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Les boucles fluides diphasiques à pompage thermo-capillaire (BFDPT) ont été identifiées comme des solutions de transport de chaleur alternatives dans le contexte des transports terrestres, afin de répondre au contrôle thermique de l’électronique de puissance. Le développement d’une architecture particulière de boucle, mettant à profit la gravité, a été mené afin d’adapter la capacité des BFDPT à ces nouvelles contraintes. Les investigations expérimentales et numériques ont permis de montrer le fort potentiel de ce nouveau système et amènent aujourd’hui à sa future utilisation industrielle.La conception d’une nouvelle boucle expérimentale avec trois évaporateurs, en parallèle, a pour objectif d’élargir son utilisation comme « bus thermique ». Le banc expérimental, finement instrumenté, a pour but d’approfondir les études antérieures et de caractériser les réponses de cette boucle à une application de puissance, en régime permanent et transitoire, avec un ou plusieurs évaporateurs. L’analyse des résultats obtenus met en évidence les nombreux couplages entre évaporateurs, réservoir et condenseur,notamment lors de phases transitoires sévères, et confirme la capacité du système à gérer le contrôle thermique de l’électronique quelles que soient les puissances appliquées sur les différents composants.Les résultats du travail de modélisation, basée sur la méthode nodale, s’avèrent prédire correctement le comportement thermohydraulique transitoire de la boucle, en fonctionnement nominal, en mode mono ou multi-évaporateurs. Malgré une représentation des évaporateurs et du réservoir encore simplifiée et tributaire de l’expérience, le modèle se révèle être un très bon outil de dimensionnement et d’analyse. / Two-phase fluid capillary pumped loop (TFCPL) have been identified as heat transfers alternative solutionsfor thermal control of power electronics, in the context of ground transportation. A special gravityloop has been improved to adapt TFCPL capacity to these new terrestrial constraints. Experimentaland numerical investigations have proved the great potential of this kind of system and lead today toits upcoming industrial use.The aim is now to employ this new system as « thermal busbar ». A new experimental loop withthree parallel evaporators was built to go further into the earlier studies. Thanks to many measurementinstruments, the goal is to define steady and transient thermohydraulic responses to a power solicitationwith a mono- or multi-evaporators loop. Results highlight many coupling between evaporators, reservoirand condenser, especially during strong transient phases. The thermal control of power electronics, withdifferent range of thermal dissipation on several separate areas, is furthermore demonstrated.The developed modeling is based on nodal method. Transient thermohydraulic behaviour of the gravityloop is very well predicted by numerical results during nominal operating, with one or more evaporators.Even if evaporators and reservoir models are still simplified and depend on empirical identifications, theglobal model of the loop proves to be a great design and analysis tool.

Many hours are spent by systemand control engineers deriving reduced order dynamicmodels portraying the dominant systemdynamics of thermohydraulic systems. A need therefore exists to develop a method that will automate the model derivation process. The model format preferred for control system design and analysis during preliminary system design is the state space format. The aim of this study is therefore to develop an automated and generic state space model extraction method that can be applied to thermohydraulic systems. Well developed system identification methods exist for obtaining state space models from input-output data, but these models are not transparent, meaning the parameters do not have any physical meaning. For example one cannot identify system parameters such as heat or mass transfer coefficients. Another approach is needed to derive state space models automatically. Many commercial thermohydraulic simulation codes follow a network approach towards the representation of thermohydraulic systems. This approach is probably one of the most advanced approaches in terms of technical development. It would therefore be useful to develop a state space extraction algorithm that would be able to derive reduced order state space models from network representations of thermohydraulic systems. In this regard a network approach is followed in the development of the state space extraction algorithm. The advantage of using a network-based extraction method is that the extracted state space model is transparent and the algorithm can be embedded in existing simulation software that follow a network approach. In this study an existing state space extraction algorithm, used for electrical network analysis, is modified and applied in a new way to extract state space models of thermohydraulic systems. A thermohydraulic system is partitioned into its respective physical domains which, unlike electrical systems, have multiple variables. Network representations are derived for each domain. The state space algorithm is applied to these network representations to extract symbolic state spacemodels. The symbolic parametersmay then be substitutedwith numerical values. The state space extraction algorithm is applied to small scale thermohydraulic systems such as a U-tube and a heat exchanger, but also to a larger, more complex system such as the Pebble Bed Modular Reactor Power Conversion Unit (PBMR PCU). It is also shown that the algorithm can extract linear, nonlinear, time-varying and time-invariant state space models. The extracted state space models are validated by solving the state space models and comparing the solutions with Flownex results. Flownex is an advanced and extensively validated thermo-fluid simulation code. The state space models compared well with Flownex results. The usefulness of the state space model extraction algorithm in model-based control system design is illustrated by extracting a linear time-invariant state space model of the PBMR PCU. This model is embedded in an optimal model-based control scheme called Model-Predictive Control (MPC). The controller is compared with standard optimised control schemes such as PID and Fuzzy PID control. The MPC controller shows superior performance compared to these control schemes. This study succeeded in developing an automated state space model extraction method that can be applied to thermohydraulic networks. Hours spent on writing down equations from first principles to derive reduced order models for control purposes can now be replaced with a click of a button. The need for an automated state space model extraction method for thermohydraulic systems has therefore been resolved / Thesis (Ph.D. (Computer and Electronical Engineering)--North-West University, Potchefstroom Campus, 2009.

