Submitted by Almir Azevedo (barbio1313@gmail.com) on 2014-07-21T13:54:14Z No. of bitstreams: 0 / Made available in DSpace on 2014-07-21T13:54:14Z (GMT). No. of bitstreams: 0 Previous issue date: 2014 / Esta dissertação apresenta uma contribuição para o estudo do processo de envelhecimento de plantas com Reatores a Água Pressurizada (PWR). A análise é feita através da aplicação do Método de Árvore de Falhas, Método de Monte Carlo e Medidas de Importância. A abordagem do estudo de envelhecimento em usinas nucleares, além de dar atenção aos fatores econômicos envolvidos diretamente com a extensão de sua vida operacional, também fornece dados importantes sobre questões de segurança. O mais recente caso envolvendo o processo de extensão da vida de um PWR pode ser vista na Usina de Angra 1 através do investimento de vinte e sete milhões de dólares ($27 millions) para a instalação de uma nova tampa do reator. A ação corretiva geral uma estimativa de extensão de vida útil de Angra 1 em vinte anos, oferecendo grande economia em comparação com o custo de construção de uma nova planta e o descomissionamento da anterior, caso ela tivesse atingido o tempo limite de funcionamento de quarenta anos. A extensão de vida útil de uma planta de energia nuclear deve ser acompanhada por uma atenção especial aos componentes dos sistemas e seus processos de envelhecimento. Após a aplicação da metodologia (análise de envelhecimento do Sistema de Injeção de Borrifo da Contenção) proposta neste trabalho, é possível ver que o aumento na probabilidade de falha de componentes, devido ao processo de envelhecimento, gera o aumento da indisponibilidade geral do sistema que contém esses componentes básicos. os resultados finais obtidos foram como o esperado e pode contribuir para a política de manutenção, evitando processos de envelhecimento prematuros em sistemas de plantas nucleares / This dissertation presents a contribution to the study of aging process of commercial plants with Pressurized Water Reactors (PWRs). The analysis is made through application of the Fault Trees Method, Monte Carlo Method and Importance Measures. The approach of the study of aging in nuclear power plants, besides giving attention to the economic factors involved directly with the extent of their operational life, also provide significant data on security issues. The latest case involving process of life extension of a PWR could be seen in Angra 1 Nuclear Power Plant through investing of twenty and seven millions of dallars ($27 millions) for the installation of a new reactor lid. The corrective action has generated an estimated operating life extension of Angra 1 in twenty years, offering great economy compared with building cost of a new plant and anterior decommissioning, case it had reached the time operating limit of forty years. The extension of the operating life of a nuclear power plant must be accompanied by a special attention to the components of the systems and their aging process. After the application of the methodology (aging analysis of the Containment Spray Injection System) proposed this work, it can be seen that the increase in the rate of components failure, due the aging process, generates the increase in the general unavailability of the system that containing these basic components. The final results obtained were as expected and may contribute to the maintenance policy, preventing premature aging processes in nuclear plants systems

The stress corrosion cracking (SCC) of components which are fabricated from austenitic stainless steel has been observed in the primary circuit of pressurised water reactors (PWR). In recent years it has become an increasing concern that cold work can induce susceptibility to SCC in these materials, even when exposed to good-quality flowing coolant. Laboratory studies which were launched in response to this observation have confirmed that SCC susceptibility is enhanced by cold work. The intention of this study is therefore to investigate the link between the effects of cold work on the material and the susceptibility to SCC. The investigation has been conducted on a grade 304 austenitic stainless steel. Characterisation of the microstructure and mechanical properties has been carried out in the annealed condition, and following cold rolling to a reduction in thickness of 20 %. The cold rolled material has then been subjected to SCC tests in simulated PWR primary circuit coolant. Two types of test were utilised: slow strain rate tests (SSRTs) were carried out in order to investigate the initiation of cracks from a smooth surface and constant load tests using pre-cracked specimens were used to investigate the crack propagation behaviour. In both types of test the SCC produced was predominantly intergranular. The SSRTs revealed that the most susceptible grain boundaries separated grains which had dissimilar deformation microstructures (one grain deformed heavily by planar bands, the other more homogenously). It was also observed that initiation could occur on a grain boundary which is adjacent to an annealing twin. In both microstructural configurations the susceptibility is likely to be due to the deformation incompatibility across the failed boundary, possible indicating that shear at the boundary is important for the initiation of cracking. The crack propagation behaviour of the rolled material was particularly anisotropic; regardless of the loading direction (specimens were manufactured to allow loading along the rolling, transverse and normal plate directions) cracking was observed to occur parallel to the rolling-transverse plane. The origin of this behaviour was explored in terms of preferential alignment of the deformation microstructure and the anisotropic mechanical properties of the rolled plate. Limited transgranular cracking was also observed, which occurred along oxidised deformation bands. The results overall indicate that heterogeneous deformation between different regions of the material, and preferential alignment of the deformation microstructure are important with respect to the SCC susceptibility of the rolled material.

