Radiation-induced changes in microstructures often lead to significant changes in mechanical properties of alloys used in the construction of nuclear reactors. It is desirable to test small specimens to make efficient use of the small volumes available in test and commercial reactor cores and also because small specimens are less affected by the sometimes steep flux gradients experienced in reactor cores and the sometimes large temperature gradients developed in the specimens from gamma heating. (Continues...).

On développe une méthode de résolution du transfert radiatif au sein d'un milieu vapeur-gouttelettes entouré de parois chaudes, en vue d'un couplage avec une résolution de l'écoulement à l'échelle de la CFD. Le domaine d'application considéré est l'étude du refroidissement du coeur d'une centrale nucléaire suite à un Accident de Perte de Réfrigérant Primaire (APRP). Le problème du transfert radiatif se découpe en deux sous-problèmes, l'un concernant l'évaluation des propriétés radiatives du milieu et le second la résolution du transfert radiatif. Les propriétés radiatives ont été calculées avec la théorie de Mie pour les gouttelettes et le modèle C-k pour la vapeur d'eau. On obtient un milieu absorbant, diffusant anisotrope, émissif, non gris et non homogène. De plus, compte tenu de la grande gamme possible des propriétés de l'écoulement (diamètre et concentration des gouttelettes, température et pression de la vapeur), le milieu peut être optiquement fin ou optiquement épais. Il faut donc une méthode de résolution du transfert radiatif efficace pour toutes les conditions observées dans un APRP et ayant un temps de calcul raisonnable en vue du couplage avec les autres modes de transferts. La méthode IDA, dérivée de l'approximation P1, a été choisie. Son niveau de précision a été validé sur des cas tests académiques et une expérimentation. Des simulations en condition APRP ont ensuite été effectuées, permettant d'évaluer les flux rayonnés et confirmant que le transfert radiatif n'est pas négligeable dans cet accident / We developped a method of resolution of radiative transfer inside a medium of vapor-droplets surrounded by hot walls, in order to couple it with a simulation of the flow at the CFD scale. The scope is the study of the cooling of the core of nuclear reactor following a Loss Of Coolant Accident (LOCA). The problem of radiative transfer can be cut into two sub problems, one concerning the evaluation of the radiative properties of the medium and a second concerning the solution of the radiative transfer equation. The radiative properties of the droplets have been computed with the use of the Mie Theory and those of the vapor have been computed with a Ck model. The medium made of vapor and droplets is an absorbing, anisotropically scattering, emissive, non grey, non homogeneous medium. Hence, owing to the possible variations of the flow properties (diameter and volumetric fraction of the droplets, temperature and pressure of the vapor), the medium can be optically thin or thick. Consequently, a method is required which solves the radiative transfer accurately, with a moderate calculation time for all of these prerequisites. The IDA has been chosen, derived from the well-known P1-approximation. Its accuracy has been checked on academical cases found in the literature and by comparison with experimental data. Simulations of LOCA flows have been conducted taking account of the radiative transfer, evaluating the radiative fluxes and showing that radiative transfer influence cannot be neglected

No description available.