Many hours are spent by systemand control engineers deriving reduced order dynamicmodels portraying the dominant systemdynamics of thermohydraulic systems. A need therefore exists to develop a method that will automate the model derivation process. The model format preferred for control system design and analysis during preliminary system design is the state space format. The aim of this study is therefore to develop an automated and generic state space model extraction method that can be applied to thermohydraulic systems. Well developed system identification methods exist for obtaining state space models from input-output data, but these models are not transparent, meaning the parameters do not have any physical meaning. For example one cannot identify system parameters such as heat or mass transfer coefficients. Another approach is needed to derive state space models automatically. Many commercial thermohydraulic simulation codes follow a network approach towards the representation of thermohydraulic systems. This approach is probably one of the most advanced approaches in terms of technical development. It would therefore be useful to develop a state space extraction algorithm that would be able to derive reduced order state space models from network representations of thermohydraulic systems. In this regard a network approach is followed in the development of the state space extraction algorithm. The advantage of using a network-based extraction method is that the extracted state space model is transparent and the algorithm can be embedded in existing simulation software that follow a network approach. In this study an existing state space extraction algorithm, used for electrical network analysis, is modified and applied in a new way to extract state space models of thermohydraulic systems. A thermohydraulic system is partitioned into its respective physical domains which, unlike electrical systems, have multiple variables. Network representations are derived for each domain. The state space algorithm is applied to these network representations to extract symbolic state spacemodels. The symbolic parametersmay then be substitutedwith numerical values. The state space extraction algorithm is applied to small scale thermohydraulic systems such as a U-tube and a heat exchanger, but also to a larger, more complex system such as the Pebble Bed Modular Reactor Power Conversion Unit (PBMR PCU). It is also shown that the algorithm can extract linear, nonlinear, time-varying and time-invariant state space models. The extracted state space models are validated by solving the state space models and comparing the solutions with Flownex results. Flownex is an advanced and extensively validated thermo-fluid simulation code. The state space models compared well with Flownex results. The usefulness of the state space model extraction algorithm in model-based control system design is illustrated by extracting a linear time-invariant state space model of the PBMR PCU. This model is embedded in an optimal model-based control scheme called Model-Predictive Control (MPC). The controller is compared with standard optimised control schemes such as PID and Fuzzy PID control. The MPC controller shows superior performance compared to these control schemes. This study succeeded in developing an automated state space model extraction method that can be applied to thermohydraulic networks. Hours spent on writing down equations from first principles to derive reduced order models for control purposes can now be replaced with a click of a button. The need for an automated state space model extraction method for thermohydraulic systems has therefore been resolved / Thesis (Ph.D. (Computer and Electronical Engineering)--North-West University, Potchefstroom Campus, 2009.

