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Dans le but de fabriquer des combustibles de type oxyde mixte (MOX = (U,Pu)O2) pour les réacteurs nucléaires du futur, l‘ajout d‘un nouvel intrant, l‘octaoxyde de triuranium (U3O8), est envisagé. Ces travaux de thèse portent sur l'influence de cet ajout pendant le frittage du MOX, ainsi que sur la stabilité dimensionnelle des pastilles frittées lors de recuits. Ces premiers essais ont révélé une dé-densification à haute température des pastilles incorporant une poudre d'U3O8 lorsqu‘elle était issue d‘une synthèse oxalique.Ce phénomène indésirable a été étudié sur un simulant inactif : un oxyde de cérium synthétisé par voie oxalique, afin de développer les techniques expérimentales et les protocoles d‘analyse. Les résultats ont mis en évidence le lien entre la présence d‘impuretés carbonées et le phénomène de dé-densification. De plus, l‘évolution de la dé-densification a été observée pour la première fois, par micro-tomographie X in situ au cours du frittage.L‘étude appliquée au combustible MOX a confirmé le rôle majeur des impuretés carbonées lors du frittage. Les évolutions microstructurales, la quantification des espèces carbonées relâchées pendant le frittage et l‘analyse des gaz piégés dans les pores du matériau fritté ont de plus conduit à identifier un mécanisme de dé-densification. Enfin, une modélisation du comportement thermomécanique du combustible sous l‘effet de la pressurisation des pores, a permis de conforter le mécanisme envisagé. Fort de cette connaissance, un nouveau cycle de frittage a pu alors être proposé et mis en application avec succès pour limiter les effets de ce phénomène. / 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.

Prélevés de manière non invasive, les gaz respiratoires sont constitués de nombreux composés organiques volatils (VOCs) dont la quantité dépend de l’état de santé du sujet. L’analyse quantitative de l’air expiré présente alors un fort intérêt médical, que ce soit pour le diagnostic ou le suivi de traitement. Dans le cadre de ma thèse, nous proposons d’étudier un dispositif d’analyse des gaz respiratoires, et notamment de ces VOCs. Cette thèse multidisciplinaire aborde différents aspects, tels que le choix des capteurs, du matériel et des modes d’acquisition, l’acquisition des données à l’aide d’un banc gaz, et ensuite le traitement des signaux obtenus de manière à quantifier un mélange de gaz. Nous étudions la réponse d’un capteur à oxyde métallique (MOX) à des mélanges de deux gaz (acétone et éthanol) dilués dans de l’air synthétique (oxygène et azote). Ensuite, nous utilisons des méthodes de séparation de sources de manière à distinguer les deux gaz, et déterminer leur concentration. Pour donner des résultats satisfaisants, ces méthodes nécessitent d’utiliser plusieurs capteurs dont on connait la forme mathématique du modèle décrivant l’interaction du mélange avec le capteur, et qui présentent une diversité suffisante dans les mesures d’étalonnage pour estimer les coefficients de ce modèle. Dans cette thèse, nous montrons que les capteurs MOX peuvent être décrits par un modèle de mélange linéaire quadratique, et qu’un mode d’acquisition fonctionnant en double température permet de générer deux capteurs virtuels à partir d’un unique capteur physique. Pour quantifier précisément les composants du mélange à partir des mesures sur ces capteurs (virtuels), nous avons conçu des méthodes de séparation de sources, supervisées et non supervisées appliquées à ce modèle non-linéaire : l’analyse en composantes indépendantes, des méthodes de moindres carrés (algorithme de Levenberg-Marquardt), et une méthode bayésienne ont été étudiées. Les résultats expérimentaux montrent que ces méthodes permettent d’estimer les concentrations de VOCs contenus dans un mélange de gaz, de façon précise, en ne nécessitant que très peu de points de calibration. / Non-invasively taken, exhaled breath contains many volatile organic compounds (VOCs) whose amount depends on the health of the subject. Quantitative analysis of exhaled air is of great medical interest, whether for diagnosis or for a treatment follow-up. As part of my thesis, we propose to study a device to analyze exhaled breath, including these VOCs. This multidisciplinary thesis addresses various aspects, such as the choice of sensors, materials and acquisition modes, the acquisition of data using a gas bench, and then the processing of the signals obtained to quantify a gas mixture. We study the response of a metal oxide sensor (MOX) to mixtures of two gases (acetone and ethanol) diluted in synthetic air (oxygen and nitrogen). Then, we use source separation methods in order to distinguish the two gases, and to determine their concentration. To give satisfactory results, these methods require first to use several sensors for which we know the mathematical model describing the interaction of the mixture with the sensor, and which present a sufficient diversity in the calibration measurements to estimate the model coefficients. In this thesis, we show that MOX sensors can be described by a linear-quadratic mixing model, and that a dual temperature acquisition mode can generate two virtual sensors from a single physical sensor. To quantify the components of the mixture from measurements on these (virtual) sensors, we have develop supervised and unsupervised source separation methods, applied to this nonlinear model: independent component analysis, least squares methods (Levenberg Marquardt algorithm), and a Bayesian method were studied. The experimental results show that these methods make it possible to estimate the VOC concentrations of a gas mixture, accurately, while requiring only a few calibration points.

