Tesi sul tema "MOX (combustibles nucléaires) – Solubilité"
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Thomas, Régis. "MOX dopé chrome : optimisation du dopage et de l’atmosphère de frittage". Thesis, Bordeaux 1, 2013. http://www.theses.fr/2013BOR14832/document.
Optimal use of the Mixed Oxide (U,Pu)O2 nuclear fuel in pressurized water reactors is mainly limited by the behavior of gaseous fission produced during irradiation. Within the MOX microstructure, the probability of fission gas release is increased by the presence of rich localized plutonium areas exhibiting a higher local burn-up. A solution consists in optimizing plutonium distribution within the industrial product and promoting the crystalline growth of the fuel grains. For this purpose, addition of chromium sesquioxide during the manufacturing process is currently considered. A previous thesis has shown that the best results are obtained for a Cr addition slightly greater than the solubility limit of Cr in (U,Pu)O2. In order to explain the enhanced plutonium homogeneity, the author highlighted the formation of PuCrO3 precipitates at grain boundaries. A sintering model under reducing atmosphere, with chromium addition, was proposed. However, several points have to be more thoroughly investigated, especially regarding the solubility limit of chromium, as well as the optimal conditions of PuCrO3 precipitates formation. In a first part, speciation of solubilized and precipitated chromium in the mixed oxide (U,Pu)O2 is studied using electron probe microanalysis (EPMA) and X-ray absorption spectroscopy (XAS). It was shown that the oxidation state and the environment of soluble chromium within the (U,Pu)O2 matrix do not depend on the oxygen partial pressure during sintering, neither on the plutonium content of the mixed oxide. However, both chemical nature of the precipitates and chromium solubility depend on the thermodynamic variable and on the plutonium content.Based on these results, a chromium solubility model in the mixed oxide (U,Pu)O2-x was built using the law of mass action governing solubility equilibrium. This model is described as a function of the plutonium content (y) of the solid solution (U1-yPuy)O2-x (y = 0,11 ; 0,275 et 1) and in the oxygen potential range of interest for MOX fuel sintering (-445 kJ/mol < µO2 < -360 kJ/mol). This thermodynamic model contributes to the optimization of the doping stage of fabrication and defines the optimal conditions of PuCrO3 precipitates formation.The aim of the second part is to verify if the thermodynamic conditions of PuCrO3 formation correspond to an optimal plutonium distribution and grain growth of the mixed oxide. Samples manufactured with and without Cr2O3 addition and sintered under various atmospheres were analyzed. It was shown that the U-Pu interdiffusion kinetics is completely modified with chromium addition. Morover, with chromium addition, sintering conditions which increase the U-Pu interdiffusion kinetics, don’t necessarily correspond to optimal grain growth. Based on these results, recommendations for the industrial manufacturing process are proposed. They deal with the choice of the sintering atmosphere and doping concentration to obtain an optimized microstructure
Garzón, Losik Germán Alexander. "Étude et modélisation d’un procédé de dissolution poussée en réacteur continu – application aux oxydes (U, Pu)O₂". Electronic Thesis or Diss., Université de Lorraine, 2023. http://www.theses.fr/2023LORR0050.
Plutonium multirecycling aims to stabilise the plutonium inventory and eventually close the fuel cycle in France. This action involves the reprocessing and recycling of plutonium-rich Mixed OXide (MOX) spent fuel, which implies the adaptation of current technologies, in particular dissolution. Therefore, an experimental study of the reaction between uranium-plutonium mixed oxides and nitric acid coupled with chemical reactor modelling is required. In this context, a study of the dissolution of three mixed oxides (30, 40, and 65% Pu/(U+Pu)) and plutonium dioxide in nitric acid is carried out using an optical setup. It can be highlighted that the mixed oxide dissolves according to a similar mechanism as plutonium dioxide when the plutonium content in the solid exceeds 30%. Moreover, only the mixed oxide with the highest uranium content undergoes an autocatalytic reaction mechanism similar to that identified for uranium dioxide. A dissolution model describing the size evolution of a single particle against time was developed. Such model considers the description of particles surface by a fractal geometry approach as well as the surface where the reaction actually takes place. Model was validated by comparison of experimental data from this work and from literature. In addition, a second model was developed, taking into account the single particlemodel, based on population balance equations. The model allows to describe the behaviour of a fluidised bed dissolver, which presents interesting advantages for solid-fluid type reactions. Finally, by implementing the developed model, simulations were performed showing a first estimation of the feasibility of a new dissolution process for current and upcoming MOX fuels
Toury, Grégoire. "Maîtrise de microstructures MOX de type CERCER". Limoges, 2001. http://www.theses.fr/2001LIMO0014.
Since fifteen years, the nuclear fuel MOX, blending of uranium dioxide UO2 and plutonium dioxide PuO2, allow to recycle plutonium derived from the reprocessing of the irradiated fuel UO2. With the aim in view to reach high burnup (70 GWj/t aimed), and particularly to restrain the fission gas release, this work pay particular attention to the realisation of materials with model microstructure
Roussette, Sophie. "Analyse par champs de transformation de matériaux élastoviscoplastiques multiphases : application aux combustibles MOX". Aix-Marseille 2, 2005. http://www.theses.fr/2005AIX22054.
The description of the overall behavior of nonlinear materials with nonlinear dissipative phases requires an infinity of internal variables. An approximate model involving only a finite number of internal variables, Nonuniform Transformation Field Analysis, is obtained by considering a decomposition of these variables on a finite set of nonuniform transformation fields, called plastic modes. The method is initially developed for incompressible elastoviscoplastic materials. Karhunen-Loève expansion is proposed to optimize the plastic modes. Then the method is extended to porous elastoviscoplastic materials. Finally the transformation field analysis, developed by Dvorak, is applied to nuclear fuels MOX. This method enables to make sensitivity studies to determine the role of some microstructural parameters on the fuel behaviour. Moreover the adequacy of the nonuniform method for fuels MOX is shown, the final objective being to be able to apply the model to the MOX in 3D
Oudinet, Ghislain. "Analyse d'images et modélisation 2D/3D de la microstructure du combustible MOX". Saint-Etienne, 2003. http://www.theses.fr/2003STET4011.
