Dans le but d'étudier les possibilités de gestion avancée du plutonium dans les réacteurs actuels, 8 stratégies de multi-recyclage du plutonium dans des REP sont étudiées. Suite à des études à l'équilibre, il a été montré que, en utilisant des assemblages homogènes, l'utilisation du thorium ne permet pas de réduire les inventaires de plutonium en cycle à l'équilibre ni la production d'américium. En répartissant les différents types de combustibles au sein d'un même assemblage, certaines stratégies thoriées permettent cependant d'obtenir des inventaires et des productions d'américium inférieur aux meilleures stratégies utilisant uniquement le cycle uranium. Cependant dans tous les cas, le faible taux de conversion des combustibles thoriés en REP conduit à l'impossibilité de faire baisser la consommation de ressources de plus de quelques pour-cent par rapport aux stratégies sans thorium. Afin d'étudier la transition, une participation active au développement du code de scénario CLASS a eu lieu. Elle a aboutie à la simulation de deux scénarios parmi ceux étudiés à l'équilibre avec CLASS. Ces simulations ont montrées des écarts avec les scénarios simulés précédemment. Les causes majeures de ces écarts ont été identifiées et quantifiées. / In order to study the possibility of advanced management of plutonium in existing reactors, 8 strategies for plutonium multi-recycling in PWRs are studied. Following equilibrium studies, it was shown that, by using homogeneous assemblies, the use of thorium cannot reduce the plutonium inventory of equilibrium cycle or production of americium. By distributing the different fuel types within the same assembly, some thoriated strategies allow however lower inventories and lower production americium best strategies using only the uranium cycle. However, in all cases, low fuel conversion theories in PWRs makes it impossible to lower resource consumption more than a few percent compared to strategies without thorium. To study the transition, active participation in development of the scenario code CLASS has been taken. It led to the two simulation scenarios among those studied in equilibrium with CLASS. These simulations have shown discrepancies with previously simulated scenarios. The major causes of these differences were identified and quantified.

京都大学 / 新制・課程博士 / 博士(エネルギー科学) / 甲第24714号 / エネ博第457号 / 新制||エネ||86(附属図書館) / 京都大学大学院エネルギー科学研究科エネルギー基礎科学専攻 / (主査)教授 齊藤 泰司, 教授 三澤 毅, 教授 横峯 健彦 / 学位規則第4条第1項該当 / Doctor of Energy Science / Kyoto University / DFAM

The U.S. Evolutionary Power Reactor (EPR) is a new, large-scale pressurized water reactor made by AREVA NP Inc. Surrounding the core of this reactor is a steel wall structure sitting inside called the heavy reflector. The purpose of the heavy reflector is to reduce the neutron flux escaping the core and thus increase the efficiency of the reactor while reducing the damage to the structures surrounding the core as well. The heavy reflector is heated due to absorption of the gamma radiation, and this heat is removed by the water flowing through 832 cooling channels drilled through the heavy reflector. In this project, the temperature distribution in the heavy reflector was investigated to ascertain that the maximum temperature does not exceed the allowable temperature of 350 C, with the intent of modifying the flow distribution in the cooling channels to alleviate any hot spots. The analysis was conducted in two steps. First, the flow distribution in the cooling channels was calculated to test for any maldistribution. The temperature distribution in the heavy reflector was then calculated by simulating the conjugate heat transfer with this flow distribution as the coolant input. The turbulent nature of the flow through the cooling channels made the calculation of the flow distribution computationally expensive. In order to resolve this problem, a simplification method using the "equivalent flow resistance" was developed. The method was validated by conducting a few case studies. Using the simplified model, the flow distribution was calculated and was found to be fairly uniform. The conjugate heat transfer calculation was conducted. The same simplification method used in the flow distribution analysis could not be applied to this calculation; therefore, the computational cost of this model was reduced by lowering the grid density in the fluid region. The results showed that the maximum temperature in the heavy reflector is 347.7 C, which is below the maximum allowable temperature of 350 C. Additional studies were conducted to test the sensitivity of the maximum temperature with change in the flow distribution in the cooling channels. Through multiple calculations, the maximum temperature did not drop more than 3 C; therefore, it was concluded that the flow distribution in the cooling channels does not have significant effect on the maximum temperature in the heavy reflector. / Master of Science