The most common design among U.S. nuclear power plants is the pressurized water reactor (PWR). The three primary design disciplines of these plants are system analysis (which includes thermal hydraulics), neutronics, and fuel performance. The nuclear industry has developed a variety of codes over the course of forty years, each with an emphasis within a specific discipline. Perhaps the greatest difficulty in mathematically modeling a nuclear reactor, is choosing which specific phenomena need to be modeled, and to what detail. A multiphysics computational environment provides a means of advancing simulations of nuclear plants. Put simply, users are able to combine various physical models which have commonly been treated as separate in the past. The focus of this work is a specific multiphysics environment currently under development at Idaho National Labs known as the LOCA Toolkit for US light water reactors (LOTUS). The ability of LOTUS to use uncertainty quantification (UQ) and sensitivity analysis (SA) tools within a multihphysics environment allow for a number of unique analyses which to the best of our knowledge, have yet to be performed. These include the first known integration of the neutronics and thermal hydraulic code VERA-CS currently under development by CASL, with the well-established fuel performance code FRAPCON by PNWL. The integration was used to model a fuel depletion case. The outputs of interest for this integration were the minimum departure from nucleate boiling ratio (MDNBR) (a thermal hydraulic parameter indicating how close a heat flux is to causing a dangerous form of boiling in which an insulating layer of coolant vapour is formed), the maximum fuel centerline temperature (MFCT) of the uranium rod, and the gap conductance at peak power (GCPP). GCPP refers to the thermal conductance of the gas filled gap between fuel and cladding at the axial location with the highest local power generation. UQ and SA were performed on MDNBR, MFCT, and GCPP at a variety of times throughout the fuel depletion. Results showed the MDNBR to behave linearly and consistently throughout the depletion, with the most impactful input uncertainties being coolant outlet pressure and inlet temperature as well as core power. MFCT also behaves linearly, but with a shift in SA measures. Initially MFCT is sensitive to fuel thermal conductivity and gap dimensions. However, later in the fuel cycle, nearly all uncertainty stems from fuel thermal conductivity, with minor contributions coming from core power and initial fuel density. GCPP uncertainty exhibits nonlinear, time-dependent behaviour which requires higher order SA measures to properly analyze. GCPP begins with a dependence on gap dimensions, but in later states, shifts to a dependence on the biases of a variety of specific calculation such as fuel swelling and cladding creep and oxidation. LOTUS was also used to perform the first higher order SA of an integration of VERA-CS and the BISON fuel performance code currently under development at INL. The same problem and outputs were studied as the VERA-CS and FRAPCON integration. Results for MDNBR and MFCT were relatively consistent. GCPP results contained notable differences, specifically a large dependence on fuel and clad surface roughness in later states. However, this difference is due to the surface roughness not being perturbed in the first integration. SA of later states also showed an increased sensitivity to fission gas release coefficients. Lastly a Loss of Coolant Accident was investigated with an integration of FRAPCON with the INL neutronics code PHISICS and system analysis code RELAP5-3D. The outputs of interest were ratios of the peak cladding temperatures (highest temperature encountered by cladding during LOCA) and equivalent cladding reacted (the percentage of cladding oxidized) to their cladding hydrogen content-based limits. This work contains the first known UQ of these ratios within the aforementioned integration. Results showed the PCT ratio to be relatively well behaved. The ECR ratio behaves as a threshold variable, which is to say it abruptly shifts to radically higher values under specific conditions. This threshold behaviour establishes the importance of performing UQ so as to see the full spectrum of possible values for an output of interest. The SA capabilities of LOTUS provide a path forward for developers to increase code fidelity for specific outputs. Performing UQ within a multiphysics environment may provide improved estimates of safety metrics in nuclear reactors. These improved estimates may allow plants to operate at higher power, thereby increasing profits. Lastly, LOTUS will be of particular use in the development of newly proposed nuclear fuel designs.