The microstructure of the MOX fuel, made with UO2 and PuO2, determines his " in pile " behavior. The french companies CEA and COGEMA are highly interested in its description by image analysis, which is the object of the present work. The segmentation algorithms described here use pictures issued from a microprobe and a SEM, to analyse the plutonium and porosity distribution in the fuel pellets. They are innovating, automated and robust enough to be used with a small data set. They have been successfully tested on different fuels, before and after irradation. Three-dimensional informations have been computed with a genetic algorithm. The obtained 3D object size distributions allowed the modeling of many different industrial and research fuels. 3D reconstruction is accurate and stable, and provides a basis for different studies among which the study of the MOX fuel " in pile " behavior
Bouloré, Antoine. "Etude et modélisation de la densification en pile des oxydes nucléaires UO2 et MOX". Grenoble INPG, 2001. http://www.theses.fr/2001INPG4203.
Amongst the many phenomena which take place in the course of the irradiation of UO2 or (U, Pu)O2 nuclear fuels, one of them involves the elimination of a fraction of the as-fabricated porosity. In-pile densification or sintering can reach 2. 5%, i. E. Approximately half the initial volume of pores is likely to disappear. Our literature survey indicates that the amplitude and kinetics of the phenomenon are both heavily dependent on the initial fuel microstructure. Micro-structural characterisation techniques of oxide fuels have therefore been developed in conjunction with quantitative image analysis methods. The ensuing methodology enables a quantitative comparison of micro-structural features in different fuels and has been applied to ascertaining the influence of the local fission rate and temperature on in-pile densification. It is thus revealed that in-pile operation eliminates a significant fraction of pores smaller than 3 microns in diameter. The experimental data generated has been used to set up a semi-empirical and a mechanistic model. The former is based on experimental results and is not essentially predictive. The inability of this model to predict the in-pile densification of oxide fuels is illustrated by the fact that the maximum fraction of pores that disappears is proportional to an empirical function of fission rate, and temperature. The proportionality factor appears to be difficult to correlate quantitatively to any given micro-structural feature. The model has however been applied to the interpretation of an in-pile densification experiment carried out in the Halden reactor (Norway). The latter model is mechanistic, i. E. It is based on the solution to a set of equations that describe the coupled temperature and radiation induced phenomena which occur in-pile. These can broadly be broken down into three categories : the fission fragment-pore interaction, the creation of point defects as the fission fragments slow down, and the diffusion of these point defects to sinks. The model calculates the evolution of the pore size distribution and has successfully been applied to modelling the in-pile densification behaviour of a fuel pellet characterised before and after irradiation
Mendez, Sandrine. "Etude de l'inerdiffusion U-Pu appliquée au combustible MOX". Aix-Marseille 3, 1995. http://www.theses.fr/1995AIX30036.
Théry, Odile. "Etude de la co-conversion uranium-cérium". Lille 1, 1999. https://pepite-depot.univ-lille.fr/LIBRE/Th_Num/1999/50376-1999-499.pdf.
Pieragnoli, Adrien. "Influence de l'adjuvant de frittage Cr2O3 sur l'homogénéisation de la répartition en plutonium au sein d'une pastille MOX hétérogène". Limoges, 2007. http://www.theses.fr/2007LIMO4061.
Detalle, Vincent. "Analyse de l'homogénéité du combustible nucléaire MOX par Spectrométrie d'Emission optique sur Plasma Induit par Laser (SEPIL)". Lyon 1, 1999. http://www.theses.fr/1999LYO10267.
Vaudez, Stéphane. "Frittage des combustibles MOX : influence des conditions d’élaboration et de la pression partielle d’oxygène". Thesis, Bordeaux, 2022. http://www.theses.fr/2022BORD0117.
Diffusion phenomena occurring during the sintering of mixed uranium and plutonium oxides (MOX) depend on the oxygen potential of the furnace atmosphere. The atmosphere and the temperature generate deviations from the oxygen stoichiometry, called O/M ratio, in the mixed oxide U1-yPuyO2±x. The PhD work focused on a better understanding of the evolution of the O/M ratio and the ceramic microstructure as a function of the composition of the UO2 / PuO2 / (UPu)O2 mixture and of the oxygen potential during sintering. Their effects on densification and microstructures have been studied and have led to proposing the implementation of a (UPu)O2 solid solution powder obtained by oxalic coprecipitation.An innovative monitoring of PO2 has been designed by stabilizing and controlling O2 input/output measurements of the ovens by oxygen pumps and zirconia probes. It allows a much better understanding of the oxygen exchanges in temperatures between the sample and the atmosphere. A significant difference between the thermodynamic calculations and experimental data is observed, even after several hours at 1700°C. The kinetic effects are particularly sensitives to the formation of the (UPu)O2 solid solution as well as to local solid/gas exchanges. A reaction scheme for varying the O/M ratio over time, by voluntarily varying the PO2 of the input gas makes it possible to design sintering cycles in order to obtain the desired microstructure and final O/M ratio.This new knowledge will be used as input data for the modelling of the sintering step. New sintering cycles in batch oven or in continuous oven could then be considered. They make possible to obtain new products and / or process gains. These advantages can be coupled with the beneficial contribution of the presence of a solid solution as a new raw material
Pizette, Patrick. "Simulation de la compaction de poudres céramiques par la méthode des éléments discrets : application à la mise en forme des combustibles nucléaires mixtes". Grenoble INPG, 2009. http://www.theses.fr/2009INPG0190.
Nuclear mixed oxide fuel (MOX) is formed as pellets by cold compaction of a powder blend of oxides of uranium and plutonium followed by sintering. Because of the variability of incoming powders in the process and taking into account the need to maintain constant industrial settings for processing significant quantities of pellets, the compaction process may generate some rejection during controls. In particular, compact strength and dimensional precision are key features of the industrial process. The Discrete Element Method (DEM), which has been used here, offers a powerful tool for understanding and simulating the compaction stage. It relies on an explicit modeling of the particulate nature of the uranium oxide powders. Two models, at the length scale of the aggregate of crystallites and at the length scale of the crystallites are used to simulate the powder compaction. A methodology, based on numerical experiments, is proposed to generate constitutive laws to feed a finite element code. Finally, the modeling at the crystallite length scale is used to identify the main microstructural parameters that control the compact strength
Lozano, Nathalie. "La subdivision d'un solide induite par l'évolution de sa composition chimique : intérêt pour la céramique nucléaire a fort taux d'irradiation". Dijon, 1998. http://www.theses.fr/1998DIJOS067.
Seck, Mohamed El Bachir. "Modélisation du comportement effectif de milieux hétérogènes viscoélastiques, non linéaires, vieillissants : application à la simulation du comportement des combustibles MOX". Thesis, Aix-Marseille, 2018. http://www.theses.fr/2018AIXM0407/document.