Neste trabalho são comparados os resultados analíticos de previsão de falha de um vaso nuclear PWR destinado à propulsão naval, contendo oito diferentes tipos de trinca, obtidos segundo a Mecânica da Fratura Elástica-Linear (MFEL) e a correspondente Elasto-Plástica (MFEP). Os resultados são fornecidos em termos da pressão interna e do gradiente de deformações ao longo da espessura da parede do componente estrutural, sendo diretamente aplicáveis às condições impostas e desenvolvidas em ensaios isotérmicos de hidro-pressurização. O elevado nível de conservativismo dos resultados derivados segundo a MFEL, via critério KIC de disparo de clivagem, quando comparados àqueles obtidos por meio da MFEP, através do critério de iniciação de rasgamento dúctil, Ji, e em especial dos critérios J50 e Jinstab.dúctil de falha com base na Teoria da Instabilidade do Rasgamento Dúctil ou Elasto-Plástico (TIRD/TIREP) por intermédio de diagramas J x T (integral-J versus módulo de rasgamento), foi avaliado e quantificado. O intenso efeito das dimensões dos corpos de prova nos resultados obtidos a partir de ambas abordagens, assim como o papel determinante dos entalhes laterais nos ensaios da MFEP foram definitivamente comprovados. Foi também demonstrado o saudável conservativismo das previsões segundo o critério J50 para projeto e serviço seguros de plantas nucleares frente às condições reais em que se verifica analiticamente a instabilidade dúctil do componente trincado, tais como definidas por Jinstab.dúctil, assim como das previsões decorrentes dos processos de extrapolação linear de curvas J-R no espaço J x T quando comparadas às resultantes de extrapolações não-lineares mais frequentemente utilizadas. Com relação aos efeitos exercidos nos resultados de previsão de falha pelo tipo, profundidade e comprimento da trinca contida na parede do vaso de pressão, ficou demonstrado que as trincas de superfície são muito mais críticas que as sub-superficiais de idênticas profundidade e comprimento, assim como que as avaliações de integridade são muito mais sensíveis a mudanças na profundidade da trinca relativamente a alterações no seu comprimento. As aproximações entre os resultados obtidos via J50 e através de Jinstab.dúttil, e entre os resultados obtidos pelos procedimentos de extrapolação linear e não-linear de curvas JMAT-TMAT, são muito mais nítidas para as trincas de superfície menos profundas e, em especial para as embutidas, mesmo aquelas mais críticas. Deste modo, a metodologia aqui aprimorada, destinada em última instância a avaliações de integridade estrutural de vasos PWR em transientes de potência durante a sua vida em serviço, se mostrou absolutamente efetiva no caso das trincas mais prováveis em tais circunstâncias, quais sejam, as de pequenas dimensões e situadas superficialmente na parede do vaso. Ainda, foi verificada a tendência da ocorrência do vazamento do meio pressurizante anteriormente à falha do vaso por instabilidade dúctil nos casos da presença das trincas mais profundas, independentemente de serem elas de superfície ou sub-superficiais e, novamente, de seu comprimento, contribuindo assim para a percepção visual da iminência da falha bem como para eventos de freamento de trinca dúctil, impedindo a falha catastrófica do vaso PWR. Por outro lado, trincas rasas ou meno·s profundas tendem a entrar no regime de instabilidade dúctil muito precocemente, com a essência da propagação se dando de modo instável, causando fratura rápida ou súbita do vaso sem sinal anterior. Finalmente, foi comprovada nesse estudo a aplicabilidade da técnica analítica de normalização linear para o levantamento de curvas J-R, e como consequência