Coordenação de Aperfeiçoamento de Pessoal de Nível Superior / As ligas de Níquel são amplamente utilizadas na construção de diversos componentes de reatores à água pressurizada PWR. Estes materiais foram selecionados por possuírem elevada resistência mecânica e a corrosão generalizada e compatibilidade com os materiais do reator. No entanto, alguns reatores têm apresentado trincamento causado por corrosão sob tensão em soldas dissimilares efetuadas com a liga de Inconel 182, isto têm ocorrido principalmente no circuito primário. O trabalho aqui apresentado é um estudo para calcular a taxa de propagação de trincas causada por corrosão sob tensão em soldas dissimilares, efetuadas com a liga de Inconel 182 com sobrecamada de solda de Inconel 52 em ambiente simulado do circuito primário de reatores PWR. Este estudo é fundamental para determinar o tempo de vida em serviço dos equipamentos e estabelecer intervalos de inspeção, evitando-se assim acidentes catastróficos. / Nickel alloys are widely used in the construction of several components of the pressurized water reactors (PWR).These materials were selected due to their high mechanical and corrosion resistance as well as its compatibility to the reactor materials. However, some reactors have shown cracking caused by stress corrosion cracking (SCC) in dissimilar metals weld made with alloy Inconel 182, mainly in the primary circuit. The purpose of the present study was to calculate the crack growth rate caused by primary water SCC in dissimilar metals weld made with alloy Inconel 182 and Weld Overlay of Inconel 52.This study is essential to determine the lifetime of the equipments and establish inspection intervals, avoiding catastrophic accidents.

O fenômeno de injeção de vapor em tanques de alívio é de relevância no projeto de reatores de água leve, sejam eles do tipo reator de água pressurizada (PWR) ou reator de água fervente (BWR). Este fenômeno permite a rápida absorção do vapor injetado em massa de água, por meio de sua condensação, uma vez que este vapor pode conter contaminantes químicos ou radiológicos que não permitem o seu descarte diretamente no ambiente. Desta forma, facilita-se a coleta do vapor produzido por descarga de vapor da água do resfriamento do reator, radiologicamente contaminada, e evita-se o que projeto de dispositivos e equipamentos necessite considerar a elevada pressão do vapor. A rapidez com que se dá a condensação é fruto de processos físicos que ocorrem na interface de vapor e água e que ainda não possuem modelo analítico e numérico definido. Em 1972 um modelo semi-empírico foi proposto, o qual, desde então, vem evoluindo. Não obstante, até o presente momento, não há modelo definitivo que se proponha a abranger toda extensão das condições experimentais. Estes modelos são fortemente dependentes do fluxo de massa que atravessa a interface de vapor e água, entretanto, até a presente data, não há expressão que determine este fluxo de massa, de tal forma que o valor de 275 Kg/m2/s vem sendo assumido como \"representativo da ordem de grandeza do fenômeno\" até o presente momento. Neste trabalho, é proposto um método de cálculo analítico do fluxo de massa, considerando-se como premissa a isentropia da injeção, e o desenvolvimento da 1ª e 2ª leis da Termodinâmica. Ainda, o fenômeno é analisado experimentalmente, por meio da análise dos dados produzidos no experimento do Circuito Termo Hidráulico de 150 bar (Loop 150), realizado nas dependências do CENTRO TECNOLÓGICO DA MARINHA EM SÃO PAULO. Por fim, um modelo numérico em software comercial foi desenvolvido para complementar a análise. Os resultados obtidos comprovam que a formulação isentrópica do fluxo de massa corrige de maneira satisfatória o fluxo de massa constante utilizado até então nos modelos semi-empíricos. Tal comprovação se deu através de análise numérica e da confrontação com dados experimentais obtidos na literatura. / The phenomenon of vapor injection in relief tanks presents relevance in the design of light water reactors, be they of the type pressurized water reactor (PWR) or boiling water reactor (BWR). This phenomenon allows the rapid absorption of the vapor injected in a mass of water, by condensation. Since this vapor may contain chemical or radiological contaminants that do not allow its discharge directly in the