This document focuses on the procedure and results of creating a thermohydraulic model of the secondary cooling system of the SAFARI-1 research reactor at the Pelindaba facility of the South African Nuclear Energy Corporation (Necsa) to the west of Pretoria, South Africa. The secondary cooling system is an open recirculating cooling system that comprises an array of parallel-coupled heat exchangers between the primary systems and the main heat sink system, which consists of multiple counterflow-induced draught cooling towers. The original construction of the reactor was a turnkey installation, with no theoretical/technical support or verifiability. The design baseline is therefore not available and it is necessary to reverse-engineer a system that could be modelled and characterised. For the nuclear operator, it is essential to be able to make predictions and systematically implement modifications to improve system performance, such as to understand and modify the control system. Another objective is to identify the critical performance areas of the thermohydraulic system or to determine whether the cooling capacity of the secondary system meets the optimum original design characteristics. The approach was to perform a comprehensive one-dimensional modelling of all the available physical components, which was followed by using existing performance data to verify the accuracy and validity of the developed model. Where performance data is not available, separate analysis through computational fluid dynamics (CFD) modelling is performed to generate the required inputs. The results yielded a model that is accurate within 10%. This is acceptable when compared to the variation within the supplied data, generated and assumed alternatives, and when considering the compounding effect of the large amount of interdependent components, each with their own characteristics and associated performance uncertainties. The model pointed to potential problems within the current system, which comprised either an obstruction in a certain component or faulty measuring equipment. Furthermore, it was found that the current spray nozzles in the cooling towers are underutilised. It should be possible to use the current cooling tower arrangement to support a similar second reactor, although slight modifications would be required to ensure that the current system is not operated beyond its current limits. The interdependent nature of two parallel systems and the variability of the conditions that currently exist would require a similar analysis as the current model to determine the viability of using the existing cooling towers for an additional reactor. / Dissertation (MEng)--University of Pretoria, 2015. / Mechanical and Aeronautical Engineering / MEng / Unrestricted

Au cours des dernières décennies, le phénomène de quench a été une des problématiques les plus importantes abordées dans les conceptions d’aimants supraconducteurs. En effet, la transition de quench d’un aimant de son état supraconducteur à son état normal induit une grande quantité de l’énergie par effet Joule. Cet apport de chaleur va ensuite augmenter rapidement la température du conducteur ainsi que la pression du liquide de refroidissement à l’hélium. Le dépassement d’un certain seuil sur ces deux paramètres peut engendrer une détérioration irréversible à l’aimant et au système de refroidissement cryogénique. Afin de mettre en évidence les comportements de quench des bobines supraconductrices à champ toroïdal (TF) du Tokamak JT-60SA, nous avons réalisé des études expérimentales et numériques sur les phénomènes thermohydrauliques résultant du quench d’un aimant supraconducteur fabriqué en câble-en-conduit conducteur (CICC) et refroidi par l’écoulement forcé à l’hélium supercritique. Dans ce cadre, toutes les 18 TF bobines de JT-60SA ont été testées dans une configuration à une seule bobine dans leurs conditions de fonctionnement nominales de courant et de température (25,7 kA et 5 K). Une augmentation progressive de la température a été appliquée à l'entrée de l'hélium jusqu'à la température de quench, suivie d'une décharge rapide du courant dès que le quench est détecté pour protéger l'aimant. Les analyses expérimentales de ces tests ont d'abord permis d'identifier plusieurs phases dynamiques très différentes pendant toute la propagation de quench. Ensuite, les phénomènes physiques parcourant chacune de ces phases ont été étudiés et les plus prédominants ont été mis en évidence tels que les charges thermiques externes, les performances magnétiques des brins, les transferts thermiques conducto-convectifs entre conducteurs et hélium ou encore l'expulsion d'hélium et le reverse flow. Sur la base de ces analyses expérimentales, un modèle numérique d’une seule galette a été développé dans le code THEA afin d'analyser un phénomène physique à la fois sans construire un modèle global trop complexe de l'ensemble