JMAT-TMAT, para a avaliação de integridade estrutural de componentes robustos falhando de modo dúctil após considerável, e mensurável, crescimento estável de trinca; também, foi aqui proposto o ajuste logarítmico de dados de curvas J-R, em oposição ao hoje consensualmente utilizado via lei de potência, de modo a computar os efeitos do fenômeno de saturação de J, permitindo assim extrapolações conservativas e seguras para as condições em que se verifica a falha do componente nuclear. / In this work the analytical results related to a nuclear pressure vessel failure conditions as obtained by Linear Elastic Fracture Mechanics (LEFM) and the Elastic-Plastic Fracture Mechanics (EPFM) were compared for eigth different crack types. The results are supplied in terms of internal pressure and strain gradient through the wall thickness of the structural component, and might be directly applied to isothermal hydrotests experiments. The high conservatism of LEFM approach, by using KIC criterion of cleavage triggering, as compared to results obtained via Ji of ductile crack initiation from EPFM approach, and specially to the J50 and Jduct.instab. criteria based on ductile Tearing Instability Theory through J x T (J-integral x tearing modulus) diagrams, was evaluated and quantified. The overriding effect of specimens size over LEFM and EPFM results were definitely proved, as well as the very strong influence of side-grooving the EPFM specimens. It was demonstrated the healthy conservatism of the predictions following the J50 criterion for safe design and service in the nuclear industry facing the real conditions where ductile instability takes place as analytically defined by Jduct.instab.. In the same way the conservatism of the predictions from the linear extrapolation procedure of J-R curves on J x T space, as compared with the most common non-linear process was proved as well. With respect to the effect exerted by type, depth and length of the crack over the results it was demonstrated that surface cracks are more critical than the embedded ones with identical dimensions, being the analytical evaluation much more sensitive to changes in the crack depth than in its length. The approximations among the results obtained from J50 and Jduct.instab. criteria as well as among the results from the linear and non-linear extrapolation procedure are best verified for the less depth surface cracks and specially for the embedded cracks even the largest ones. So, the methodology here improved, aimed in last instance to structural integrity assessment of PWR vessels on power transients during their service life has been shown to be a powerful tool in evaluating the most expected cracks under such circunstances, that is, those having small dimensions and located on surface wall. Still, it was demonstrated the trend for leak-before-ductile instability in the presence of the deepest surface and internal cracks, allowing the visual perception of the threshold of the vessel failure and the development of crack arrest events avoiding catastrofic rupture of the component. Otherwise, shallow cracks are prone to get very fast in the unstable regime of growth, with the essence of the propagation occuring under this condition, conducting to suddenly fracture of the vessel without any prior evidence. Finally, it was proved in this study the accuracy of the linear normalization technique in deriving J-R curves for the assessment of structural integrity of massive components which fail in a ductile way after undergo large stable crack extensions; it was suggested here the logarithmic fit of J-R data points taking in account more properly the J-saturation phenomena and leading to more safe and conservative extrapolation results when compared to resulting from power-law fit.