La modélisation des réacteurs nucléaires repose sur la résolution de l'équation de Boltzmann linéaire. Pour la résolution spatiale de la forme stationnaire de cette équation, le solveur MINARET utilise la méthode des éléments finis discontinus sur un maillage triangulaire non structuré afin de pouvoir traiter des géométries complexes. Cependant, l'utilisation d'arêtes droites introduit une approximation de la géométrie. Autoriser l'existence d'arêtes courbes permet de coller parfaitement à la géométrie, et dans certains cas de diminuer le nombre de triangles du maillage. L'objectif principal de cette thèse est l'étude d'éléments finis sur des triangles possédant plusieurs bords courbes. Le choix des fonctions de base est un des points importants pour ce type d'éléments finis. Un résultat de convergence a été obtenu sous réserve que les triangles courbes ne soient pas trop éloignés des triangles droits associés. D'autre part, un solveur courbe a été développé pour traiter des triangles avec plusieurs bords courbes. Une autre partie de ce travail porte sur l'accélération de la convergence des calculs. En effet, la résolution du problème est itérative et peut converger très lentement. Une méthode d'accélération dite DSA (Diffusion Synthetic Acceleration) permet de diminuer le nombre d'itérations et le temps de calcul. L'opérateur de diffusion est utilisé comme un préconditionneur de l'opérateur de transport. La DSA a été mise en oeuvre en utilisant une technique issue des méthodes de pénalisation intérieure. Une analyse de Fourier en 1D et 2D permet de vérifier la stabilité du schéma pour des milieux périodiques avec de fortes hétérogénéités / To model the nuclear reactors, the stationnary linear Boltzmann equation is solved. After discretising the energy and the angular variables, the hyperbolic equation is numerically solved with the discontinuous finite element method. The MINARET code uses this method on a triangular unstructured mesh in order to deal with complex geometries (like containing arcs of circle). However, the meshes with straight edges only approximate such geometries. With curved edges, the mesh fits exactly to the geometry, and in some cases, the number of triangles decreases. The main task of this work is the study of finite elements on curved triangles with one or several curved edges. The choice of the basis functions is one of the main points for this kind of finite elements. We obtained a convergence result under the assumption that the curved triangles are not too deformed in comparison with the associated straight triangles. Furthermore, a code has been written to treat triangles with one, two or three curved edges. Another part of this work deals with the acceleration of transport calculations. Indeed, the problem is solved iteratively, and, in some cases, can converge really slowly. A DSA (Diffusion Synthetic Acceleration) method has been implemented using a technique from interior penalty methods. A Fourier analysis in 1D and 2D allows to estimate the acceleration for infinite periodical media, and to check the stability of the numerical scheme when strong heterogeneities exist

Les isotopes d’américium représentent une part importante des déchets nucléaires à haute activité et à durée de vie longue dans le combustible usé. Parmi les options de retraitement envisagées, sa transmutation en réacteurs à neutrons rapides au sein de pastilles d’oxydes mixtes d’uranium-américium de composition U1-xAmxO2±δ est une option prometteuse qui permettrait de diminuer l’empreinte écologique des sites d’entreposage des déchets ultimes. Dans ce contexte, cette thèse est consacrée à l’étude de ces composés sur une large gamme de compositions (7,5 %mol ≤ Am/(U+Am) ≤ 70 %mol) focalisée sur leur fabrication à partir d’oxydes simples et l’évaluation de leur stabilités structurales, thermodynamique et sous auto-irradiation. Les résultats mettent en évidence l’influence majeure de la réduction de l’américium en Am+III, aussi bien dans les mécanismes de formation de la solution solide U1-xAmxO2±δ par voie solide que dans la stabilisation de cations d’uranium oxydés, accompagnés de la formation de défauts dans le sous-réseau d’oxygène tels que des lacunes et des clusters cuboctaédriques. Par ailleurs, les données acquises concernant la stabilité en température des composés U1-xAmxO2±δ (existence d’une lacune de miscibilité, comportement en vaporisation) ont été comparées à des calculs basés sur un nouveau modèle thermodynamique décrivant le système ternaire U-Am-O. Enfin, les effets structuraux de l’auto-irradiation α dans les composés U1-xAmxO2±δ ont été analysés par DRX, XAS et MET, permettant d’étudier l’influence de la teneur en américium sur le gonflement structural et de décrire l’évolution des défauts structuraux radio-induits. / Americium isotopes represent a significant part of high-level and long-lived nuclear waste in spent fuels. Among the envisaged reprocessing scenarios, their transmutation in fast neutron reactors using uranium-americium mixed-oxide pellets (U1-xAmxO2±δ) is a promising option which would help decrease the ecological footprint of ultimate waste repository sites. In this context, this thesis is dedicated to the study of such compounds over a wide range of americium contents (7.5 at.% ≤ Am/(U+Am) ≤ 70 at.%), with an emphasis on their fabrication from single-oxide precursors and the assessment of their structural and thermodynamic stabilities, also taking self-irradiation effects into account. Results highlight the main influence of americium reduction to Am+III, not only on the mechanisms of solid-state formation of the U1-xAmxO2±δ solid solution, but also on the stabilization of oxidized uranium cations and the formation of defects in the oxygen sublattice such as vacancies and cuboctahedral clusters. In addition, the data acquired concerning the stability of U1-xAmxO2±δ compounds (existence of a miscibility gap, vaporization behavior) were compared to calculations based on new thermodynamic modelling of the U-Am-O ternary system. Finally, α-self-irradiation-induced structural effects on U1-xAmxO2±δ compounds were analyzed using XRD, XAS and TEM, allowing the influence of americium content on the structural swelling to be studied as well as the description of the evolution of radiation-induced structural defects.