The prediction of the macroscopic mechanical behavior of heterogeneous materials from the properties of their constituents is possible for various classes of behavior (elastic, viscoelastic, etc.) thanks to the theory of homogenization. Nevertheless, the extension of this theory for materials with a non-linear (or elasto-viscoplastic) viscoelastic behavior remains an open question that we are tackling in this work in order to predict the macroscopic behavior of uranium-plutonium (MOX) mixed oxide fuels used in french pressurized water reactors (PWRs). From this perspective analytical and purely numerical solutions have been obtained and the model adopted is used to simulate the behavior of fuels
Lefrançois, Lydie. "Analyse et nouvelle approche prédictive du phénomène de formation d'une troisième phase dans les systèmes d'extraction (akylmalonamide-hydrocarbure-acide-eau)". Nancy 1, 1999. http://www.theses.fr/1999NAN10258.
Le, Gall Claire. "Contribution à l'étude du relâchement des produits de fission hors de combustibles nucléaires en situation d'accident grave : effet de la pO2 sur la spéciation du Cs, Mo et Ba". Thesis, Université Grenoble Alpes (ComUE), 2018. http://www.theses.fr/2018GREAY053/document.
In the nuclear community, it is a top priority to gain in-depth understanding of fission product (FP) speciation mechanisms occurring in nuclear fuel in order to precisely estimate the source term of a severe accident. Among the FP produced, some are highly reactive and may have a strong radiological impact if released into the environment. This is particularly the case of cesium (Cs), molybdenum (Mo) and barium (Ba). In this context, the objective of this study is to provide experimental data on the effect of the oxygen potential on Cs, Mo and Ba speciation in nuclear fuels at different stages of a severe accident.A thermodynamic approach was coupled with the experimental work to support the interpretation of experimental data. Two types of samples were studied in detail: irradiated MOX fuels and simulated high burn-up UO2 fuels produced through sintering at high temperature (SIMFuel). The samples were submitted to thermal treatments in conditions representative of a pressurised water reactor (PWR) severe accident. This approach made it possible to cover a temperature range from 400°C up to 2530°C and oxygen potentials from -470 kJ.mol(O2)-1 to -100 kJ.mol(O2)-1. The samples were characterized before and after each test using complementary techniques like OM, SEM, EPMA and SIMS in the case of irradiated fuels. XANES measurements using synchrotron radiation facilities were performed on SIMFuels and provided valuable results on FP speciation. Moreover, spark plasma sintering (SPS) was successfully investigated for the production of SIMFuel samples containing Cs, Mo and Ba in a chemical state representative of PWR fuel in normal operating conditions.This work highlighted the effect of oxidizing severe accident conditions on the fuel and FP behavior. Oxidation of Mo initially contained in the fuel’s metallic inclusions into MoO2 was observed to take place around 1000°C in oxidizing conditions. An interaction between MoO2 and the oxide phase containing Ba took place in the same conditions, leading to the formation of BaMoO4. The oxygen potential also plays an important role in fuel-cladding interactions, enhancing the diffusion of species in oxidizing conditions and lowering the temperature at which fuel melting occurs
Chambon, Cébastien. "Densification et homogénéisation U/Pu au cours du frittage de combustibles oxydes mixtes élaborés à partir de poudres UO2, U3O8 et PuO2". Thesis, Bordeaux, 2017. http://www.theses.fr/2017BORD0847.
In order to manufacture mixed-oxide fuels, also known as MOX ((U,Pu)O2) for the next generation of nuclear reactors, the use of triuranium octoxide (U3O8) was considered in this study. This PhD work focuses on the impact of this addition on MOX sintering and on the dimensional stability of sintered pellets during annealing. Initial experiments revealed a de-densification phenomenon at high temperature in the pellets containing U3O8 synthesized from an oxalic route.This undesirable phenomenon was studied on an inactive surrogate: a cerium oxide synthesized from an oxalic route in order to develop experimental techniques and protocols. The relationship between the presence of carbon impurities in the powders and the de-densification phenomenon was proven. Moreover, this de-densification phenomenon was observed in situ for the first time by using X-ray microtomography during sintering.The study of MOX fuels confirmed the major role of carbon impurities. The microstructural evolutions, the quantification of the carbon species released during sintering and the analysis of gases trapped inside the porosity of the sintered material led to the determination of a de-densification mechanism. Finally, a thermomechanical modelling of the fuel behavior under the effect of pore pressurization allows consolidating the proposed mechanism. Based on these results, a new sintering cycle was proposed and the first trials successfully limited the impact of the de-densification phenomenon
Bonev, Plamen. "Thermal conductivity of mixed oxide fuel (MOX) : effect of temperature, elementary chemical composition, microstructure and burn-up in reactor". Electronic Thesis or Diss., Université de Lorraine, 2023. http://www.theses.fr/2023LORR0367.
Mixed oxide fuel (MOX) is the nuclear fuel, used in fourth generation reactors, also called fast neutron reactors (FNR). Those reactors operate at very high temperatures (between 1500 and 2500 K). Thermal conductivity is therefore an essential material property to reactor safety. In fast reactor operating conditions, MOX is not only subject to high temperatures, but also to local changes in chemical composition and microstructure, which can have great impact on thermal conductivity. The effect of plutonium content is of particular interest for FNR applications, not only due to its local changes during irradiation, but also because fast reactors can be used to recycle plutonium. Thermal conductivity models should therefore be predictive in a wide range of plutonium contents. Most modeling approaches are semi-empirical in their temperature-dependency description of thermal conductivity, and are purely empirical in terms of plutonium and oxygen content-dependency. Those approaches are therefore limited by the number of available experimental data, especially concerning high temperatures (above 2000 K) and high plutonium contents (above 30 at. % ). The extrapolation of those models beyond their experimental range of validity can therefore lead to high modeling uncertainties. To address this problem, we propose in this work a model built on physical foundations. This model is based on a theoretical assessment of the contribution to thermal conductivity of each of the three (quasi)particles responsible for heat transport in oxide fuels: phonons, polarons and photons. The effect of temperature, plutonium and oxygen content on thermal conductivity is therefore clearly identified. Plutonium-oxygen content correlated effects were in particular observed, which do not appear in empirical approaches. This work also allowed to improve the understanding of irradiation-induced effects on thermal conductivity in FNR irradiation conditions. The model, proposed in this work was compared to the most up-to-date experimental data on thermal conductivity of MOX fuels, counting a total of 6619 experimental points, originated from different institutions: CEA, European projects, IAEA, OECD. Experimental data confirmed the effect of plutonium content, predicted in this work and in particular provided an experimental evidence for the plutonium-oxygen content correlated effects. The model was implemented into the CEA fuel performance code GERMINAL, from the simulation software platform PLEAIDES, to simulate the fuel behavior during the INTA-2 irradiation experiment. The predicted fuel temperature was compared to thermocouple measurements and showed good consistency, highlighting the adequate use of our model in fuel performance codes
Costenoble, Sylvain. "Modélisation de la coprécipitation d'oxalates mixtes d'uranium et de plutonium dans le cadre du recyclage du combustible nucléaire : solubilité des solutions solides oxalate". Thesis, Lille 1, 2009. http://www.theses.fr/2009LIL10179/document.