Single-phase CFD calculations are already widely applied in nuclear industry for the thermal-hydraulic design optimization of fuel assemblies (FA). In contrast, two-phase CFD-applications are still in the state of development. The work presented in this thesis shows contributions towards the detailed two-phase modeling of boiling flows under swirl conditions in sub-channel geometries of pressurized water reactors (PWR) FAs, including a realistic description of the critical heat flux (CHF)-phenomena and identification of two-phase indicators characterizing CHF-phenomena. The numerical simulations were conducted with a 3-D CFD code (STAR CD) for various types of swirl generating components in FA. New insights regarding local void distribution in sub-channels under swirl conditions were obtained, which are relevant for CHF (“bubble pockets”). Furthermore, an enhanced wall partitioning model provides a more realistic description of the steep increase of the rod temperature due to CHF. Presented validation studies showed good agreement with the available experiments under PWR conditions for the radial void distributions at non-CHF conditions as well as for the CHF prediction. The results performed in this thesis show the incentives and chances of two-phase CFD applications for the development of thermal-hydraulically optimized PWR spacer grids with regard to heat transfer and improvement to the CHF behaviour. / Einphasige CFD Analysen stellen für typische thermohydraulische Fragestellungen aus dem Kernkraftwerksbereich bereits jetzt ein wichtiges Werkzeug für die Brennelement (BE)¬Auslegung dar. Die zweiphasige CFD-Modellierung befindet sich dagegen in der Entwicklungsphase. Die in dieser Dissertation präsentierten Arbeiten zeigen Fortschritte für die detaillierte zweiphasige Modellierung drallbehafteter Strömungen im Unterkanal eines BEs des Druckwasserreaktors (DWR) einschließlich der realistischen Beschreibung des kritischen Wärmestroms (CHF) sowie die Bestimmung zweiphasiger Indikatoren, welche das Auftreten von CHF-Phänomen beschreiben. Verschiedene drallerzeugende Komponenten im BE wurden mit dem 3-D CFD-Code STAR-CD modelliert. Es wurden neue Erkenntnisse zur lokalen Blasenverteilung in Unterkanälen unter Drallbedingungen gewonnen, die für CHF relevant sind (“bubble pockets”). Durch eine Modifikation des Wärmeübergangsmodells (heat partitioning model) wird der starke Anstieg der Stabtemperatur infolge CHF realistischer beschrieben. Die durchgeführten Validierungen zeigen eine gute Übereinstimmung mit verfügbaren Experimenten unter DWR-Bedingungen für die radialen Blasenverteilungen und für die Bestimmung von CHF. Die vorliegenden Ergebnisse dieser Dissertation verdeutlichen den Nutzen und die Möglichkeiten von zweiphasigen CFD-Anwendungen für die Entwicklung und die thermohydraulische Optimierung von DWR-Abstandshaltern bezüglich des Wärmeübergangs und der Verbesserung des CHF-Verhaltens. / Модели однофазовой среды в вычислительной гидродинамике (англ. CFD) являются уже сейчас важным инструментом для решения типичных термогидравлических задач в ядерной энергетике, например, при конструировании тепловыделяющих сборок (ТВС). Двухфазовое моделирование, в сравнении с однофазовым, находится на сегодняшний момент в стадии развития. Данная диссертация связана с совершенствованием двухфазовой модели в направлении улучшения детализации вихревого течения внутри ячейки ТВС в условиях работы двухконтурного ядерного реактора (тип PWR). Также в работе показана возможность более реалистичного описания феномена кризиса теплоотдачи и определения двухфазовых индикаторов, характеризующих критический тепловой поток (англ. CHF). Комплекс расчетов выполнен для различных интенсификаторов теплоотдачи, размещенных на дистанционирующих решетках TBC. Расчеты производились с помощью программного обеспечения STAR-CD, позволяющего моделировать трехмерные гидродинамические системы. Полученные новые данные о локальном распределении пузырьков в ячейках с вихревым течением дали важную информацию для идентификации критического теплового потока. Усовершенствование модели теплоотдачи на поверхности тепловыделяющего элемента (ТВЭЛ) позволило более реалистично описать резкое повышение температуры на поверхности ТВЭЛа при достижении критических условий. Полученные результаты численного моделирования для радиальных распределений пузырьков при нормальных условиях работы PWR и также для определения критического теплового потока показали хорошее совпадение с известными экспериментальными данными. Представленные в диссертации результаты показывают возможности применения двухфазовых CFD-расчётов для разработки и термогидравлической оптимизации дистанционирующих решеток с целью улучшения теплообмена и характеристик критического теплового потока в двухконтурном ядерном реакторе типа PWR.