Afin d’évaluer les performances du combustible irradié en situation de stockage géologique, des recherches sont menées sur le comportement à long terme des combustibles irradiés (UOx et MOx) en conditions environnementales se rapprochant de celles du site de stockage français. L’objectif de cette thèse est de déterminer si la géochimie de la couche géologique d'argilites du Callovo-Oxfordien (COx) et la corrosion des conteneurs en acier (produisant du fer et de l'hydrogène) ont un impact sur la dissolution oxydante de la matrice (U,Pu)O2 sous radiolyse alpha de l’eau.Des expériences de lixiviation ont été réalisées avec des pastilles de UO2 dopées en émetteurs alpha (Pu) et du combustible MOx MIMAS (non irradié ou irradié en réacteur) afin de mettre en évidence l’influence de l’eau du COx et de la présence de fer métallique sur la dissolution oxydante de ces différents matériaux induite par la radiolyse de l’eau. Les résultats indiquent un effet inhibiteur de l’eau du COx sur la dissolution oxydante de la matrice UO2. D’autre part en présence de fer, deux régimes différents sont observés. Sous irradiation alpha dominante telle que celle attendue en stockage géologique, la dissolution oxydante de la matrice UO2 et du combustible MOx est très fortement inhibée du fait de la consommation des espèces radiolytiques oxydantes par le fer en solution avec précipitation d’hydroxydes de Fe(III) à la surface des pastilles. En revanche, sous forte irradiation beta/gamma comme dans le cas du combustible irradié, les traceurs de l’altération indiquent que celle-ci se poursuit en présence de fer tandis que la concentration en uranium en solution est contrôlée par la solubilité de UO2(am,hyd). Ceci est expliqué par le déplacement du front redox de la surface du combustible vers la solution homogène ne protégeant plus le combustible. Les modèles géochimique (code CHESS) et de transport réactif (code HYTEC) développés représentent correctement les principaux résultats et mécanismes mis en jeu. / 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.

La prévision du comportement mécanique macroscopique de matériaux hétérogènes à partir des propriétés de leurs constituants est possible pour diverses classes de comportement (élastique, viscoélastique, etc) et ce, grâce à la théorie de l'homogénéisation. Néanmoins l'extension de cette théorie pour des matériaux possédant un comportement viscoélastique non linéaire (ou élasto-viscoplastique) reste une question ouverte à laquelle nous nous attaquons dans ce travail afin de prédire le comportement macroscopique des combustibles oxydes mixtes uranium-plutonium (MOX) utilisés dans les réacteurs nucléaires à eau sous pression (REP) français. Dans cette optique des solutions analytiques et purement nunériques ont été obtenues et le modèle retenu est utilisé pour simuler le comportement des combustibles / 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.

<p>Nuclear energy is currently one of the world’s main sources of electricity. Closely connected to the use of nuclear energy are important issues such as the nonproliferation of fissile material that may potentially used in nuclear weapons (safeguards), and the management of the highly radioactive nuclear waste. This thesis addresses both these issues by contributing to the development of new experimental methods for ensuring safe and secure handling of the waste, with focus on methods to be used prior to encapsulation and final storage.</p><p>The methods rely on high resolution gamma ray spectroscopy (HRGS), involving the measurement and analysis of emitted gamma radiation from the fission products ¹³⁷Cs, ¹³⁴Cs and ¹⁵⁴Eu. This technique is nondestructive, making it relatively nonintrusive with respect to the normal operation of the nuclear facilities.</p><p>For the safeguards issue, it is important to experimentally verify the presence and identity of nuclear fuel assemblies and also that the fuel has experienced normal, civilian reactor operation. It has been shown in this thesis that the HRGS method may be used for verifying operator declared fuel parameters such as burnup, cooling time and irradiation history. In the experimental part of the work, the burnup and the cooling time has been determined with an accuracy of 1.6% and 1.5%, respectively (1 σ).</p><p>A technique has also been demonstrated, utilizing the ratio ¹³⁴Cs/¹⁵⁴Eu, with which it is possible to determine whether a fuel assembly is of MOX or LEU type. This is of interest for safeguards as well as for the safe operation of a final storage facility.</p><p>As an improvement to the HRGS technique, measuring a part of the fuel assembly length in order to reduce measurement time has been suggested and investigated. A theoretical case for partial defect verification has also been studied as an extension of the HRGS technique. </p><p>Finally, HRGS has been used for determining the decay heat in spent nuclear fuel assemblies, which is of importance for the safe operation of a final storage facility. This application is based on the radiation from ¹³⁷Cs, and the accuracy demonstrated was within 3% (1 σ).</p>