de l'aimant. Ce modèle d’une seule galette a été validé sur les données d'expériences de quench et a été appliqué avec succès pour faire d'autres analyses plus détaillées des phénomènes physiques ainsi que des phases dynamiques identifiées pendant la propagation de quench des TF bobines. Ce modèle numérique a même permis d'identifier certains phénomènes prépondérants qui n'ont pas pu être étudiés dans l'analyse expérimentale, tels que l'impact des instabilités des conditions de test sur la dynamique de quench. Les très bons résultats de ce modèle et sa cohérence avec les analyses physiques expérimentales en font une étape très intéressante vers la modélisation complète de toute la TF bobine de JT-60SA et l'étude de son comportement de quench dans une vraie machine Tokamak et non en conditions d'essais. / During the last decades, the quench phenomenon has been one of the most important issues addressed in the superconducting magnets designs. Indeed, the quench transition of a magnet from its superconducting state to its normal state induces a large deposition of the Joule effect energy leading to an abrupt temperature increase in the conductor as well as a large pressure rise in the helium coolant. Any excess of these two parameters can cause an irreversible damage either to the magnet or to the cryogenic system. In order to achieve a better understanding of the quench behavior of the TF coils in the superconducting Tokamak JT-60SA, we carried out both experimental and numerical studies of the thermohydraulic phenomena taking place during the quench of a superconducting magnet manufactured with Cable-In-Conduit Conductor and cooled in forced flow with supercritical helium. In this framework, all the 18 JT-60SA TF coils were tested in a single coil configuration at their nominal operating conditions of current and temperature (25.7kA and 5K). A progressive temperature increase has been applied to the helium inlet up to the quench temperature, followed by a current fast discharge as soon as the quench is detected to protect the magnet. The experimental analyses of these tests allowed first to identify several very different dynamic phases in the overall quench propagation time. Then, the physical phenomena driving each one of these phases have been studied and the most predominant ones have been highlighted such as the external heat loads, the strands magnetic performances, the conductive and convective heat transfers between conductors and helium or even the helium expulsion and reverse flow. Based on these experimental analyses, a single pancake numerical model has been developed in the THEA code in order to analyze one physical phenomenon at a time without building a too complex global model of the entire magnet. This single pancake model has been validated on the quench experiments data and has been successfully applied to make further more detailed analyses of the physical phenomena as well as the dynamic phases identified during the TF coils quench propagation. This numerical model even allowed identifying some driving physical phenomena that could not be studied in the experimental analysis, such as the impact of the testing conditions instabilities on the quench dynamics. The very good results of this model and its coherence with physical experimental analyses makes it a very interesting step towards the full modelling of the entire JT-60SA TF coil and the study of its quench behavior in real Tokamak and not test facility conditions.

This PhD study, developed at Universitat Politècnica de València (UPV), aims to cover the first phase of the benchmark released by the expert group on Uncertainty Analysis in Modeling (UAM-LWR). The main contribution to the benchmark, made by the thesis' author, is the development of a MATLAB program requested by the benchmark organizers. This is used to generate neutronic libraries to distribute among the benchmark participants. The UAM benchmark pretends to determine the uncertainty introduced by coupled multi-physics and multi-scale LWR analysis codes. The benchmark is subdivided into three phases: 1. Neutronic phase: obtain collapsed and homogenized problem-dependent cross sections and criticality analyses. 2. Core phase: standalone thermohydraulic and neutronic codes. 