Neste trabalho são comparados os resultados analíticos de previsão de falha de um vaso nuclear PWR destinado à propulsão naval, contendo oito diferentes tipos de trinca, obtidos segundo a Mecânica da Fratura Elástica-Linear (MFEL) e a correspondente Elasto-Plástica (MFEP). Os resultados são fornecidos em termos da pressão interna e do gradiente de deformações ao longo da espessura da parede do componente estrutural, sendo diretamente aplicáveis às condições impostas e desenvolvidas em ensaios isotérmicos de hidro-pressurização. O elevado nível de conservativismo dos resultados derivados segundo a MFEL, via critério KIC de disparo de clivagem, quando comparados àqueles obtidos por meio da MFEP, através do critério de iniciação de rasgamento dúctil, Ji, e em especial dos critérios J50 e Jinstab.dúctil de falha com base na Teoria da Instabilidade do Rasgamento Dúctil ou Elasto-Plástico (TIRD/TIREP) por intermédio de diagramas J x T (integral-J versus módulo de rasgamento), foi avaliado e quantificado. O intenso efeito das dimensões dos corpos de prova nos resultados obtidos a partir de ambas abordagens, assim como o papel determinante dos entalhes laterais nos ensaios da MFEP foram definitivamente comprovados. Foi também demonstrado o saudável conservativismo das previsões segundo o critério J50 para projeto e serviço seguros de plantas nucleares frente às condições reais em que se verifica analiticamente a instabilidade dúctil do componente trincado, tais como definidas por Jinstab.dúctil, assim como das previsões decorrentes dos processos de extrapolação linear de curvas J-R no espaço J x T quando comparadas às resultantes de extrapolações não-lineares mais frequentemente utilizadas. Com relação aos efeitos exercidos nos resultados de previsão de falha pelo tipo, profundidade e comprimento da trinca contida na parede do vaso de pressão, ficou demonstrado que as trincas de superfície são muito mais críticas que as sub-superficiais de idênticas profundidade e comprimento, assim como que as avaliações de integridade são muito mais sensíveis a mudanças na profundidade da trinca relativamente a alterações no seu comprimento. As aproximações entre os resultados obtidos via J50 e através de Jinstab.dúttil, e entre os resultados obtidos pelos procedimentos de extrapolação linear e não-linear de curvas JMAT-TMAT, são muito mais nítidas para as trincas de superfície menos profundas e, em especial para as embutidas, mesmo aquelas mais críticas. Deste modo, a metodologia aqui aprimorada, destinada em última instância a avaliações de integridade estrutural de vasos PWR em transientes de potência durante a sua vida em serviço, se mostrou absolutamente efetiva no caso das trincas mais prováveis em tais circunstâncias, quais sejam, as de pequenas dimensões e situadas superficialmente na parede do vaso. Ainda, foi verificada a tendência da ocorrência do vazamento do meio pressurizante anteriormente à falha do vaso por instabilidade dúctil nos casos da presença das trincas mais profundas, independentemente de serem elas de superfície ou sub-superficiais e, novamente, de seu comprimento, contribuindo assim para a percepção visual da iminência da falha bem como para eventos de freamento de trinca dúctil, impedindo a falha catastrófica do vaso PWR. Por outro lado, trincas rasas ou meno·s profundas tendem a entrar no regime de instabilidade dúctil muito precocemente, com a essência da propagação se dando de modo instável, causando fratura rápida ou súbita do vaso sem sinal anterior. Finalmente, foi comprovada nesse estudo a aplicabilidade da técnica analítica de normalização linear para o levantamento de curvas J-R, e como consequência JMAT-TMAT, para a avaliação de integridade estrutural de componentes robustos falhando de modo dúctil após considerável, e mensurável, crescimento estável de trinca; também, foi aqui proposto o ajuste logarítmico de dados de curvas J-R, em oposição ao hoje consensualmente utilizado via lei de potência, de modo a computar os efeitos do fenômeno de saturação de J, permitindo assim extrapolações conservativas e seguras para as condições em que se verifica a falha do componente nuclear. / In this work the analytical results related to a nuclear pressure vessel failure conditions as obtained by Linear Elastic Fracture Mechanics (LEFM) and the Elastic-Plastic Fracture Mechanics (EPFM) were compared for eigth different crack types. The results are supplied in terms of internal pressure and strain gradient through the wall thickness of the structural component, and might be directly applied to isothermal hydrotests experiments. The high conservatism of LEFM approach, by using KIC criterion of cleavage triggering, as compared to results obtained via Ji of ductile crack initiation from EPFM approach, and specially to the J50 and Jduct.instab. criteria based on ductile Tearing Instability Theory through J x T (J-integral x tearing modulus) diagrams, was evaluated and quantified. The overriding effect of specimens size over LEFM and EPFM results were definitely proved, as well as the very strong influence of side-grooving the EPFM specimens. It was demonstrated the healthy conservatism of the predictions following the J50 criterion for safe design and service in the nuclear industry facing the real conditions where ductile instability takes place as analytically defined by Jduct.instab.. In the same way the conservatism of the predictions from the linear extrapolation procedure of J-R curves on J x T space, as compared with the most common non-linear process was proved as well. With respect to the effect exerted by type, depth and length of the crack over the results it was demonstrated that surface cracks are more critical than the embedded ones with identical dimensions, being the analytical evaluation much more sensitive to changes in the crack depth than in its length. The approximations among the results obtained from J50 and Jduct.instab. criteria as well as among the results from the linear and non-linear extrapolation procedure are best verified for the less depth surface cracks and specially for the embedded cracks even the largest ones. So, the methodology here improved, aimed in last instance to structural integrity assessment of PWR vessels on power transients during their service life has been shown to be a powerful tool in evaluating the most expected cracks under such circunstances, that is, those having small dimensions and located on surface wall. Still, it was demonstrated the trend for leak-before-ductile instability in the presence of the deepest surface and internal cracks, allowing the visual perception of the threshold of the vessel failure and the development of crack arrest events avoiding catastrofic rupture of the component. Otherwise, shallow cracks are prone to get very fast in the unstable regime of growth, with the essence of the propagation occuring under this condition, conducting to suddenly fracture of the vessel without any prior evidence. Finally, it was proved in this study the accuracy of the linear normalization technique in deriving J-R curves for the assessment of structural integrity of massive components which fail in a ductile way after undergo large stable crack extensions; it was suggested here the logarithmic fit of J-R data points taking in account more properly the J-saturation phenomena and leading to more safe and conservative extrapolation results when compared to resulting from power-law fit.