Nuclear energy is currently one of the world’s main sources of electricity. Closely connected to the use of nuclear energy are important issues such as the nonproliferation of fissile material that may potentially used in nuclear weapons (safeguards), and the management of the highly radioactive nuclear waste. This thesis addresses both these issues by contributing to the development of new experimental methods for ensuring safe and secure handling of the waste, with focus on methods to be used prior to encapsulation and final storage. The methods rely on high resolution gamma ray spectroscopy (HRGS), involving the measurement and analysis of emitted gamma radiation from the fission products 137Cs, 134Cs and 154Eu. This technique is nondestructive, making it relatively nonintrusive with respect to the normal operation of the nuclear facilities. For the safeguards issue, it is important to experimentally verify the presence and identity of nuclear fuel assemblies and also that the fuel has experienced normal, civilian reactor operation. It has been shown in this thesis that the HRGS method may be used for verifying operator declared fuel parameters such as burnup, cooling time and irradiation history. In the experimental part of the work, the burnup and the cooling time has been determined with an accuracy of 1.6% and 1.5%, respectively (1 σ). A technique has also been demonstrated, utilizing the ratio 134Cs/154Eu, with which it is possible to determine whether a fuel assembly is of MOX or LEU type. This is of interest for safeguards as well as for the safe operation of a final storage facility. As an improvement to the HRGS technique, measuring a part of the fuel assembly length in order to reduce measurement time has been suggested and investigated. A theoretical case for partial defect verification has also been studied as an extension of the HRGS technique. Finally, HRGS has been used for determining the decay heat in spent nuclear fuel assemblies, which is of importance for the safe operation of a final storage facility. This application is based on the radiation from 137Cs, and the accuracy demonstrated was within 3% (1 σ).

Es wurde eine SP3-Transportmethode entwickelt, die neutronenkinetische Rechnungen für die Kerne von Leichtwasserreaktoren mit höherer Genauigkeit als die gegenwärtig in der Kernauslegung angewandten Standardmethoden auf Basis der Zweigruppendiffusionsnäherung er-laubt. Eine Verbesserung der Genauigkeit von Abbrandrechnungen und der Berechnung von Tran-sienten ist für heterogene Kerne notwendig, in denen neben UO2-Brennelementen auch Mischoxyd – Brennelemente eingesetzt werden. In einem ersten Schritt wird die in dem Rechenprogramm DYN3D verwendete Zweigruppendiffusi-onsmethode auf viele Energiegruppen erweitert. Auf der Basis von Untersuchungen zu einer optima-len Gruppenstruktur wird die Verwendung von 8-10 Energiegruppen der Neutronen als optimal erach-tet. Das Verfahren wurde anhand von stationären und transienten Rechnungen für das OECD/NEA und US NRC PWR MOX/UO2 Core Transient Benchmark verifiziert. In den nächsten Schritten erfolgte die Entwicklung und Implementierung einer SP3-Näherung in DYN3D. Dabei besteht die Möglichkeit, ein feineres Gitter im BE zu benutzen. Das Verfahren wurde zunächst durch pinweise Berechnung stationärer Zustände des obigen Benchmarks verifiziert. Untersuchungen für das Benchmarkproblem zeigen, dass das Verhältniss des 2-ten Momentes zum 0-ten Moment des Flusses klein ist. Die beiden SP3-Gleichungen können deshalb separat in iterativer Weise gelöst werden. Dies reduziert den benötigten Speicherplatz und erfordert weniger CPU-Zeit. Dieses vereinfachte Verfahren wurde deshalb ebenfalls in das Programm implementiert. Es wird ge-zeigt, dass mit diesem Verfahren eine vergleichbare Genauigkeit erreicht wird. Stabweise Rechnun-gen mit 4, 8 und 16 Energiegrupppen wurden für einen stationären Zustand des Benchmarks durch-geführt. Eine 3-dimensionale Aufgabe des Benchmarks mit Rückkopplung und Vollleistung wurde mit dem optimierten SP3-Verfahren gerechnet. A SP3 transport approximation was developed for neutron kinetic calculations of cores of light water reactors with a higher accuracy than the present standard methods of core design based on the two group diffusion approximation. An improvement of accuracy for burnup and transient calculations is required for cores loaded with UO2 and MOX fuel assemblies. In the first step, the two group diffusion method applied in the computer code DYN3D was extended to an arbitrary number of groups. Investigations for an optimal group structure have shown that a number of 8 to 10 energy groups of neutrons seems to be reasonable. The multi-group technique was verified for steady states and transients of the OECD/NEA und US NRC PWR MOX/UO2 Core Tran-sient Benchmark. In the next steps, a SP3-approximation was developed and implemented into DYN3D. The possibility of using finer meshes inside the fuel assemblies is involved in this method. The technique was veri-fied by pinwise calculations for steady states of the above mentioned benchmark. The investigations to the benchmark problem have shown that ratio of the 2nd moment of flux to the 0th moment is small. Therefore the two coupled SP3 equations can be solved separately in an iterative way. The required computer memory and the CPU time can be reduced by this technique. This sim-pler method was also implemented in the code. It is shown that the reached accuracy is comparable to accuracy of the original technique. Pinwise calculations with 4, 8 and 16 energy groups were per-formed for a steady state of this benchmark. A three-dimensional problem of the benchmark at full power and with feedback was calculated with the optimized SP3 technique. The optimized method was used for the time integration of the transient SP3 equations. The pinwise calculation of a control rod ejection was tested for a simple system and the results were compared with the diffusion solution.