Chemical processes for future spent fuel treatment and recycling for new generation facilities are oriented on uranium and plutonium co-management. In the COEX TM process, one of the key operation consists of uranium(IV)-plutonium(III) co-conversion by oxalic co-precipitation. This leads to a mixed oxide for fuel production via the synthesis and the calcination of an oxalate solid solution precursor. Oxalic coprecipitation modeling, support of a better understanding of the process, is based on the supersaturation calculation whose expression is a function of thermodynamic data related to the oxalate solid solution formation.These thermodynamic data are acquired thanks to solubility measurements realised on uranium(IV)-neodymium(III) oxalate solid solutions, where neodymium(III) simulates the plutonium(III) behaviour. From the development of an actinide speciation analytical method, the non congruent interaction between the solid solution phase and the aqueous phase is observed. On the basis of the reaction constants of the occurring equilibria, a model, extracted from the LIPPMANN’s theory, allows to predict the state of the solid solution-aqueous solution system at thermodynamic equilibria. This methodology was extended to mixed oxalates U(IV)-Pu(III) and U(IV)-Am(III) demonstrating the model validity for these systems
Kauric, Guilhem. "Contribution to the investigation of the chemical interaction between sodium and irradiated MOX fuel for the safety of Sodium-cooled Fast Reactors". Thesis, université Paris-Saclay, 2020. http://www.theses.fr/2020UPASF027.
In case of a severe accident in Sodium-cooled Fast Reactors, interactions between partly molten fuel and sodium could happen at high temperature. Therefore, to predict the degradation evolution of fuel pins and phase formation in the different systems existing in the irradiated fuel, a thorough study of the Na-FP-Pu-U-O with FP= Ba, Cs, I, Mo, Te has to be performed. For such multicomponent system and large temperature and composition range, the Calphad method is a suitable way for developing a thermodynamic database to predict the phase formation depending on the temperature, pressure and composition of the system. Compositions with four Pu/(U+Pu) ratio in the Na-O-Pu-U system were synthesised by solid state synthesis method using nanoparticules of MOX fuel and characterised by XRD, ²³Na-NMR and HR-XANES techniques. The oxidation state of plutonium and uranium in the different structures was systematiquely investigated. When the measured oxidation state of actinides was different from the theoretical one, charge compensation mechanisms were suggested either by adding sodium in the structure or oxygen vacancies. Then, the structure of quaternary compounds in the Ba-Mo-Na-O and Cs-Mo-Na-O systems were also investigated by several structural techniques (XRD, neutron diffraction, HT-XRD, HT-Raman spectroscopy, XAS). Thermodynamic properties such as standard enthalpy of formation or enthalpy of decomposition were also determined. Finally, the Cs₂MoO₄-Na₂MoO₄ pseudo-binary section was re-investigated experimentally by DSC and a Calphad model for this system was developed
Kerleguer, Valentin. "Apport de l'étude de matériaux modèles U1-xPuxO2 à la compréhension des mécanismes d'altération des combustibles UOx et MOx en stockage géologique". Electronic Thesis or Diss., Université Paris sciences et lettres, 2020. http://www.theses.fr/2020UPSLM061.
The effects of the geological disposal environment on the leaching of the oxide matrices of UOx and MOx fuels were investigated following a step-by-step procedure: carbonated water, synthetic porewater of the Collovo-Oxfordian (COx) argillite, synthetic porewater in presence of iron or argillite samples. Two types of alpha-emitting materials were considered, UO2 pellets doped with a low Pu content, and pellets of homogenous U0.73Pu0.27O2 MOx fuel. The experimental protocols did not show any significant effect of the argillite on the UO2alteration. Plutonium decreased the oxidative dissolution of U0.73Pu0.27O2 and enhanced the disproportionation of the H2O2 produced by water radiolysis. The dissolution of the MOx matrix decreased in COx water. It was strongly inhibited in presence of iron which anoxic corrosion liberated Fe2+ in solution that fully reacted with the radiolytic H2O2, leading to magnetite precipitation on the pellet surface. Geochemical (CHESS code) and reactive transport (HYTEC code) models, which were developed for the homogeneous MOx alteration,correctly simulated the main experimental data and the underlying mechanisms. The alteration processes of UOx and MOx matrices were found to be very similar under the present environmental conditions
Odorowski, Mélina. "Etude de l'altération de la matrice (U,Pu)O2 du combustible irradiéen conditions de stockage géologique : Approche expérimentale et modélisation géochimique". Electronic Thesis or Diss., Paris, ENMP, 2015. http://www.theses.fr/2015ENMP0057.
To assess the performance of direct disposal of spent fuel in a nuclear waste repository, researches are performed on the long-term behavior of spent fuel (UOx and MOx) under environmental conditions close to those of the French disposal site. The objective of this study is to determine whether the geochemistry of the Callovian-Oxfordian (COx) clay geological formation and the steel overpack corrosion (producing iron and hydrogen) have an impact on the oxidative dissolution of the (U,Pu)O2 matrix under alpha radiolysis of water.Leaching experiments have been performed with UO2 pellets doped with alpha emitters (Pu) and MIMAS MOx fuel (un-irradiated or spent fuel) to study the effect of the COx groundwater and of the presence of metallic iron upon the oxidative dissolution of these materials induced by the radiolysis of water. Results indicate an inhibiting effect of the COx water on the oxidative dissolution. In the presence of iron, two different behaviors are observed. Under alpha irradiation as the one expected in the geological disposal, the alteration of UO2 matrix and MOx fuel is very strongly inhibited because of the consumption of radiolytic oxidative species by iron in solution leading to the precipitation of Fe(III)-hydroxides on the pellets surface. On the contrary, under a strong beta/gamma irradiation field, alteration tracers indicate that the oxidative dissolution goes on and that uranium concentration in solution is controlled by the solubility of UO2(am,hyd). This is explained by the shifting of the redox front from the fuel surface to the bulk solution not protecting the fuel anymore. The developed geochemical (CHESS) and reactive transport (HYTEC) models correctly represent the main results and occurring mechanisms
Takoukam, Takoundjou Cyrille. "Etude du diagramme de phases du combustible MOX par simulation moléculaire de type Monte Carlo". Thesis, Aix-Marseille, 2020. http://www.theses.fr/2020AIXM0321.