Les Rotavirus appartiennent à la famille des Reoviridae, famille du groupe III des virus à ARN double brin. Identifiés en 1973 par Ruth Bishop, ces virus non enveloppés sont la première cause de diarrhée aiguë sévère du jeune enfant dans le monde. La capside virale icosaédrique est constituée de 3 couches protéiques de structure : la couche externe formée par la glycoprotéine VP7 d’où émergent les spicules de protéine VP4, la couche intermédiaire constituée par la protéine VP6 représentant près de 50 % du poids du virus et enfin, la couche interne appelée core, résultant de l’assemblage des protéines VP2, d’où émergent vers l’intérieur les protéines VP1 et VP3. Cette capside renferme un génome divisé en 11 segments d’ARN bicaténaires. A ces 6 protéines structurales s’ajoutent les protéines non structurales qui interviennent lors de la réplication du virus. Les deux protéines structurales, VP4 et VP7 sont essentielles pour la fixation de la particule triple couche (TLP) aux membranes des cellules hôtes, par interaction aux récepteurs intégrines, elle sont également impliqués dans la déstabilisation des membranes endosomales, indispensable à la libération de la particule double couche (DLP) infectieuse dans le cytoplasme. Actuellement, contrairement au mécanisme d’action de la protéine VP5*, celui de la glycoprotéine VP7 est inconnu. L’objectif de cette thèse, a été de comprendre le mécanisme moléculaire de déstabilisation des membranes par les peptides dérivés de VP7. Dans un premier temps nous avons montré, par des études in silico, l’existence d’un domaine prédit en hélice membranaire bordé de résidus arginine et lysine hautement conservés, situé à l’extrémité C-terminale de la glycoprotéine VP7. Ces résultats ont conduit à la synthèse de quatre peptides avec lesquels des tests de perméabilisation de membranes modèles de larges vésicules unilamellaires (LUVs) ont été menés. Ceux-ci ont permis d’identifier le domaine minimum le plus actif, VP723, parmi les peptides sélectionnés. Dans un second temps nous avons déterminé la structure de ces peptides par RMN, dans des conditions mimant l’environnement hydrophobe de la membrane. Le peptide minimal VP723 s’organise en hélice α-amphipathique, structure souvent impliquée dans la déstabilisation des membranes cellulaires. La comparaison de sa structure obtenue par RMN à celle du domaine correspondant dans la structure cristallographique de la protéine native montre le réarrangement conformationnel de ce segment après maturation par la trypsine. Ces résultats ont été confirmés par deux mutants de synthèse, dont l’un est inactif pour la perméabilisation des membranes modèles. Ces travaux ont été complétés par des expériences de Résonance Plasmonique aux Ondes guidée (PWR). Des études par RMN du solide sont en cours afin de déterminer l’orientation du peptide dans les membranes modèles. En conclusion, nos résultats mettent en évidence l’importance du domaine C-terminal VP723 de la protéine VP7 dans la déstabilisation des membranes, permettant d’assurer la translocation de la particule virale infectieuse (DLP) de l’endosome vers le cytoplasme. Un modèle du mécanisme d’entrée du virus, médié par les peptides dérivés de la maturation par la trypsine de la glycoprotéine VP7 est proposé. / Rotaviruses belong to the Reoviridae family, belonging to the group III of dsRNA viruses. Identified in 1973 by Ruth Bishop, these non-enveloped viruses are the leading cause of severe diarrhea in young children worldwide. The icosahedral capsid is composed of three structural protein layers: the outer one, formed by the glycoprotein VP7, emerges spicules protein VP4, the intermediate one consists of VP6 protein representing nearly 50% of the weight of the virus and finally, the inner one called core, results from the assembly of proteins VP2, emerges towards the inside of proteins VP1 and VP3. The capsid contains a genome divided into 11 segments of dsRNA. To these six structural proteins are added nonstructural proteins involved in virus replication. The two structural proteins, VP4 and VP7, are involved in the interaction of the triple layer particle (TLP) to integrin receptors, necessary for the release of the infectious double