Orientador: Prof. Dr. José Rubens Maiorino / Tese ( doutorado)- Universidade Federal do ABC. Programa de Pós-Graduação em Energia, 2017. / O presente trabalho apresenta a análise e estudo da viabilidade de converter um reator de água pressurizada, para que opere com combustível a base de mistura de óxidos de tório e urânio (Th, U)O2, em substituição ao tradicional dióxido de urânio, com a finalidade de redução de actinídeos de longa vida, em especial plutônio, e de gerar um estoque de 233U que poderia vir a no futuro ser utilizado em ciclos de combustível avançados, em um processo mais sustentável e para aproveitar a grande reserva de tório disponível no planeta e em especial no Brasil. O reator escolhido como referência foi o AP1000, que é considerado como um dos reatores mais seguros e modernos da atual Geração III, e por sua similaridade com os reatores já consolidados e utilizados no Brasil para geração de energia elétrica. Os resultados obtidos mostram a viabilidade e potencialidade do conceito, sem a necessidade de mudanças no núcleo do AP1000, e também com vantagens relativamente a este. Os cálculos nêutronicos foram feitos pelo programa SERPENT. Os resultados forneceram uma densidade de potência linear máxima menores que o AP1000, favorecendo a segurança. Além disso a fração de nêutrons atrasados, os coeficientes de reatividade mostraram-se adequados para garantir a segurança do conceito. Os resultados mostraram que é possível uma produção de cerca de 260 Kg de 233U por ciclo, com uma produção mínima de plutônio físsil que favorece a utilização do conceito em ciclos de U-Th, no entanto os estudos apontam que sua vantagem é limitada a ciclos de combustível fechados. / This work presents the neutronics and thermal hydraulics feasibility to convert the UO2 core of the Westinghouse AP1000 in a (U-Th)O2 core, rather than the traditional uranium dioxide, for the purpose of reducing long-lived actinides, especially plutonium, and generates a stock pile of 233U, which could in the future be used in advanced fuel cycles, in a more sustainable process and taking advantage of the large stock of thorium available on the planet and especially in Brazil. The reactor chosen as reference was the AP1000, which is considered to be one of the most reliable and modern reactor of the current Generation III, and its similarity to the reactors already consolidated and used in Brazil for electric power generation. The results show the feasibility and potentiality of the concept, without needs any change in the core of the AP1000, and even with advantages over this. The neutron calculations were made by the SERPENT code. The results provided a maximum linear power density lower than the AP1000, favoring safety. In addition the delayed neutron fraction, the reactivity coefficients proved to be adequate to ensure the safety of the concept. The results show that a production of about 260 Kg of 233U per cycle is possible, with a minimum production of fissile plutonium that favors the use of the concept in U-Th cycles, however studies show that its advantage is limited to closed cycles.