No description available.

Au cours de la vie d'une installation dans laquelle est utilisée de la matière radioactive, des étapes de maintenance, ou bien à terme de démantèlement, sont nécessaires afin d'assurer une protection convenable des travailleurs et de l'environnement. Parmi les procédés existants, des gels de décontamination ont été développés au CEA afin de pallier aux inconvénients des méthodes classiquement utilisées. Ces gels colloïdaux, formulés à l'origine de façon plutôt empirique, ont une structure qu'il convient de connaitre, celle-ci étant la clef de voute du comportement rhéologique du matériau. La façon dont ils s'écoulent est en effet un paramètre important du procédé, et les mesures en laboratoire qui sont effectuées sur ces systèmes permettent d'en améliorer l'implémentation. En outre, de nombreux raffinements du procédé, par le biais notamment d'additifs, en étendent le champ d'utilisation, et l'efficacité / During the exploitation of a nuclear plant, or all other installation which uses radioactive materials, maintenance tasks or decommissioning operations are mandatory in order to preserve people health and environment. Among existing processes, decontaminating gels have been development by CEA in order to overcome the drawbacks of the traditionally used methods. These colloidal gels were originally formulated in an empirical way; however, the knowledge of their structures is important as it rules all the rheological behaviors of the material. The way these gels flow is an important parameter to the process and our laboratory measurements can be transposed to the industrial world. Moreover, other composition refinements have been developed in order to extend their field of use and efficiency