layer particle (DLP) into the cytoplasm following the permeabilization of the membrane of the endosome compartments. Currently, unlike the mechanism of action of the protein VP5*, the glycoprotein VP7 remains unknown. The objective of this work was to understand the molecular mechanism involved in the destabilization of membranes by peptides derived from VP7. In a first step, we have shown, by in silico studies, the existence of a helical trans-membrane domain predicted containing a highly conserved arginine and lysine residues, located at the C-terminus of the VP7 glycoprotein. These results led to the synthesis of four peptides with which permeabilizing tests of model membranes were conducted. We have identified the minimum of the most active domain, named VP723, among the selected peptides. In a second step, we determined the structure of these peptides by NMR under conditions mimicking the hydrophobic environment of the membrane. The VP723 peptide is organized like an α-helical amphipathic structure often involved in the destabilization of cell membranes. The comparison of the structure obtained by NMR to that of the corresponding domain in the crystallographic structure of the native protein shows a conformational rearrangement of the segment after trypsin maturation. These results were confirmed by two synthetic mutants, one of which is inactive for the permeabilization of model membranes. These studies were complemented by experiments Plasmon Resonance guided the Waves (PWR). Studies by solid state NMR are in progress to determine the orientation of the peptide in model of membranes. In conclusion, our results highlight the importance of the C-terminal domain of the VP7 protein, named VP723, in the destabilization of membranes, to ensure the translocation of the infectious viral particle (DLP) from the endosome into the cytoplasm compartments. A mechanism of virus entry mediated by peptides derived from trypsin maturation of the VP7 glycoprotein is proposed in this study.

Single-phase CFD calculations are already widely applied in nuclear industry for the thermal-hydraulic design optimization of fuel assemblies (FA). In contrast, two-phase CFD-applications are still in the state of development. The work presented in this thesis shows contributions towards the detailed two-phase modeling of boiling flows under swirl conditions in sub-channel geometries of pressurized water reactors (PWR) FAs, including a realistic description of the critical heat flux (CHF)-phenomena and identification of two-phase indicators characterizing CHF-phenomena. The numerical simulations were conducted with a 3-D CFD code (STAR CD) for various types of swirl generating components in FA. New insights regarding local void distribution in sub-channels under swirl conditions were obtained, which are relevant for CHF (“bubble pockets”). Furthermore, an enhanced wall partitioning model provides a more realistic description of the steep increase of the rod temperature due to CHF. Presented validation studies showed good agreement with the available experiments under PWR conditions for the radial void distributions at non-CHF conditions as well as for the CHF prediction. The results performed in this thesis show the incentives and chances of two-phase CFD applications for the development of thermal-hydraulically optimized PWR spacer grids with regard to heat transfer and improvement to the CHF behaviour. / Einphasige CFD Analysen stellen für typische thermohydraulische Fragestellungen aus dem Kernkraftwerksbereich bereits jetzt ein wichtiges Werkzeug für die Brennelement (BE)¬Auslegung dar. Die zweiphasige CFD-Modellierung befindet sich dagegen in der Entwicklungsphase. Die in dieser Dissertation präsentierten Arbeiten zeigen Fortschritte für die detaillierte zweiphasige Modellierung drallbehafteter Strömungen im Unterkanal eines BEs des Druckwasserreaktors (DWR) einschließlich der realistischen Beschreibung des kritischen Wärmestroms (CHF) sowie die Bestimmung zweiphasiger Indikatoren, welche das Auftreten von CHF-Phänomen beschreiben. Verschiedene drallerzeugende Komponenten im BE wurden mit dem 3-D CFD-Code STAR-CD modelliert. Es wurden neue Erkenntnisse zur lokalen