3. System phase: coupled thermohydraulic and neutronic code. In this thesis the objectives of the first phase are covered. Specifically, a methodology is developed to propagate the uncertainty of cross sections and other neutronic parameters through a lattice physics code and core simulator. An Uncertainty and Sensitivity (U&S) analysis is performed over the cross sections contained in the ENDF/B-VII nuclear library. Their uncertainty is propagated through the lattice physics code SCALE6.2.1, including the collapse and homogenization phase, up to the generation of problem-dependent neutronic libraries. Afterward, the uncertainty contained in these libraries can be further propagated through a core simulator, in this study PARCSv3.2. The module SAMPLER -available in the latest release of SCALE- and DAKOTA 6.3 statistical tool are used for the U&S analysis. As a part of this process, a methodology to obtain neutronic libraries in NEMTAB format -to be used in a core simulator- is also developed. A code-to-code comparison with CASMO-4 is used as a verification. The whole methodology is tested using a Boiling Water Reactor (BWR) reactor type. Nevertheless, there is not any concern or limitation regarding its use in any other type of nuclear reactor. The Gesellschaft für Anlagen und Reaktorsicherheit (GRS) stochastic methodology for uncertainty quantification is used. This methodology makes use of the high-fidelity model and nonparametric sampling to propagate the uncertainty. As a result, the number of samples (determined using the revised Wilks' formula) does not depend on the number of input parameters but only on the desired confidence and uncertainty of output parameters. Moreover, the output Probability Distribution Functions (PDFs) are not subject to normality. The main disadvantage is that each input parameter must have a pre-defined PDF. If possible, input PDFs are defined using information found in the related literature. Otherwise, the uncertainty definition is based on expert judgment. A second scenario is used to propagate the uncertainty of different thermohydraulic parameters through the coupled code TRACE5.0p3/PARCSv3.0. In this case, a PWR reactor type is used and a transient control rod drop occurrence is simulated. As a new feature, the core is modeled chan-by-chan following a fully 3D discretization. No other study is found using a detailed 3D core. This U&S analysis also makes use of the GRS methodology and DAKOTA 6.3. / Este trabajo de doctorado, desarrollado en la Universitat Politècnica de València (UPV), tiene como objetivo cubrir la primera fase del benchmark presentado por el grupo de expertos Uncertainty Analysis in Modeling (UAM-LWR). La principal contribución al benchmark, por parte del autor de esta tesis, es el desarrollo de un programa de MATLAB solicitado por los organizadores del benchmark, el cual se usa para generar librerías neutrónicas a distribuir entre los participantes del benchmark. El benchmark del UAM pretende determinar la incertidumbre introducida por los códigos multifísicos y multiescala acoplados de análisis de reactores de agua ligera. El citado benchmark se divide en tres fases: 1. Fase neutrónica: obtener los parámetros neutrónicos y secciones eficaces del problema específico colapsados y homogenizados, además del análisis de criticidad. 2. Fase de núcleo: análisis termo-hidráulico y neutrónico por separado. 3. Fase de sistema: análisis termo-hidráulico y neutrónico acoplados. En esta tesis se completan los principales objetivos de la primera fase. Concretamente, se desarrolla una metodología para propagar la incertidumbre de secciones eficaces y otros parámetros neutrónicos a través de un código lattice y un simulador de núcleo. Se lleva a cabo un análisis de incertidumbre y sensibilidad para las secciones eficaces contenidas en la librería neutrónica ENDF/B-VII. Su incertidumbre se propaga a través del código lattice SCALE6.2.1, incluyendo las fases de colapsación y homogenización, hasta llegar a la generación de una librería neutrónica específica del problema. Luego, la incertidumbre contenida en dicha librería puede continuar propagándose a través de un simulador de núcleo, para este estudio PARCSv3.2. Para el análisis de incertidumbre y sensibilidad se ha usado el módulo SAMPLER -disponible en la última versión de SCALE- y la herramienta estadística DAKOTA 6.3. Como parte de este proceso, también se ha desarrollado una metodología para obtener librerías neutrónicas en formato NEMTAB para ser usadas en simuladores de núcleo. Se ha realizado una comparación con el código CASMO-4 para obtener una verificación de la metodología completa. Esta se ha probado usando un reactor de agua en ebullición del tipo BWR. Sin embargo, no hay ninguna preocupación o limitación respecto a su uso con otro tipo de reactor nuclear. Para la cuantificación de la incertidumbre se usa la metodología estocástica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Esta metodología hace uso del modelo de alta fidelidad y un muestreo no paramétrico para propagar la incertidumbre. Como resultado, el número de muestras (determinado con la fórmula revisada de Wilks) no depende del número de parámetros de entrada, sólo depende del nivel de confianza e incertidumbre deseados de los parámetros de salida. Además, las funciones de distribución de probabilidad no están limitadas a normalidad. El principal inconveniente es que se ha de disponer de las distribuciones de probabilidad de cada parámetro de entrada. Si es posible, las distribuciones de probabilidad de entrada