Le sujet de la thèse s'inscrit dans le cadre de la rénovation des outils et des méthodes de calculs appliqués aux accidents graves des Réacteurs à Neutrons Rapides refroidis au Sodium (RNR-Na). En particulier, on s'intéresse aux biais et conservatismes liés à la méthodologie de calculs de la phase primaire d'un accident grave. Pour évaluer les conséquences d'un accident de fusion du coeur d'un RNR-Na, une approche déterministe est généralement réalisée en considérant des hypothèses dites "best-estimate". Cette approche repose sur l'utilisation de codes informatiques pour simuler numériquement le comportement du coeur en conditions accidentelles.La phase primaire de dégradation concerne les évènements se produisant tant que les boîtiers inter-assemblages sont intègres. Les assemblages combustibles conservent alors une indépendance les uns par rapport aux autres. Pour cette raison, la simulation de la phase primaire repose sur une approche multi-canaux. Cette approche consiste à regrouper les assemblages semblables en classes d'assemblages appelés canaux. Le modèle thermo-hydraulique en canaux est couplé à un calcul neutronique pour évaluer le niveau de puissance et de réactivité au cours du transitoire accidentel. La méthodologie de calcul de la phase primaire d'un accident grave repose sur des hypothèses fortes en termes de modélisation neutronique et thermo-hydraulique. Après avoir identifié les principales sources d'erreur, la thèse a consisté à développer un nouvel outil de calcul pour la phase primaire en vue d'évaluer les biais et conservatismes méthodologiques. / This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and recriticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios.During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. Inthe multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level.In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling.

Comprendre les mécanismes de spéciation des Produits de Fission (PF) dans le combustible nucléaire est un enjeu majeur pour pouvoir estimer précisément le terme source d’un accident grave. Parmi les nombreux PF créés, certains sont très réactifs et peuvent avoir un impact radiologique important en cas de relâchement dans l’atmosphère. C’est notamment le cas du césium (Cs), du molybdène (Mo) et du baryum (Ba). C’est dans ce contexte que s’inscrit le travail de thèse qui propose d’apporter des données expérimentales sur l’effet du potentiel oxygène sur la spéciation du Cs, du Mo et du Ba dans des combustibles nucléaires, à différents stades d’un accident grave.Une approche thermodynamique a été utilisée en support à l’interprétation des données expérimentales obtenues dans le cadre de ce travail. Deux types d’échantillons ont été étudiés: des combustibles MOX irradiés et des matériaux simulant un combustible UO2 à fort taux de combustion, obtenus par frittage à haute température (SIMFuel). Les échantillons ont été traités thermiquement dans des conditions représentatives d’un accident grave survenant dans un Réacteur à Eau Pressurisée (REP). Les conditions expérimentales ont couvert une gamme de température allant de 400°C à 2530°C et des potentiels oxygène situés entre -470 kJ.mol(O2)-1 et -100 kJ.mol(O2)-1. Les échantillons ont été caractérisés finement avant et après chaque traitement à l’aide de techniques complémentaires comme la microscopie optique et électronique, la microsonde et le SIMS dans le cas de l’irradié. Des mesures de XANES sur synchrotron ont été réalisées sur SIMFuel et ont conduit à des résultats importants en termes de spéciation des PF. Enfin, la technique de Spark Plasma Sintering (SPS) a été explorée avec succès pour la fabrication de SIMFuel contenant du Cs, du Mo et du Ba sous des formes chimiques représentatives d’un combustible REP en fonctionnement nominal.Ce travail a permis de mettre en évidence l’effet de la température en conditions oxydantes sur le comportement du combustible et des PF. Une oxydation du Mo, initialement présent sous forme métallique dans les inclusions blanches du combustible, en MoO2 a été observée dès 1000°C en conditions oxydantes. Une interaction entre le MoO2 formé et le Ba contenu dans la phase oxyde a eu lieu dans les mêmes conditions, menant à la formation de BaMoO4. Le potentiel oxygène joue aussi un rôle important dans le phénomène d’interaction pastille-gaine, en favorisant la diffusion des espèces en conditions oxydantes, diminuant ainsi la température de fusion du combustible. / In the nuclear community, it is a top priority to gain in-depth understanding of fission product (FP) speciation mechanisms occurring in nuclear fuel in order to precisely estimate the source term of a severe accident. Among the FP produced, some are highly reactive and may have a strong radiological impact if released into the environment. This is particularly the case of cesium (Cs), molybdenum (Mo) and barium (Ba). In this context, the objective of this study is to provide experimental data on the effect of the oxygen potential on Cs, Mo and Ba speciation in nuclear fuels at different stages of a severe accident.A thermodynamic approach was coupled with the experimental work to support the interpretation of experimental data. Two types of samples were studied in detail: irradiated MOX fuels and simulated high burn-up UO2 fuels produced through sintering at high temperature (SIMFuel). The samples were submitted to thermal treatments in conditions representative of a pressurised water reactor (PWR) severe accident. This approach made it possible to cover a temperature range from 400°C up to 2530°C and oxygen potentials from -470 kJ.mol(O2)-1 to -100 kJ.mol(O2)-1. The samples were characterized before and after each test using complementary techniques like OM, SEM, EPMA and SIMS in the case of irradiated fuels. XANES measurements using synchrotron radiation facilities were performed on SIMFuels and provided valuable results on FP speciation. Moreover, spark plasma sintering (SPS) was successfully investigated for the production of SIMFuel samples containing Cs, Mo and Ba in a chemical state representative of PWR fuel in normal operating conditions.This work highlighted the effect of oxidizing severe accident conditions on the fuel and FP behavior. Oxidation of Mo initially contained in the fuel’s metallic inclusions into MoO2 was observed to take place around 1000°C in oxidizing conditions. An interaction between MoO2 and the oxide phase containing Ba took place in the same conditions, leading to the formation of BaMoO4. The oxygen potential also plays an important role in fuel-cladding interactions, enhancing the diffusion of species in oxidizing conditions and lowering the temperature at which fuel melting occurs.