environment, it must be collected. The condensation avoids the design of devices and equipment, which need to consider the high vapor pressure, and allows the vapor to be collected. The rapidity with which the condensation occurs is the result of physical processes that occur at the interface of steam and water. These processes do not yet have a defined numerical and analytical model. In 1972 a semi-empirical model was proposed, which has, since then, evolved. Nevertheless, up to the present moment, there is no definitive model that intends to cover every extension of the experimental conditions. These models are strongly dependent on the mass flow through the steam and water interface, however, up to date, there is no expression that determines this mass flow. For the sake of this, the value of 275 kg / m2 / s has been assumed as \"representative of the order of magnitude of the phenomenon\" up to the present moment. In this work, a method of analytical calculation of mass flow is proposed, considering as premise the isotropy of the injection, and the development of the 1st and 2nd laws of thermodynamics. Still, the phenomenon is analyzed experimentally, by means of the analysis of the data produced in the experiment of Loop 150, realized in dependencies of the CENTRO TECNOLÓGICO DA MARINHA EM SÃO PAULO. Finally, a numerical model in commercial software was developed to complement the analysis. The result is proven with an isentropic mass flow formulation, which satisfactorily corrected the mass used in the former semi-empirical models. Such verification was performed through a series of data and confrontation with experimental data in the literature.

Nenhuma / No circuito primário de uma usina nuclear do tipo PWR (Pressurized Water Reactor), o refrigerante do reator é mantido a uma temperatura interna por volta de 300 C e pressão interna da ordem de 15,0 MPa, durante operação normal. O Vaso de Pressão do Reator (VPR) contém os elementos combustíveis e é considerado o componente mais importante do circuito primário. A integridade do VPR deve ser assegurada durante toda a vida útil da usina, de forma a proteger os trabalhadores da usina e o público em geral dos danos decorrentes da liberação de material radioativo.Uma das condições de carregamento mais severas que pode ameçar a integridade do VPR é causada por um transitório conhecido como Choque Térmico Pressurizado (PTS - Pressurized Thermal Shock). O VPR estará sujeito a tal condição durante um acidente com perda de refrigerante do núcleo do reator. Em um evento como este, o sistema de refrigeração de emergência do núcleo é ativado, o que provoca a injeção de água fria no interior do VPR e, consequentemente, um súbito resfriamento da parede do vaso. As tensões térmicas, resultantes deste choque térmico, associadas às tensões causadas pela repressurização do sistema, resultam em tensões de tração bastante elevadas, atingindo um valor máximo na superfície interna da parede do vaso. Além disso, a baixa temperatura provoca uma redução na tenacidade à fratura do material. Tal cenário pode levar à propagação de trincas relativamente pequenas através da parede do vaso. Portanto, ferramentas para prever o comportamento de trincas durante um evento de PTS são importantes e necessárias. O tema do presente trabalho se insere neste contexto. Em primeiro lugar, foi feito um estudo das principais questões envolvidas com o problema de PTS em vasos de pressão de reatores PWR. Essas questões dizem respeito ao comportamento à fratura de aços ferríticos na região de transição frágil-dúctil, aos procedimentos de análise de PTS disponíveis em documentos normativos e ao uso de ferramentas de análise numérica para cálculo de distribuição de temperaturas e tensões, e para obtenção de parâmetro de mecânica da fratura representativo da força motriz da trinca. Como principal objetivo do trabalho, foram desenvolvidos modelos de elementos finitos para avaliação do comportamento estrutural de um protótipo de VPR, contendo trincas em sua superfície, utilizado em um experimento de PTS. Procedimentos de mecânica da fratura foram também aplicados para prever eventuais crescimentos de trinca através da espessura da parede do vaso. Resultados das análises numéricas foram comparados com aqueles obtidos com o uso de método simplificado e com medições