se definen usando información encontrada en la literatura relacionada. En caso contrario, la incertidumbre se define en base a la opinión de un experto. Se usa un segundo escenario para propagar la incertidumbre de diferentes parámetros termo-hidráulicos a través del código acoplado TRACE5.0p3/PARCSv3.0. En este caso, se utiliza un reactor tipo PWR para simular un transitorio de una caída de barra. Como nueva característica, el núcleo se modela elemento a elemento siguiendo una discretización totalmente en 3D. No se ha encontrado ningún otro estudio que use un núcleo tan detallado en 3D. También se usa la metodología GRS y el DAKOTA 6.3 para este análisis de incertidumbre y sensibilidad. / Aquest treball de doctorat, desenvolupat a la Universitat Politècnica de València (UPV), té com a objectiu cobrir la primera fase del benchmark presentat pel grup d'experts Uncertainty Analysis in Modeling (UAM-LWR). La principal contribució al benchmark, per part de l'autor d'aquesta tesi, es el desenvolupament d'un programa de MATLAB sol¿licitat pels organitzadors del benchmark, el qual s'utilitza per a generar llibreries neutròniques a distribuir entre els participants del benchmark. El benchmark del UAM pretén determinar la incertesa introduïda pels codis multifísics i multiescala acoblats d'anàlisi de reactors d'aigua lleugera. El citat benchmark es divideix en tres fases: 1. Fase neutrònica: obtenir els paràmetres neutrònics i seccions eficaces del problema específic, col¿lapsats i homogeneïtzats, a més de la anàlisi de criticitat. 2. Fase de nucli: anàlisi termo-hidràulica i neutrònica per separat. 3. Fase de sistema: anàlisi termo-hidràulica i neutrònica acoblats. En aquesta tesi es completen els principals objectius de la primera fase. Concretament, es desenvolupa una metodologia per propagar la incertesa de les seccions eficaces i altres paràmetres neutrònics a través d'un codi lattice i un simulador de nucli. Es porta a terme una anàlisi d'incertesa i sensibilitat per a les seccions eficaces contingudes en la llibreria neutrònica ENDF/B-VII. La seua incertesa es propaga a través del codi lattice SCALE6.2.1, incloent les fases per col¿lapsar i homogeneïtzar, fins aplegar a la generació d'una llibreria neutrònica específica del problema. Després, la incertesa continguda en la esmentada llibreria pot continuar propagant-se a través d'un simulador de nucli, per a aquest estudi PARCSv3.2. Per a l'anàlisi d'incertesa i sensibilitat s'ha utilitzat el mòdul SAMPLER -disponible a l'última versió de SCALE- i la ferramenta estadística DAKOTA 6.3. Com a part d'aquest procés, també es desenvolupa una metodologia per a obtenir llibreries neutròniques en format NEMTAB per ser utilitzades en simuladors de nucli. S'ha realitzat una comparació amb el codi CASMO-4 per obtenir una verificació de la metodologia completa. Aquesta s'ha provat utilitzant un reactor d'aigua en ebullició del tipus BWR. Tanmateix, no hi ha cap preocupació o limitació respecte del seu ús amb un altre tipus de reactor nuclear. Per a la quantificació de la incertesa s'utilitza la metodologia estocàstica Gesellschaft für Anlagen und Reaktorsicherheit (GRS). Aquesta metodologia fa ús del model d'alta fidelitat i un mostreig no paramètric per propagar la incertesa. Com a resultat, el nombre de mostres (determinat amb la fórmula revisada de Wilks) no depèn del nombre de paràmetres d'entrada, sols depèn del nivell de confiança i incertesa desitjats dels paràmetres d'eixida. A més, las funcions de distribució de probabilitat no estan limitades a la normalitat. El principal inconvenient és que s'ha de disposar de les distribucions de probabilitat de cada paràmetre d'entrada. Si és possible, les distribucions de probabilitat d'entrada es defineixen utilitzant informació trobada a la literatura relacionada. En cas contrari, la incertesa es defineix en base a l'opinió d'un expert. S'utilitza un segon escenari per propagar la incertesa de diferents paràmetres termo-hidràulics a través del codi acoblat TRACE5.0p3/PARCSv3.0. En aquest cas, s'utilitza un reactor tipus PWR per simular un transitori d'una caiguda de barra. Com a nova característica, cal assenyalar que el nucli es modela element a element seguint una discretizació totalment 3D. No s'ha trobat cap altre estudi que utilitze un nucli tan detallat en 3D. També s'utilitza la metodologia GRS i el DAKOTA 6.3 per a aquesta anàlisi d'incertesa i sensibilitat.¿ / Mesado Melia, C. (2017). Uncertainty Quantification and Sensitivity Analysis for Cross Sections and Thermohydraulic Parameters in Lattice and Core Physics Codes. Methodology for Cross Section Library Generation and Application to PWR and BWR [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/86167 / TESIS