Blasenverteilung in Unterkanälen unter Drallbedingungen gewonnen, die für CHF relevant sind (“bubble pockets”). Durch eine Modifikation des Wärmeübergangsmodells (heat partitioning model) wird der starke Anstieg der Stabtemperatur infolge CHF realistischer beschrieben. Die durchgeführten Validierungen zeigen eine gute Übereinstimmung mit verfügbaren Experimenten unter DWR-Bedingungen für die radialen Blasenverteilungen und für die Bestimmung von CHF. Die vorliegenden Ergebnisse dieser Dissertation verdeutlichen den Nutzen und die Möglichkeiten von zweiphasigen CFD-Anwendungen für die Entwicklung und die thermohydraulische Optimierung von DWR-Abstandshaltern bezüglich des Wärmeübergangs und der Verbesserung des CHF-Verhaltens. / Модели однофазовой среды в вычислительной гидродинамике (англ. CFD) являются уже сейчас важным инструментом для решения типичных термогидравлических задач в ядерной энергетике, например, при конструировании тепловыделяющих сборок (ТВС). Двухфазовое моделирование, в сравнении с однофазовым, находится на сегодняшний момент в стадии развития. Данная диссертация связана с совершенствованием двухфазовой модели в направлении улучшения детализации вихревого течения внутри ячейки ТВС в условиях работы двухконтурного ядерного реактора (тип PWR). Также в работе показана возможность более реалистичного описания феномена кризиса теплоотдачи и определения двухфазовых индикаторов, характеризующих критический тепловой поток (англ. CHF). Комплекс расчетов выполнен для различных интенсификаторов теплоотдачи, размещенных на дистанционирующих решетках TBC. Расчеты производились с помощью программного обеспечения STAR-CD, позволяющего моделировать трехмерные гидродинамические системы. Полученные новые данные о локальном распределении пузырьков в ячейках с вихревым течением дали важную информацию для идентификации критического теплового потока. Усовершенствование модели теплоотдачи на поверхности тепловыделяющего элемента (ТВЭЛ) позволило более реалистично описать резкое повышение температуры на поверхности ТВЭЛа при достижении критических условий. Полученные результаты численного моделирования для радиальных распределений пузырьков при нормальных условиях работы PWR и также для определения критического теплового потока показали хорошее совпадение с известными экспериментальными данными. Представленные в диссертации результаты показывают возможности применения двухфазовых CFD-расчётов для разработки и термогидравлической оптимизации дистанционирующих решеток с целью улучшения теплообмена и характеристик критического теплового потока в двухконтурном ядерном реакторе типа PWR.

No description available.

This thesis is centered around the computional program PASGO. PASGO is used for calculations for making performance monitoring of the steam generators in Ringhals reactor 3 and 4, especially concerning fouling of the steam generator tubes. The main goal of this thesis project was to convert the program code from the programming language Turbo Pascal 6.0 into a more modern language that works with the current Windows operative systems. Apart from a straight conversion of the code, a function to make automatic input from an Excel-file to the program was also requested. The thesis also includes analyzing the correlations used for heat transfer in PASGO, to determine if these can be replaced with more exact correlations. The Dittus-Boelter correlation was used in the original program, and while it was not a bad correlation, studies show that the more modern Gnielinski correlation has a smaller margin of error. The program code was converted successfully into C++ in Visual Studio, which was the programming language of choice. A function to enable automatic input of data from Excel-files was successfully implemented as well. This function allows the user to run performance calculations regarding tube fouling and pressure correction values for hundreds of performance tests at the same time. Versions of PASGO using the Gnielinski correlation were also created. These versions are recommended for future tests, while the version using the Dittus Boelter correlation is recommended for tests which will be compared to studies made white the previous version of PASGO.