This thesis is intended to contribute to policy analysis on nuclear energy planning, and also as a contribution to applied mathematics. From point of view of nuclear policy analysis, this thesis is not designed to offer realistic detail on nuclear engineering itself, which is of second order relative to our chosen problem. The goal is to address some large scale problems in the management of the world stocks of two important nuclear fuels, Uranium (an economically finite natural resource) and Plutonium (the result at first of policies for Uranium burning, and later of policies on fast reactor breeding). This thesis assumes, as a ‘political’ working hypothesis, that at some future time world governments will agree urgently to decarbonise the world economy. Up to that point, assuming no previous large progress towards decarbonisation, basic world electricity consumption will have continued to grow at its historic average of 1.9% compound. This rate is hypothetically a combination of slower growth in the developed world and faster growth in the developing world. On this hypothesis, a necessary but not sufficient condition for decarbonising the economy would be the complete decarbonisation of future basic electricity demand, plus the provision of sufficient extra decarbonised electricity supply to take over powering all land transport. The demand for electricity for land transport at any time is assumed to equal (in line with historical experience) an increment of approximately 20% above the contemporary basic world demand for electricity. The hypothetical scenario for achieving this model of decarbonisation, without major stress to the worlds economic and social system, is to expand nuclear power to meet the whole of basic electricity demand. This would leave intermittent renewable sources to power the intermittent electricity demands of road transport.This thesis explores the above hypothetical future in various ways. We first list published forecasts of future Uranium use and future Uranium supply. These suggest that presently known Uranium reserves can meet demand for many decades. However on extrapolating the cumulative demand for Uranium that results from the above working hypothesis, we find that if a dash to decarbonise world electricity supply begins immediately, this would consume a very large multiple of presently known Uranium reserves. Sustaining that decarbonisation for only a few more decades of demand growth would consume further large multiples of the known Uranium supply. A delay in the start of the dash for decarbonisation by only a few decades greatly increases the cumulative Uranium demand needed to reach decarbonisation even briefly.Therefore the sustained achievement of decarbonisation, in a world economy of the historical type, requires such large Uranium resources that a successor fuel cycle is required. This thesis models only the case of a Uranium-based fast reactor fuel cycle, since this cycle can in principle consume all the cumulative past and future Plutonium stockpile, and can then meet its own Plutonium needs for a long period (hundreds or thousands of years), allowing ample time for economic adjustment. However a commercially effective fast reactor technology is some decades away.Up to this point, the thesis has only added two physical factors to the existing debate on Uranium needs: namely cumulative growth of electricity demand at its historic rate, and a political choice for 100% physical decarbonisation of the electricity supply.The mathematical and economic contribution of the thesis then begins. We ask the following questions:1. Under what circumstances would profit-maximising investors (or an economically rational centralized economy) actually choose to build enough reactors to decarbonise the world electricity supply?2. Would the need for investors to make a profit increase or decrease the life of the economically accessible Uranium reserves?3. What is the effect of accelerating or delaying the technical availability of fast reactors?4. When if at all would there be shortages of Uranium or Plutonium?5. Under what circumstances would rational investors chose a smooth and physically feasible handover from Uranium burning to fast reactors, thus avoiding the need for a large but temporary return to fossil fuel?The above questions set a mathematically demanding problem: four interacting physical stocks and two physical flow variables ( control variables) must simultaneously be optimized, along with their economic effects. The two control variables are the rate of building or decommissioning Uranium burners, and the rate of building or decommissioning fast reactors. The first control variable drives the cumulative stock of Uranium burning reactors, and hence the resulting maximum physical supply of electricity (with sales income bounded by demand), less the costs of operating, and of new investment. This variable also drives the cumulative depletion of the finite economically extractable reserve of Uranium, and it simultaneously drives an increase in the free Plutonium stock (from Uranium burning). The second control variable, the rate of building or decommissioning fast reactors, drives a decrease in the Plutonium stock (from charging new fast reactors) and it drives a cumulative increase in the stock of fast reactors. This affects the resulting rate of supply of electricity and of income less operating costs and new investment costs. The combined sales of electricity from the two reactor systems is bounded by the total world demand for electricity.The thesis explores this problem in several stages. A fully stochastic form of the problem (stochastic in the price of electricity) is posed using the tools of contingent claims analysis, but this proves intractable to solve, even numerically. Fortunately the price increases needed to impose decarbonisation are very large, and they result from discrete and long lasting government actions. Hence for policy analysis it is adequate to assume a large one off change in electricity price, and observe the progress towards the resulting evolving equilibrium. This problem is also addressed in stages, firstly we optimise the Uranium burning and the fast reactor cycles in isolation from each other, then we allow some purely heuristic and manually controlled interaction between them. Finally we solve, and economically optimize, the total dynamic system of two physical control variables and the resulting four interacting dependent stock variables.