Es wurde ein Modell zur realistischen Beschreibung der Kühlmittelvermischung innerhalb des Reaktordruckbehälters von Druckwasserreaktoren in den gekoppelten Programmkomplex DYN3D/ATHLET implementiert. Diese Modell basiert auf dem Prinzip der linearen Superposition Dirac-Impuls-ähnlicher Störungen und kann für die Modellierung der Vermischung von Kühlmittel unterschiedlicher Temperatur und/oder unterschiedlicher Borsäurekonzentration eingesetzt werden. Der um das Vermischungsmodell erweiterte Programmkomplex DYN3D/ATHLET wurde für Analyse von Borverdünnungsstörfällen und Frischdampflecks angewandt. Für den Fall "Start der ersten Hauptkühlmittelpumpe bei Vorhandensein eines minderborierten Pfropfens im kalten Strang" zeigten die Ergebnisse der durchgeführten Parameterstudie, dass es selbst bei Annahme des maximal möglichen Pfropfenvolumens nicht zu einer Schädigung des Brennstoffes kommt. Mit den Analysen zu einem generischen Frischdampfleckszenario wurde die Anwendbarkeit des Programmkomplexes DYN3D/ATHLET auf die zweite Störfallklasse, in der die Kühlmittelvermischung eine wichtige Rollen spielt, demonstriert. Im Rahmen der Arbeiten zum Projekt wurde außerdem gezeigt, dass der Einfluss der turbulenten Schwankungen des Geschwindigkeitsfeldes innerhalb des Reaktordruckbehälters auf neutronenkinetische Parameter im Nominalbetrieb und unter Störfallbedingungen nicht zu vernachlässigen ist. A model for the realistic description of the coolant mixing inside the pressure vessel of pressurized water reactors was implemented into the coupled code complex DYN3D/ATHLET. This model is based on the linear superposition Dirac-pulse-like perturbations. The model can be applied to the mixing of coolant of different temperature and/or boron concentration. The coupled code complex DYN3D/ATHLET with the newly implemented model was applied to the analysis of boron dilution and steam line break accidents. The results of a parameter study for the case "Start-up of the first main coolant pump with a slug of lower borated water in the cold leg" have shown, that even under the conditions of the maximum slug volume there is no fuel damage. The applicability to the second class of accidents, where the coolant mixing has to be considered, was demonstrated by the analysis of a generic main steam line break scenario. Further it was shown, that the influence of turbulent fluctuations of the velocity inside the reactor pressure vessel during nominal and accident conditions on neutron-kinetic parameters cannot be neglected.

Ce travail de thèse porte sur l’étude des pertes de pression pour des écoulements monophasiques et diphasiques inertiels au travers de milieux poreux. Son objectif est d’aider à la compréhension et à la modélisation des transferts de quantité de mouvement à l’intérieur de lits de particules, en lien avec la problématique de la gestion d’un accident grave dans un réacteur nucléaire. En effet, lors d’un tel accident, la dégradation du coeur du réacteur peut amener celui-ci à s’effondrer pour former un lit de débris, que l’on peut assimiler à un milieu poreux à haute température et dégageant de la chaleur. Ce travail de thèse s’inscrit dans un projet de recherche en sûreté nucléaire visant à prédire la refroidissabilité d’un lit de débris par injection d’eau, ou « renoyage ». Une étude expérimentale des pertes de pression pour des écoulements monodimensionnels monophasiques et diphasiques à froid est proposée dans des situations représentatives du cas réacteur, en termes de granulométrie, de formes de particules et de vitesses d’écoulement. Les expériences réalisées apportent un complément important aux données existantes, en permettant notamment d’explorer les domaines d’écoulements diphasiques avec nombres de Reynolds liquides non nuls, tout en mesurant le taux de vide, ce qui est essentiel pour une modélisation. Des modèles prédictifs pour les pertes de pression à l’intérieur d’écoulements monophasiques et diphasiques au travers de lits de particules sont établis à partir des structures d’équations obtenues par une prise de moyenne volumique des équations de conservation locales. L’observation des écoulements monophasiques montrent que des lois de type Darcy-Forchheimer avec une correction quadratique en vitesse de filtration sont à même de prédire les pertes de pression avec une précision supérieure à 10%. Une étude numérique a montré que ce résultat est applicable pour un lit désordonné de particules peu rugueuses. L’étude des écoulements diphasiques montre qu’une structure d’équations de type Darcy-Forchheimer généralisée, incluant des termes supplémentaires pour prendre en compte les effets inertiels et les frottements interfaciaux, permet de reproduire le comportement des pertes de pression dans cette situation. Un nouveau modèle est proposé, et comparé aux données expérimentales et aux modèles utilisés dans les codes de simulation des accidents graves. / This work deals with single and two-phase flow pressure losses in porous media. The aim is to improve understanding and modeling of momentum transfer inside particle beds, in relation