realizadas no experimento de PTS. / In the primary system of a pressurized water reactor (PWR) nuclear power plant, the reactor coolant is kept at internal temperature around 300 C and internal pressure in the order of 15,0 MPa, during normal operation. The reactor pressure vessel (RPV) contains the fuel assemblies and is considered the most important component of the reactor primary system. The RPV integrity must be assured all along its useful life to protect the general public against radiation liberation damage. One of the most severe load conditions that may threaten the integrity of a RPV is caused by a transient known as pressurized thermal shock (PTS). The RPV may be subjected to such a condition during a loss of coolant accident. In an event like that, the emergency core cooling system is activated, what leads to a sudden cooling of the RPV wall. The thermal stresses due to this thermal shock on the vessel wall, in combination with the pressure stresses from repressurization of the system, results in large tensile stresses, which are maximum at the inside surface of the vessel. In addition, the low temperature causes a decrease in the material fracture toughness. Such a scenario may lead to the propagation of relatively small cracks through the vessel wall. Therefore, analysis tools to predict crack growth behavior during a PTS event are important and necessary. The theme of the present work is connected with this research area. In the first place, the critical issues involved with the PTS problem were reviewed. These issues are related to the fracture behavior of ferritic steels in the ductile-to-brittle transition region, the PTS analysis procedures available in industry codes and standards, and the use of numerical analysis tools for calculation of temperature and stress distribution and for computation of crack driving force parameter. As the main goal, finite element models were developed for the assessment of the structural behavior of a RPV prototype, containing surface cracks, used in a PTS experiment. Fracture mechanics procedures were applied to predict crack growth through the vessel wall. The results of numerical analyses were compared with those obtained with the use of a simplified methodology and measurements from the PTS experiment.

O fenômeno de injeção de vapor em tanques de alívio é de relevância no projeto de reatores de água leve, sejam eles do tipo reator de água pressurizada (PWR) ou reator de água fervente (BWR). Este fenômeno permite a rápida absorção do vapor injetado em massa de água, por meio de sua condensação, uma vez que este vapor pode conter contaminantes químicos ou radiológicos que não permitem o seu descarte diretamente no ambiente. Desta forma, facilita-se a coleta do vapor produzido por descarga de vapor da água do resfriamento do reator, radiologicamente contaminada, e evita-se o que projeto de dispositivos e equipamentos necessite considerar a elevada pressão do vapor. A rapidez com que se dá a condensação é fruto de processos físicos que ocorrem na interface de vapor e água e que ainda não possuem modelo analítico e numérico definido. Em 1972 um modelo semi-empírico foi proposto, o qual, desde então, vem evoluindo. Não obstante, até o presente momento, não há modelo definitivo que se proponha a abranger toda extensão das condições experimentais. Estes modelos são fortemente dependentes do fluxo de massa que atravessa a interface de vapor e água, entretanto, até a presente data, não há expressão que determine este fluxo de massa, de tal forma que o valor de 275 Kg/m2/s vem sendo assumido como \"representativo da ordem de grandeza do fenômeno\" até o presente momento. Neste trabalho, é proposto um método de cálculo analítico do fluxo de massa, considerando-se como premissa a isentropia da injeção, e o desenvolvimento da 1ª e 2ª leis da Termodinâmica. Ainda, o fenômeno é analisado experimentalmente, por meio da análise dos dados produzidos no experimento do Circuito Termo Hidráulico de 150 bar (Loop 150), realizado nas dependências do CENTRO TECNOLÓGICO DA MARINHA EM SÃO PAULO. Por fim, um modelo numérico em software comercial foi desenvolvido para complementar a análise. Os resultados obtidos comprovam que a formulação isentrópica do fluxo de massa corrige de maneira satisfatória o fluxo de massa constante