Uranium plutonium mixed oxide, commonly called MOX (for Mixed Oxide), is currently used in the Light Water Reactors (LWR) and is envisaged as the reference fuel for the future 4th generation french Fast Breeder Reactors (FBR). During the manufacturing of this fuel, the oxygen hypo-stoichiometry (O/M ratio < 2 with M = U+Pu) is imposed. Knowledge of the characteristics of MOX requires the strong understanding of the U‒Pu‒O ternary phase diagram. Particularly the characterisation of the miscibility gap highlighted exclusively in the hypo-stoichiometric region and more precisely within the region interest of the FBR fuel (UO2‒PuO2‒Pu2O3 phase diagram). Our study consists in using the molecular Monte Carlo (MC) method with its cation exchange algorithm. The main objective is to improve the knowledge about the MOX ternary phase diagram. To do this, it is appropriate to calculate first the thermodynamic properties of this fuel. Thus, we worked along two main axes. The first part of this work consisted in implementing the MOX simulation through the MC method and determining the structural, thermodynamic and mechanical properties of stoichiometric U1-yPuyO2 MOX as well as those of pure oxides UO2 and PuO2 missing or poorly known. The second part consist in studying the hypo-stoichiometric MOX. We first optimized our own potential, which appeared to be more efficient than the unique potential available in the literature. Next, we calculated several properties of interest of the hypo-stoichiometric MOX both outside and inside the miscibility gap. The marks of a possible phase separation in the hypo-stoichiometric MOX as well as the demixing limit temperature were highlighted
Hibert, Nicolas. "Synthèse et caractérisation structurale des complexes de plutonium à base de peroxyde". Thesis, Lille, 2020. http://www.theses.fr/2020LILUR054.
In the framework of the improvement of the reused nuclear fuel manufacturing and the Pu multirecycling, U-Pu coconversion represent a potential alternative to the current mixing process of uranium and plutonium oxide powders. Compared to carbon-based U-Pu coconversion processes, the peroxide process has the advantage of, among others, leading to the absence of residual carbon in oxide powder. However, the current knowledge of plutonium peroxide is incomplete and scattered, hindering the plutonium conversion and U-Pu coconversion technological developments. Thus, the evaluation of the feasibility of this process requires a preliminary strengthening of the knowledge of plutonium peroxide physico-chemical properties. The first part of this work has been dedicated to the characterization of plutonium peroxide soluble complexes and salts. Molar extinction coefficient of soluble complexes have been estimated in order to quantify the plutonium loss in precipitation experiments. The experimental conditions enabling very high yield of precipitation of Pu and an easy-filterable powder have been determined. Moreover, a new database of plutonium peroxide salt properties has been established from the characterizations obtained. In the second part, syntheses carried out with mixed systems such as U-Pu and U-Th have led to obtaining a mix of uranyl peroxide and +IV actinide peroxide salt. The experimental conditions enabling very high yields for uranium and for plutonium and an easy-filterable powder have been determined. Then, thermal treatment of the precipitate has led to obtaining an oxide powder with a good ability to the manufacturing of sintered oxide pellet and which enables to demonstrate the feasibility of U-Pu conversion process at the laboratory scale
Costin, Dan Tiberiu. "Solutions solides d'uranothorites : De la préparation à la dissolution". Thesis, Montpellier, Ecole nationale supérieure de chimie, 2012. http://www.theses.fr/2012ENCM0026/document.
USiO4 coffinite appears as one of the potential phases formed in the back-end of the alteration of spent fuel, in reducing storage conditions. A study aiming to assess the thermodynamic data associated with coffinite through an approach based on the preparation of Th1-xUxSiO4 uranothorite solid solutions was then developed during this work. First, the preparation of uranothorite samples was successfully undertaken in hydrothermal conditions. However, the polyphased samples systematically formed for x > 0,2 underlined the kinetic hindering linked with the preparation of uranium-enriched samples, including coffinite end-member. Nevertheless, the characterization of the various samples led to confirm the formation of an ideal solid solution and allowed the constitution of a spectroscopic database. The purification of the samples was then performed by the means of different protocoles based on physical (dispersion-centrifugation) or chemical (selective dissolution of secondary phases) methods. This latter led to a complete of the impurities (Th1-yUyO2 mixed oxide and amorphous silica) through successive washing steps in acid then basic media. Finally, dissolution experiments were undertaken on uranothorite samples (0 ≤ xexp. ≤ 0,5) and allowed pointing out the influence of composition, pH and temperature on the normalized dissolution rate of the compounds. Also, the associated thermodynamic data, such as activation energy, indicate that the reaction is controlled by surface reactions. Once the equilibrium is reached, the analogous solubility constants were determined for each composition studied, then allowing the extrapolation to coffinite value. It was then finally possible to conclude on the inversion of coffinitisation reaction with temperature
Truphemus, Thibaut. "Etude des équilibres de phases en fonction de la température dans le système UO2-PuO2-Pu2O3 pour les céramiques nucléaires aux fortes teneurs en plutonium". Thesis, Aix-Marseille, 2013. http://www.theses.fr/2013AIXM4303/document.
