it is generally accepted at the technical and scientific level that direct disposal of HLW and SNF in deep mined geological repositories is the safest and most sustainable option (CEC, 2011; IAEA, 2004). The current efforts to improve the performance and accident tolerance of fuels in nuclear power generation resulted in an increased utilization of a variety of new types of light-water reactor (LWR) fuels such as fuels doped with Cr, Al, and Si. This doping leads to a significant change of the microstructure of the fuel matrix. The corrosion behavior of these types of fuels under conditions relevant to deep geological disposal has hardly been studied so far; however, this is of crucial importance as the development of a robust safety case for deep geological disposal of SNF requires a solid understanding of its dissolution behavior over very long time scales (up to 1 million years). To fill this knowledge gap, additional systematic studies on modern doped UO2 fuels were needed. Corrosion experiments with SNF cannot entirely unravel all of the various concurring effects of the dissolution mechanism due to the chemical and structural complexity of SNF and its high beta and gamma radiation field during the first 1000 years; moreover, technical restrictions only allow a very limited number of experiments. Therefore, within the EU-DisCo project (https://www.disco-h2020.eu, last access: 11 October 2021), a very ambitious programme of corrosion studies on irradiated Cr and Al/Cr doped fuels was carried out, which was complemented by systematic single-effect dissolution studies (e.g., with respect to doping level, grain size and thermodynamic aspects) performed on carefully prepared and characterized, simplified UO2-based model materials. Here, we present recent results on the dissolution behavior of tailor-made UO2 model materials in accelerated static batch experiments using H2O2 as simulant for radiolytic oxidants, present in long-term disposal scenarios for SNF in failed container conditions due to the alpha irradiation of water. In these dissolution experiments pure UO2 reference pellets exhibiting different densities and grain sizes, as well as Cr-doped UO2 pellets with various Cr-doping levels, produced using different doping methods having different grain sizes, were used. In addition, Nd-doped and industrially produced Cr- and Cr/Nd-doped UO2 pellets were used to determine the influence of these parameters on the dissolution rates. The dissolution experiments were performed under strictly controlled conditions with respect to exclusion of oxygen, temperature control, and exclusion of light. This bottom-up approach was followed to understand how the addition of Cr-oxide into the fuel matrix affects SNF dissolution behavior under repository relevant conditions. The results of the dissolution experiments performed with real SNF and the model materials obtained by the DisCo partners build the basis for numerical simulations on the dissolution behavior of modern SNF. First results of the data evaluation indicate that the addition of dopants and the consequential modification of the fuel matrix does not lead to a significant change of the dissolution behavior of these fuels under repository relevant conditions compared to standard SNF (i.e. dissolution rates agree within an order of magnitude).

This paper discusses the use of electron backscattered diffraction to characterize restructuring in a set of UO2 samples, irradiated in a pressurized water reactor at a burn-up between 35 and 73 GWd/tU, including standard UO2 samples and Cr-doped UO2 samples, to provide a better understanding of restructuring occurring both on the periphery and in the center of high-burn-up pellets. The formation of a high burn-up structure on the periphery of high burn-up UO2 was confirmed in our experiment. We found restructuring associated with bubble formation of all the samples in the central area, with higher irradiation temperatures when the burn-up exceeded 61 GWd/tU, regardless of their initial microstructure. This restructuring tended to progress with the increasing burn-up and to sub-divide the initial grains into sub-grains, with orientations close to that of the parent grains. Radial changes and differences between these samples showed that the burn-up and the temperature were not the only relevant parameters involved in restructuring.

The expected benefits from Cr-doped fuel are improved retention of fission gases within the pellets due to its large grain size. To demonstrate this, several experiments have been carried out by Halden reactor and Studsvik. These experiments are now being used to benchmark several fuel performance codes among them transuranus code. All this as part of a Coordinate Research Project (CRP) by IAEA named Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS). This work is introducing a novel fission gas diffusivity model for doped fuel in transuranus code. It is observed the benefits of introducing this new model when comparing to the standard model already existing in transuranus. Nevertheless, more work needs to be carried out to fully understand all the phenomena involved in adding dopant in UO2 due to change of thermo mechanical properties.

Dans le contexte de la relance mondiale du nucléaire, motivée par les enjeux de décarbonation et de sécurité énergétique, la question de la sûreté nucléaire est plus que jamais essentielle. Suite à l’accident de Fukushima Daiichi en 2011, un regain d’intérêt s’est porté sur les combustibles tolérants aux accidents (ATF), visant à améliorer la sûreté et les performances des réacteurs nucléaires. Parmi ces innovations, l’UO2 dopé au chrome a émergé comme une option prometteuse, démontrant une meilleure rétention des gaz de fission. Cependant, les mécanismes précis par lesquels le dopage au chrome influence le comportement des produits de fission restent mal compris, notamment en ce qui concerne leur diffusion au sein du matériau. L’étude s’appuie sur des simulations de dynamique moléculaire utilisant un nouveau potentiel semiempirique à charge variable, le SMTB-QB. Les travaux débutent par une validation approfondie de ce potentiel, démontrant sa capacité à reproduire des propriétés de l’UO2, du Cr2O3, du MoO2 et du Cs2O. L’analyse de la diffusion dans l’UO2 révèle deux régimes distincts pour l’oxygène et confirme le mécanisme lacunaire pour l’uranium. Dans l’UO2 dopé au chrome, les résultats montrent que le chrome se substitue préférentiellement à l’uranium, accélérant la diffusion de ce dernier d’un facteur 1,5 à 2 par rapport à l’UO2 non dopé. Le chrome lui-même diffuse plus rapidement que l’uranium dans ce système. Concernant les produits de fission étudiés, le césium s’incorpore préférentiellement dans les lacunes d’uranium, mais sa diffusion ne semble pas significativement accélérée dans l’UO2 dopé au chrome. Le molybdène, quant à lui, s’incorpore de manière préférentielle dans le complexe VUVO et présente un mécanisme de diffusion inattendu impliquant des défauts de Schottky liés
In the context of the global nuclear energy revival, driven by decarbonization and energy security challenges, the issue of nuclear safety is more crucial than ever. Following the Fukushima Daiichi accident in 2011, there has been renewed interest in Accident Tolerant Fuels (ATF), aimed at improving the safety and performance of nuclear reactors. Among these innovations, chromium-doped UO2 has emerged as a promising option, demonstrating better fission gas retention. However, the precise mechanisms by which chromium doping influences fission product behavior remain poorly understood, particularly regarding their diffusion within the material. This thesis explores the migration mechanisms of molybdenum and cesium in chromiumdoped uranium dioxide (UO2), compared to undoped UO2. The study relies on molecular dynamics simulations using a new variable-charge semi-empirical potential, SMTB-QB. The work begins with a thorough validation of this potential, demonstrating its ability to reproduce properties of UO2, Cr2O3, MoO2, and Cs2O. The analysis of diffusion in UO2 reveals two distinct regimes for oxygen and confirms the vacancy mechanism for uranium. In chromium-doped UO2, the results show that chromium preferentially substitutes for uranium, accelerating the latter’s diffusion by a factor of 1.5 to 2 compared to undoped UO2. Chromium itself diffuses faster than uranium in this system. Regarding the studied fission products, cesium preferentially incorporates into uranium vacancies, but its diffusion does not seem significantly accelerated in chromium-doped UO2. Molybdenum, on the other hand, preferentially incorporates into the VUVO complex and exhibits an unexpected diffusion mechanism involving bound Schottky defects