utilizado até então nos modelos semi-empíricos. Tal comprovação se deu através de análise numérica e da confrontação com dados experimentais obtidos na literatura. / The phenomenon of vapor injection in relief tanks presents relevance in the design of light water reactors, be they of the type pressurized water reactor (PWR) or boiling water reactor (BWR). This phenomenon allows the rapid absorption of the vapor injected in a mass of water, by condensation. Since this vapor may contain chemical or radiological contaminants that do not allow its discharge directly in the environment, it must be collected. The condensation avoids the design of devices and equipment, which need to consider the high vapor pressure, and allows the vapor to be collected. The rapidity with which the condensation occurs is the result of physical processes that occur at the interface of steam and water. These processes do not yet have a defined numerical and analytical model. In 1972 a semi-empirical model was proposed, which has, since then, evolved. Nevertheless, up to the present moment, there is no definitive model that intends to cover every extension of the experimental conditions. These models are strongly dependent on the mass flow through the steam and water interface, however, up to date, there is no expression that determines this mass flow. For the sake of this, the value of 275 kg / m2 / s has been assumed as \"representative of the order of magnitude of the phenomenon\" up to the present moment. In this work, a method of analytical calculation of mass flow is proposed, considering as premise the isotropy of the injection, and the development of the 1st and 2nd laws of thermodynamics. Still, the phenomenon is analyzed experimentally, by means of the analysis of the data produced in the experiment of Loop 150, realized in dependencies of the CENTRO TECNOLÓGICO DA MARINHA EM SÃO PAULO. Finally, a numerical model in commercial software was developed to complement the analysis. The result is proven with an isentropic mass flow formulation, which satisfactorily corrected the mass used in the former semi-empirical models. Such verification was performed through a series of data and confrontation with experimental data in the literature.

Ce mémoire rapporte les études de protéines amyloïdes en interaction avec des membranes modèle afin d’établir une relation structure toxicité. Nous avons choisi différents modèles membranaires (monocouches, bicouches) de composition lipidique et charges différentes et utilisé différentes méthodes physico-chimiques afin de caractériser les interactions des protéines amyloïdes avec les membranes.Nous avons montré l’importance de la contribution électrostatique dans les interactions de la protéine amyloïde HET-s (218-289) et ses mutants avec les membranes modèles.L’ellipsométrie a démontré que les mutants toxiques de HET-s (218-289) (M8, WT.Y1Y2V2) perturbentfortement les monocouches lipidiques à l’interface air-eau. La structure riche en feuillets β antiparallèles des protéines àl’interface air-eau et dans l’interaction avec les monocouches de lipides a été démontrée par la spectroscopie PMIRRAS (Polarization Modulation – Infrared Reflection Absorption Spectroscopy). Nous avons établie que l’interface air-eau peut modifier l’agrégation des protéines amyloïdes. A l’aide de la spectroscopie de fluorescence, la spectroscopie PWR (Plasmon-Waveguided Resonance) et la spectroscopie ATR-FTIR (Attenuated Total Reflection – Fourier Transform Infrared), nous avons mis en évidence que la protéine toxique M8 adopte une structure riche en feuillets β antiparallèles en altérant fortement l’intégrité des bicouches lipidiques. Au contraire, la protéine non toxique WT se structure en feuillets β parallèles dans ces interactions et elle ne perturbe pas l’homogénéité des membranes. La toxicité de la protéine M8 semble liée à son organisation différente et à sa capacité à réorganiser les membranes.Nos résultats confortent également l’hypothèse de la toxicité des oligomères amyloïdes.Une étude sur la fabrication d’une cellule microfluidique pour la séparation de différents types d’autoassemblage afin de les détecter et de les étudier en interaction avec des liposomes par spectroscopie infrarouge est présentée. Une cellule microfluidique de CaF2 de 8 μm d’épaisseur de canaux