In the UO2-PuO2-Pu2O3 section, a monophasic (U1-y,Puy)O2-x domain is stable for y<0,20 at 25°C and up to solid-liquid equilibrium. At higher Pu content, phase equilibria are more unclear with a phase separation process. The main objective of this work consisted in upgrading the representation of this system for 0,15≤y≤0,65 and 25≤T(°C)≤1500.At 25°C, a miscibility gap composed by two different (U1-y,Puy)O2-X phases has been observed for y<0,45, with one very closed to stoichiometric state (Oxygen/Metal=2) and one other very reduced. For the first time, a triphasic domain has been characterized at higher Pu contents, with two (U1-y,Puy)O2-X phases near y=0,45 and one (U1-y,Puy)2O3 phase with a low U content inside. Concerning the study in function of temperature, we have demonstrated that phase separation temperature increase when Pu content grows. Several representations have been established. At 200°C, the representation is closed to that at 25°C. At 400°C, the phase separation have been specified at a lower Pu content than that of literature : y=0,35. At 600°C, our results have clarified the section, until then very unclear, with a phase separation appearing at y=0,60.The microstructural analysis has clearly demonstrated the significant impact of the phase separation on the material. Indeed many cracks have been observed in our samples, and quantity of these defects increases when Pu content grows
Odorowski, Mélina. "Etude de l'altération de la matrice (U,Pu)O2 du combustible irradiéen conditions de stockage géologique : Approche expérimentale et modélisation géochimique". Thesis, Paris, ENMP, 2015. http://www.theses.fr/2015ENMP0057/document.
To assess the performance of direct disposal of spent fuel in a nuclear waste repository, researches are performed on the long-term behavior of spent fuel (UOx and MOx) under environmental conditions close to those of the French disposal site. The objective of this study is to determine whether the geochemistry of the Callovian-Oxfordian (COx) clay geological formation and the steel overpack corrosion (producing iron and hydrogen) have an impact on the oxidative dissolution of the (U,Pu)O2 matrix under alpha radiolysis of water.Leaching experiments have been performed with UO2 pellets doped with alpha emitters (Pu) and MIMAS MOx fuel (un-irradiated or spent fuel) to study the effect of the COx groundwater and of the presence of metallic iron upon the oxidative dissolution of these materials induced by the radiolysis of water. Results indicate an inhibiting effect of the COx water on the oxidative dissolution. In the presence of iron, two different behaviors are observed. Under alpha irradiation as the one expected in the geological disposal, the alteration of UO2 matrix and MOx fuel is very strongly inhibited because of the consumption of radiolytic oxidative species by iron in solution leading to the precipitation of Fe(III)-hydroxides on the pellets surface. On the contrary, under a strong beta/gamma irradiation field, alteration tracers indicate that the oxidative dissolution goes on and that uranium concentration in solution is controlled by the solubility of UO2(am,hyd). This is explained by the shifting of the redox front from the fuel surface to the bulk solution not protecting the fuel anymore. The developed geochemical (CHESS) and reactive transport (HYTEC) models correctly represent the main results and occurring mechanisms
Gupta, Florence. "Etude du comportement du produit de fission césium dans le dioxyde d’uranium par méthode ab initio". Paris 11, 2008. http://www.theses.fr/2008PA112129.
The knowledge of the behaviour of fission products in the nuclear fuel is very important for safety considerations and for understanding the evolution of the fuel properties under irradiation. In this work, we focussed mainly on the behaviour of caesium in UO2 through ab initio studies of its solubility at point defects in the matrix, its diffusion and its contribution to the formation of solid phases in the fuel. The role of electronic correlation effects of the f electrons of uranium on these properties and on the description of the defect free crystal, is assessed. The formation energies of the main point defects are calculated and their concentration as a function of fuel stoichiometry and temperature is estimated. The migration barriers and migration paths for the self-diffusion of oxygen and uranium vacancies and oxygen interstitials in UO2 are discussed. The solubility of Cs is found to be very low in UO2 in agreement with experimental findings. The most favourable trapping sites are determined as a function of oxygen concentration in the fuel. Our results show that in the hyper-stoichiometric regime, the diffusion of Cs from its most favourable trapping site is limited by the uranium vacancy diffusion mechanism. We also considered the formation of the main solid phases of caesium resulting from its oxidation (Cs2O, Cs2O2, CsO2) and from its interaction with the fuel (Cs2UO4), with molybdenum (Cs2MoO4) and with the zirconium of the clad (Cs2ZrO3), since the formation of such phases, their solubility and their interdependence will affect the release of caesium
Nkou, Bouala Galy Ingrid. "Premier stade du frittage des dioxydes de lanthanides et d’actinides : une étude in situ par MEBE à haute température". Thesis, Montpellier, 2016. http://www.theses.fr/2016MONTT220/document.
Sintering is a key step in the elaboration of UOx and MOx (U/Pu mixed oxide) nuclear fuels pellets used in the pressurized water reactors. The first step of this process, which consists in the elaboration of a neck between the grains and led to the consolidation of the material, is generally described through numerical simulation. The models used for the theoretical description of this step are generally constituted by two spherical grains in contact. In order to perform the first experimental observations of the initial stage of sintering of ceramics materials of interest for electronuclear fuel cycle and to complement the numerical approaches, samples of lanthanide (CeO2) and actinides (ThO2 and UO2) dioxides with controlled morphology were examined by environmental scanning electron microscopy during heat treatment at high temperature (HT-ESEM).First, the protocols leading to the synthesis of lanthanides and actinides oxides microspheres were developed, and the powders obtained characterized. It was thus possible to obtain, for all the compounds studied, systems similar to those generally modeled. HT-ESEM was then used as the main investigation tool for the in situ study of the first stage of sintering of these compounds. The study of the morphological modifications occurring in isolated microspheres first confirmed their polycristalline character. Indeed, heat treatment led to a progressive decrease of the crystallites number included inside the grains through different mechanisms (oriented attachment, diffusion), whose activation energy was evaluated. For the systems constituted by two CeO2 or ThO2 microspheres in contact, the ESEM micrographs allowed to observe the evolution of several parameters during heat treatment, such as neck size and grain size as well as distance between the grains center. Images processing methods using custom software were then applied in order to determine the quantitative kinetic data. The mechanisms involved, such as the rearrangement of crystalline planes and the matter diffusion, and the corresponding activation energies, were also identified. Furthermore, the law of neck growth, which allows one to describe the evolution of sintering degree, was used to determine the prevailing diffusion mechanism during heat treatment. The influence of various parameters on the sintering degree was also highlighted. For example, the influence of grains polycristallinity on sintering mechanisms and kinetics was particularly investigated study by working in parallel with polycristalline and single crystal grains, then by comparing the experimental results with data coming from modeling. Finally, the methodology developed for the study of CeO2 and ThO2 was transposed to the compound of interest UO2. In this case, the data previously described were complemented by a first approach of the influence of atmosphere used during the heat treatment
Largenton, Rodrigue. "Modélisation du comportement effectif du combustible MOX : par une analyse micro-mécanique en champs de transformation non uniformes". Thesis, Aix-Marseille, 2012. http://www.theses.fr/2012AIXM4773/document.