with nuclear safety issues concerning the reflooding of debris beds during severe nuclear accidents. Indeed, the degradation of the core during such accidents can lead to the collapse of the fuel assemblies, and to the formation of a debris bed, which can be described as a hot porous medium. This thesis is included in a nuclear safety research project on coolability of debris beds during reflooding sequences. An experimental study of single and two-phase cold-flow pressure losses in particle beds is proposed. The geometrical characteristics of the debris and the hydrodynamic conditions are representative of the real case, in terms of granulometry, particle shapes, and flow velocities. The new data constitute an important contribution. In particular, they contain pressure losses and void fraction measurements in two-phase air-water flows with non-zero liquid Reynolds numbers, which did not exist before. Predictive models for pressure losses in single and two-phase flow through particle beds have been established from experimental data. Their structures are based on macroscopic equations obtained from the volume averaging of local conservation equations. Single-phase flow pressure losses can be described by a Darcy-Forchheimer law with a quadratic correction, in terms of filtration velocity, with a better-than-10 % precision. Numerical study of single-phase flows through porous media shows that this correlation is valid for disordered smooth particle beds. Twophase flow pressure losses are described using a generalized Darcy-Forchheimer structure, involving inertial and cross flow terms. A new model is proposed and compared to the experimental data and to the usual models used in severe accident simulation codes.

L’un des éléments clé pour atteindre de hauts rendements dans les centrales solaires à récepteur volumétrique est l’absorbeur. Sa structure poreuse permet au rayonnement solaire de pénétrer à l’intérieur, où l'énergie absorbée doit être transférée efficacement par convection au fluide qui la traverse. Dans ce travail, deux types d’absorbeurs innovants sont présentés et analysés : l’un constitué par une série d’éléments empilés avec une structure en forme de grille, l’autre par des éléments similaires à des stores vénitiens. A titre de référence, un absorbeur classique à nid d’abeilles est aussi évalué. La propagation du rayonnement solaire au sein des absorbeurs est modélisée au moyen de la technique de lancer de rayons, basée sur la méthode de Monte Carlo. Leur comportement thermo-hydraulique est simulé par la méthode des éléments finis. Les caractéristiques géométriques des deux absorbeurs proposés améliorent le transfert thermique par convection par rapport aux absorbeurs alvéolaires et la modification des leurs principaux paramètres géométriques nous a permis d’augmenter la longueur d’extinction du rayonnement solaire. Cependant, l’accroissement de leur surface frontale apparente augmente les pertes par réflexion. A l’issu des résultats théoriques, l’absorbeur à stores vénitiens a été retenu pour l’analyser expérimentalement. Ses performances thermiques sont comparées avec celles d’un absorbeur alvéolaire. Ces résultats montrent également que la structure de l’absorbeur proposé intensifie les échanges thermiques vers le fluide. De plus, ce type d’absorbeur atteint un meilleur comportement thermique à de hauts flux radiatifs et à des débits élevés. / One of the key elements for achieving a high efficiency in solar power plants with volumetric receiver is the absorber. Its porous structure allows the solar radiation to penetrate inside it, where the absorbed energy should be transferred efficiently by convection to the fluid which crosses through it. In this work, two types of innovative absorbers are presented and analyzed: one consisting of a set of stacked elements with a grid-like structure, the other with elements similar to venetian blinds. As a reference, a conventional honeycomb absorber is also evaluated. The solar radiation propagation within the absorbers is modeled through the ray tracing technique, based on the Monte Carlo method. Their thermohydraulic behavior is simulated by the finite element method. The geometrical characteristics of the two proposed absorbers improve convective heat transfer compared to honeycomb absorbers and the modification of their main geometric parameters allowed us to increase the extinction length of solar radiation. However, the increase of their apparent frontal surface rises up reflection losses. Based on the theoretical results, the venetian blind absorber was selected to analyze it experimentally. Its thermal performance is compared with that of a honeycomb absorber. These results also indicate that the structure of the proposed absorber intensifies the heat exchange to the fluid. Moreover, this kind of absorber reaches a better thermal behavior at high heat flux and at high flow rates.