est obtenue et est utilisée pour la détection d’une protéine de test. / This manuscript reports the studies of amyloid proteins in interaction with membrane models in order to establish their structure-toxicity relationship.Membrane models (monolayer, bilayer) of different charge and lipid composition were used. We used various physico chemical methods to characterize the interaction of these amyloid proteins with membranes.We showed the importance of the electrostatic contribution in the interactions of the amyloid protein HET-s(218-289) and its mutants with model membranes.Ellipsometry showed that the toxic mutants of HET-s (218-289) (M8, WT.Y1Y2V2) strongly disturbed thelipid monolayers at the air-water interface. The structure rich in antiparallel β sheets of auto-assembled proteins at theair-water interface and in interaction with lipid monolayers at the air-water interface has been demonstrated by the PMIRRAS spectroscopy (Polarization Modulation - Infrared Reflection Absorption Spectroscopy). We established that theair-water interface can change the aggregation properties of amyloid proteins.By using fluorescence spectroscopy, PWR spectroscopy (Plasmon Resonance-Waveguided spectroscopy) and ATR-FTIR spectroscopy (Attenuated Total Reflection - Fourier Transform Infrared spectroscopy), we found that thetoxic protein (M8) adopted a structure rich in antiparallel β sheets greatly altered the integrity of lipid bilayers. Incontrast, the protein non-toxic (WT) organized in a structure rich in parallel β sheets in these interactions and it did notdisturb the homogeneity of the membranes. The toxicity of the protein M8 appears to be related to its differentorganization and its ability to rearrange membranes.Our results also support the hypothesis of the toxicity of amyloid oligomers.A study on the fabrication of a microfluidic cell for the separation of different aggregation states of amyloidproteins in order to detect these assemblies and to study their interaction with liposomes by infrared spectroscopy is presented. A CaF2 microfluidic cell with channels of 8 μm of thickness was obtained and was used for the detection of atested protein.

This work presents a whole-core benchmark problem based on a 2-loop pressurized water reactor with both UO₂and MOX fuel assemblies. The specification includes heterogeneity at both the assembly and core level. The geometry and material compositions are fully described and multi-group cross section libraries are provided in 2, 4, and 8 group formats. Simplifications made to the benchmark specification include a Cartesian boundary, to facilitate the use of transport codes that may have trouble with cylindrical boundaries, and control rod homogenization, to reduce the geometric complexity of the problem. These modifications were carefully chosen to preserve the physics of the problem and a justification of these modifications is given. Detailed Monte Carlo reference solutions including core eigenvalue, assembly averaged fission densities and selected fuel pin fission densities are presented for benchmarking diffusion and transport methods. Three different core configurations are presented in the paper namely all-rods-out, all-rods-in, and some-rods-in.

Cette thèse s'inscrit dans un programme initié par EDF afin de comprendre, de modéliser et de limiter la contamination du circuit primaire des réacteurs à eau sous pression par des particules colloïdales issues de la corrosion. Le comportement électrostatique de particules d'oxydes représentatives (ferrite de cobalt (CoFe2O4), ferrite de nickel (NiFe2O4) et la magnétite (Fe3O4)) a été étudié dans les conditions physico-chimiques du circuit primaire : en température (jusqu'à 320°C), en présence d'acide borique B(OH)3 et de lithine LiOH. Le point isoélectrique (PIE) et le point de charge nulle (PZC), mesurés entre 5 et 320°C, présentent, comme le pKe de l'eau, un minimum vers 200°C. Les constantes thermodynamiques de protonation ont été calculées. Le potentiel zêta et le PIE diminuent en présence d'acide borique, à cause de la sorption d'ions borate. La lithine n'a pas d'effet marqué. La modélisation de ces résultats dans des conditions représentatives du circuit primaire montre que ce type d'oxydes a une surface négative ce qui explique leurs propriétés de sorption et d'adhésion.