Among the nuclear fuels irradiated in the Pressure Water Reactor of Électricité de France, MOX fuel is used, a Mixed OXide of plutonium and uranium. In this fuel, three phases with different plutonium content can be observed. The different fissile plutonium content in each phase leads different mechanical and physico-chemical evolutions under irradiation. To predict correctly the macroscopic behavior of MOX nuclear fuels in industrial nuclear fuel codes, models need to be fed in effective properties. But it's also interresting to obtain the local fields to establish coupling between mechanisms (mechanical and physico-chemical coupling). The aim of the PhD was to develop homogenisation method based on Non uniform Transformation Field Analysis (NTFA Michel and Suquet 2003}). These works were realised on three dimensions MOX microstructures and for local ageing visco-elastic behavior with free strains. The first work of the PhD was the numerical representation of the MOX microstructure in 3D. Three steps were realized. The first one consisted in the acquisition and the treatment of experimental pictures thanks to two soft-wares already developed. The second used the stereological model of textit{Saltykov} cite{R2S67} to go back up the two-dimensional statistical information in three-dimensional. And the last step was to develop tools which are able to build a numerical representation of the MOX microstructure. The second work of the PhD was to develop the NTFA model. Some theoretical (three dimensional, free strains and ageing hadn't ever studied) and numerical (choice and reduction of plastic modes, impact of the microstructures) studies were realised
Chevreux, Pierrick. "Comportement de l’uranium et de ses simulants dans les verres d’aluminosilicates en contact avec des métaux fondus". Thesis, Université de Lorraine, 2016. http://www.theses.fr/2016LORR0257/document.
This study concerns an innovative process used for conditioning nuclear waste that contain metallic parts contaminated with actinides. High actinides concentrations are expected to be incorporated in the glass melt in contact with the molten metals. Among these metals, aluminum and/or stainless steel impose a strongly reducing environment to the glass melt involving redox reactions. These reactions modify actinides oxidation states and therefore change their solubilities in the glass and could also reduce them into the metallic form. In this work, we focus on the behavior of uranium and its surrogates, namely hafnium and neodymium, in aluminosilicate glasses from the Na2O-CaO-SiO2-Al2O3 system melted in highly reducing conditions. The first step consists in comparing the hafnium and uranium solubilities in the glass as functions of redox conditions and glass composition. A methodology has been set up and a specific device has been used to control the oxygen fugacity and the alkali content of the glass. The results show that uranium is far less soluble in the glass than hafnium (HfIV) in reducing conditions. The uranium solubility ranges from 4 to 7 wt% UO2 for an oxygen fugacity below 10-14 atm at 1250°C-1400°C. Uranium oxidation states have been investigated by X-ray absorption spectroscopy (XANES). It has been pointed out that UIV is the main form in the glass for such imposed oxygen fugacities. The second step of this work is to identify the glass-metal interaction mechanisms in order to determine the localization of uranium and its surrogates (Nd, Hf) in the glass-metal system. Mechanisms are mostly ruled by the presence of metallic aluminum and are similar for uranium, neodymium and hafnium. Glass-metal interaction kinetics demonstrate that uranium and its surrogates can temporarily be reduced into the metallic form for particular conditions. A re-oxidation occurs with time which is in good agreement with thermodynamics. Regarding uranium, the re-oxidation process must be corroborated. Finally, the formation and dissolution processes of the different crystalline phases observed during these glass-metal interactions have been studied using a thermodynamic approach based on phase diagrams
Beaudoux, Xavier. "Dissolution réductrice d'oxydes de lanthanides et de PuO2 assistée par ultrasons". Thesis, Montpellier, 2015. http://www.theses.fr/2015MONTS046.
In the French nuclear program, the reactor fuel consists of uranium oxides or uranium plutonium mixed oxides (called MOX). Developments are constantly made on the resulting reprocessing of these fuels in order to optimize the recovery of reusable materials and to minimize the waste volume. In the case of MOX dissolution, the amount of Pu-rich dissolution residues is sometimes high despite the use of hard chemical conditions (oxidizing and corrosive). The difficulty to dissolve PuO2 batches declared non-standard during the fabrication of MOX can also be a technological barrier. In this context, sonochemistry can be considered as an alternative to current methods of dissolution of PuO2 or Pu enriched MOX. First, experiments of sonochemical dissolution were performed on an inactive analogue of PuO2, namely CeO2. The results were then used as a working basis for the dissolution of PuO2. Under reducing and acidic conditions, much milder than those used industrially, the complete dissolution of these two oxides was carried out within a few hours. Meanwhile, a related study showed that it is possible to completely dissolve lanthanide mixed oxides by a process of sonocatalytic and reductive dissolution in the presence of Pt. The dissolution rates increase with the trivalent lanthanide content within the oxide. Finally, the last part was devoted to the dissolution under magnetic stirring of Ce-based oxides in the presence or absence of noble metals, in weakly acidic media containing reducing natural molecules. Under these conditions, a complete, rapid and selective dissolution of these oxides was observed. These last two studies present an interest beyond the scope of nuclear chemistry, concerning the recycling of industrial materials (catalytic converters, fuel cells...)
Strach, Michal. "In situ studies of uranium-plutonium mixed oxides : Influence of composition on phase equilibria and thermodynamic properties". Thesis, Aix-Marseille, 2015. http://www.theses.fr/2015AIXM4044.
Due to their physical and chemical properties, mixed uranium-plutonium oxides are considered for fuel in 4th generation nuclear reactors. In this frame, complementary experimental studies are necessary to develop a better understanding of the phenomena that take place during fabrication and operation in the reactor. The focus of this work was to study the U Pu–O phase diagram in a wide range of compositions and temperatures to ameliorate our knowledge of the phase equilibria in this system. Most of experiments were done using in situ X-ray diffraction at elevated temperatures. The control of the oxygen partial pressure during the treatments made it possible to change the oxygen stoichiometry of the sample, which gave us an opportunity to study rapidly different compositions and the processes involved. The experimental approach was coupled with thermodynamic modeling using the CALPHAD method, to precisely plan the experiments and interpret the obtained results. This approach enabled us to enhance the knowledge of phase equilibria in the U–Pu–O system
Parmentier, Delphine. "Étude cinétique de la nucléation primaire et de la croissance cristalline au cours de la coprécipitation de solutions solides d’oxalates d’actinides". Thesis, Université de Lorraine, 2012. http://www.theses.fr/2012LORR0425.