Ce travail se place dans le cadre du traitement-recyclage des combustibles usés pour des réacteurs nucléaires de Génération IV. Ce travail s’est focalisé sur l’étape de mise en forme du combustible en étudiant plus spécifiquement les potentialités du procédé de coulage-gélification. Cette étude a porté dans un premier temps sur un simulant non radioactif de l’oxyde mixte (U,Pu)O2, à savoir l’oxyde de cérium CeO2. Des formulations de suspensions ont été identifiés permettant leur gélification dans des moules non poreux. Ce processus de gélification s’appuie sur des réactions de polymérisation. L’étude du comportement thermique des crus a permis de définir des traitements conduisant à l’élimination complète des auxiliaires organiques et à la densification quasi-complète des crus. L’optimisation du procédé de coulage-gélification a permis d’élaborer des céramiques de CeO2 de forme annulaire répondant au cahier des charges des futurs combustibles.L’ensemble du procédé de coulage-gélification a pu être transposé à une poudre d’oxyde mixte (U,Ce)O2 synthétisée par co-précipitation oxalique. Pour cela, les caractéristiques spécifiques de cette poudre ont été prises en compte et ont nécessité le recours à une étape de supplémentaire de désagglomération. / This work falls within the framework of the reprocessing-recycling of 4th Generation nuclear used fuels. This work was focused on nuclear fuel shaping process by studying more specifically the potential of gelcasting process. This study first examined a non-radioactive surrogate of mixed oxide (U,Pu)O2, namely the cerium oxide CeO2. Suspension formulations have been identified allowing gelation in non-porous molds. This gelling process is based on polymerization reactions. The study of green compacts thermal behavior has allowed to define treatments leading to the complete elimination of every organic auxiliaries and to the green compacts densification. Every step of the gelcasting process was optimized to elaborate annular shape CeO2 ceramics meeting the specification.The whole gelcasting process has been transposed to a mixed oxide (U,Ce)O2 powder synthesized by oxalic co-precipitation. To this end, powder characteristics have been taken in consideration and have required an additional deagglomeration step.

In Pressurised Water Reactors, the reactor pressure vessel (RPV) is considered a life limiting component due to the degradation of its mechanical properties. Nano-scale Cu-enriched precipitates are known to cause embrittlement in the form of increases in hardness and the ductile-to-brittle transition temperature. The effect of irradiation on the RPV is the dominant contributor to this embrittlement. This is due to the increased mobility of Cu from the high number of vacancies, and the matrix damage providing many heterogeneous nucleation sites. However, there are also thermal effects which may be difficult to separate from the irradiation effects. To understand the contribution of the long term thermal ageing to RPV embrittlement a series of weld and plate materials containing systematic variations of Ni and Cu has been thermally aged for times up to 100,000 hrs at 330 degrees C, 365 degrees C and 405 degrees C. Microstructural characterisation using Atom Probe Tomography has been performed. Complimentary Monte-Carlo simulations have been used to investigate the early stages of formation of Cu-enriched precipitates. Thermal ageing produces a high number density of nano-scale Cu-enriched precipitates. These nanometre precipitates have a Ni-Mn-Si rich interface which was found to be wider with increased precipitate size, lower ageing temperature and higher bulk Ni content. This interface reduces the interfacial energy of the Cu-enriched precipitates through a combination of the minimising of unfavourable Fe-Cu bonds and reduction in lattice strain. The matrix Cu levels after ageing for 90,000 - 100,000 hrs were found to be around 0.06 - 0.07 at.%, close to the expected solubility limits for Cu in Fe. The Fe content of the precipitates has been characterised and found to be higher at lower ageing temperature and for smaller precipitate sizes. Cu precipitation and solute segregation at dislocations were observed, particularly in the SG steels aged at higher temperatures where the supersaturation is lowest. Movies were produced from the Kinetic Monte-Carlo modelling (see accompanying DVD) and along with other analyses indicated that sub-critical Ni and Mn clusters may be active in the formation of Cu-enriched precipitates. This mechanism appears to occur in both the high and low Ni steels. Thus, the higher number density of larger precipitates observed with increased bulk Ni content is thought to be a consequence of the greater number of sub-critical Ni-Mn clusters providing more nucleation sites. A small number of grain boundaries were examined. Segregation of many solute species to them was observed, which is thought to result from a multi-element co-segregation process. Ni-Mn-Si precipitates were observed at grain boundaries and dislocations in the high Ni steels with high and low Cu levels. These are consistent with similar phases in found in irradiated high Ni steels. In the high Cu steels these particles were much larger and associated with Cu-enriched precipitates.