Current concepts for future nuclear systems aim at improving the fuel cycle with the co management of actinides in order to enhance the fuel performance and to reduce the proliferation risk. Actinides coconversion processes play an important role by producing mixed actinides compounds used as starting materials for fuel re-fabrication. Oxalic coprecipitation is one investigated way to synthesize solid solutions of actinides – lanthanides mixed oxalates which have to meet strict standards. The nucleation and growth kinetic laws involve a fundamental crystallization parameter such as the supersaturation. For the precipitation of solid solutions, different theories are developed in the literature, however none have been verified experimentally. A new suitable expression for the supersaturation ratio applicable is presented in order to determine a general model for the expression of nucleation and growth rates. The experimental study of the primary nucleation kinetics is based on a “stopped flow” apparatus which provides a very good micromixing of the reactants. The kinetic laws of solid solutions verify the theory of Volmer and Weber applied to the coprecipitation. The method of calculating the supersaturation developed allows to find the typical behavior of nucleation. The experimental results demonstrate that the coprecipitation of the solid solution is kinetically favored over the precipitation of simple oxalates due to a lower energy barrier for the solid solution. The crystal growth rate is determined from a spectrophotometric monitoring of the reactant concentrations using a seed charge. The crystal growth is controlled by the surface integration with a spiral mechanism
Cheik, Njifon Ibrahim. "Modélisation des modifications structurales, électroniques et thermodynamiques induites par les défauts ponctuels dans les oxydes mixtes à base d'actinides (U,Pu)O2". Electronic Thesis or Diss., Aix-Marseille, 2018. http://www.theses.fr/2018AIXM0356.
(U,Pu)O2 (commonly called MOX) is currently used as nuclear fuel in pressurized water reactors with a Pu content of around 10 wt.%, and is envisaged as the reference fuel in Generation IV sodium fast reactors (SFR) with a Pu content of around 25 wt.%. Under operation, (U,Pu)O2 is submitted to fission reactions which generate a large quantity and variety of point defects, as well as fission products. By migrating, point defects and gaseous fission products can aggregate into nano-voids, dislocations and fission gas bubbles, which lead to the modification of the fuel microstructure. Therefore, a better description of the fuel behaviour at the atomic scale, and especially of the elementary mechanisms involved in the diffusion of point defects and fission products, is necessary to refine the models used in the fuel performance codes used to simulate the behaviour of fuels at the macroscopic scale. We use electronic structure calculations based on the DFT+U method combined with the occupation matrix control scheme (OMC) to investigate (U,Pu)O2 properties for various Pu contents. Static energy minimizations and ab initio molecular dynamics were used. We have first determined bulk structural, electronic and thermodynamics properties of (U,Pu)O2. We then studied the stability of point defects in (U,Pu)O2 and (U,Ce)O2, as well as the structural and electronic modifications induced by these point defects, in (U,Pu)O2 and the common experimental surrogate (U,Ce)O2. Finally, the fission gas (Kr and Xe) and helium (He) trapping and solubility in (U,Pu)O2 matrix are investigated
Morati, Nicolas. "Système de détection ultra-sensible et sélectif pour le suivi de la qualité de l'air intérieur et extérieur". Electronic Thesis or Diss., Aix-Marseille, 2021. http://www.theses.fr/2021AIXM0200.
Today the air is polluted by many chemicals, which are in the form of a complex mixture that is difficult to identify. These marker gases include carbon monoxide (CO), ozone (O3) and nitrogen dioxide (NO2). It has therefore become imperative to design detection systems that are inexpensive, but at the same time highly sensitive and selective, in order to monitor air quality in real time. Metal Oxide gas sensors (MOX) can meet these requirements. They are used in portable and low cost gas detection devices. Very sensitive, stable and with a long lifespan, MOX sensors suffer from an inherent lack of selectivity, which can be overcome by integrating artificial intelligence. This thesis is concerned with the implementation of gas identification methods based on the analysis of experimental data. The objective is to discriminate three pollution marker gases: CO, O3, and NO2, with a single sensor, under real conditions of use, i.e. in the permanent presence of a concentration of these gases in the humid ambient air. For this, we use a tungsten oxide (WO3) gas sensor patented by IM2NP laboratory and operated under a worldwide license by the company NANOZ.A complete experimental database was created from a protocol based on temperature modulation of the sensitive layer. From this database, we implemented two different feature extraction methods: the computation of temporal attributes and the wavelet transform. These two methods were evaluated on their gas discrimination capacity thanks to the use of several families of classification algorithms, such as support vector machines (SVM), decision trees, K nearest neighbours, neural networks, etc
Sogbadji, Robert. "Neutronic study of the mono-recycling of americum in PWR and of the core conversion INMNSR using the MURE code". Phd thesis, Université Paris Sud - Paris XI, 2012. http://tel.archives-ouvertes.fr/tel-00843688.
Cheik, Njifon Ibrahim. "Modélisation des modifications structurales, électroniques et thermodynamiques induites par les défauts ponctuels dans les oxydes mixtes à base d'actinides (U,Pu)O2". Thesis, Aix-Marseille, 2018. http://www.theses.fr/2018AIXM0356/document.
(U,Pu)O2 (commonly called MOX) is currently used as nuclear fuel in pressurized water reactors with a Pu content of around 10 wt.%, and is envisaged as the reference fuel in Generation IV sodium fast reactors (SFR) with a Pu content of around 25 wt.%. Under operation, (U,Pu)O2 is submitted to fission reactions which generate a large quantity and variety of point defects, as well as fission products. By migrating, point defects and gaseous fission products can aggregate into nano-voids, dislocations and fission gas bubbles, which lead to the modification of the fuel microstructure. Therefore, a better description of the fuel behaviour at the atomic scale, and especially of the elementary mechanisms involved in the diffusion of point defects and fission products, is necessary to refine the models used in the fuel performance codes used to simulate the behaviour of fuels at the macroscopic scale. We use electronic structure calculations based on the DFT+U method combined with the occupation matrix control scheme (OMC) to investigate (U,Pu)O2 properties for various Pu contents. Static energy minimizations and ab initio molecular dynamics were used. We have first determined bulk structural, electronic and thermodynamics properties of (U,Pu)O2. We then studied the stability of point defects in (U,Pu)O2 and (U,Ce)O2, as well as the structural and electronic modifications induced by these point defects, in (U,Pu)O2 and the common experimental surrogate (U,Ce)O2. Finally, the fission gas (Kr and Xe) and helium (He) trapping and solubility in (U,Pu)O2 matrix are investigated