<p>Severe accidents in a Light Water Reactor (LWR) have been a subject of the research for the last three decades. The research in this area aims to further understanding of the inherent physical phenomena and reduce the uncertainties surrounding their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors. The research is also focusing on evaluation of the proposed accident management schemes for mitigating the consequences of such accidents.</p><p>During a hypothetical severe accident, whatever the scenario, there is likelihood that the core material will be relocated and accumulated in the lower plenum in the form of a debris bed or a melt pool. Physical phenomena involved in a severe accident progression are complex. The interactions of core debris or melt with the reactor structures depend very much on the debris bed or melt pool thermal hydraulics. That is why predictions of heat transfer during melt pool formation in the reactor lower head are important for the safety assessment.</p><p>The main purpose of the present study is to advance a method for describing turbulent natural convection heat transfer of a melt pool, and to develop a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of Core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident.</p><p>Given the insights gained from Computational Fluid Dynamics (CFD) simulations, a physics-based model and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a Boiling Water Reactor (BWR) during the late phase of an in-vessel core melt progression. A model is developed for the core debris bed heat up and formation of a melt pool in the lower head of the reactor vessel, and implemented in a commercial CFD code. To describe the natural convection heat transfer inside the volumetrically decay-heated melt pool, we advanced the Effective Convectivity Conductivity Model (ECCM), which was previously developed and implemented in the MVITA code. In the present study, natural convection heat transfer is accounted for by only the Effective Convectivity Model (ECM). The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables simulations of heat transfer of a high Rayleigh melt pool in 3D large dimension geometry.</p><p>In order to describe the phase-change heat transfer associated with core debris, a temperature-based enthalpy formulation is employed in the ECM (the phase-change ECM or so called the PECM). The PECM is capable to represent possible convection heat transfer in a mushy zone. The simple approach of the PECM method allows implementing different models of the fluid velocity in a mushy zone for a non-eutectic mixture. The developed models are validated by a dual approach, i.e., against the existing experimental data and the CFD simulation results.</p><p>The ECM and PECM methods are applied to predict thermal loads to the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heat up and melting in the BWR lower plenum. Applying the ECM and PECM to simulations of reactor-scale melt pool heat transfer, the results of the ECM and PECM calculations show an apparent effectiveness of the developed methods that enables simulations of long term accident transients. It is also found that during severe accident progression, the cooling by water flowing inside the CRGTs plays a very important role in reducing the thermal load on the reactor vessel wall. The results of the CFD, ECM and PECM simulations suggest a potential of the CRGT cooling as an effective mitigative measure during a severe accident progression.</p>

Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents.  In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment.  The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis.   The CFD method, on the one hand, is indispensable for scrutinizing flow physics, on the other hand, the validated CFD method can be used to generate necessary data for validation of the accident analysis models. Given the insights gained from the CFD study, physics-based models and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a LWR during the late phase of an in-vessel core melt progression. To describe natural convection heat transfer in an internally heated volume, and molten metal layer heated from below and cooled from the top (and side) walls, the Effective Convectivity Models (ECM) are developed and implemented in a commercial CFD code. The ECM uses directional heat transfer characteristic velocities to transport the heat to cooled boundaries. The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables 3D heat transfer simulations of a homogeneous (and stratified) melt pool formed in the LWR lower head. In order to describe phase-change heat transfer associated with core debris or binary mixture (e.g. in a molten metal layer), a temperature-based enthalpy formulation is employed in the Phase-change ECM (so called the PECM). The PECM is capable to represent natural convection heat transfer in a mushy zone. Simple formulation of the PECM method allows implementing different models of mushy zone heat transfer for non-eutectic mixtures. For a non-eutectic binary mixture, compositional convection associated with concentration gradients can be taken into account. The developed models are validated against both existing experimental data and the CFD-generated data. ECM and PECM simulations show a superior computational efficiency compared to the CFD simulation method. The ECM and PECM methods are applied to predict thermal loads imposed on the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in a Boiling Water Reactor (BWR) lower plenum. It is found that during the accident progression, the CRGT cooling plays a very important role in reducing the thermal loads on the reactor vessel wall. Results of the ECM and PECM simulations suggest a high potential of the CRGT cooling to be an effective measure for severe accident management in BWRs. / <p>QC 20100812</p>

This master's thesis is an investigation and evaluation of MELCOR (a software tool for severe accident analyses regarding nuclear power plants), or more correctly of the (ASEA-Atom BWR 75) reactor model developed for version 1.8.6 of MELCOR. The main objective was to determine if MELCOR, with the reactor model in question, is able to produce satisfactory results in severe accident analyses compared to results made by MAAP, which is currently the only official software tool for this application in Sweden. The thesis work is related to the RASTEP project. This project has been carried out in several stages on behalf of SSM since 2009, with a number of specific issues explored within an NKS funded R&amp;D project carried out 2011-2013. This investigation is related to the NKS part of the project. The purpose with the RASTEP project is to develop a method for rapid source term prediction that could aid the authorities in decision making during a severe accident in a nuclear power plant. A software tool, which also gave the project its name, i.e. RASTEP (RApid Source TErm Prediction), is therefore currently under development at Lloyd's Register Consulting. A software tool for severe accident analyses is needed to calculate the source terms which are the end result from the predictions made by RASTEP. A set of issues have been outlined in an earlier comparison between MAAP and MELCOR. The first objective was therefore to resolve these pre-discovered issues, but also to address new issues, should they occur. The existing MELCOR reactor model also had to be further developed through the inclusion of various safety systems, since these systems are required for certain types of scenarios. Subsequently, a set of scenarios was simulated to draw conclusions from the additions made to the reactor model. Most of the issues (pre-discovered as well as new ones) could be resolved. However the work also rendered a set of issues which are in need of further attention and investigation. The overall conclusion is that MELCOR is indeed a promising alternative for severe accident analyses in the Swedish work with nuclear safety. Several potential benefits from making use of MELCOR besides MAAP have been identified. In conclusion, they would be valuable assets to each other, e.g. since deviations in the results (between the two codes) would highlight possible weaknesses of the simulations. Finally it is recommended that the work on improving the MELCOR reactor model should continue. / RASTEP

Ce travail a été réalisé en cotutelle entre l’Université Charles à Prague (République Tchèque) et l'Université de Strasbourg (France). Il a également profité d’une coopération entre l'Institut de Chimie Inorganique de l'Académie des Sciences de République Tchèque et le Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, Cadarache, France). Les résultats des travaux ont contribué au projet OCDE / AEN Serena 2 (Programme portant sur l’étude des effets d'une explosion de vapeur dans un réacteur nucléaire à eau). La thèse présentée se situe dans le domaine de la sûreté nucléaire et de la science des matériaux. Elle traite de l’Interaction Combustible-Réfrigérant (ICR, ou FCI en anglais pour Fuel-Coolant Interaction) susceptible d’intervenir lors d’un accident grave de réacteur nucléaire et actuellement à l’étude dans les programme de R&D. Au cours d’un accident de fusion d’un coeur de réacteur, les matériaux fondus peuvent interagir avec le liquide de refroidissement (eau légère), aussi appelé réfrigérant. Cette interaction peut se produire à l'intérieur de la cuve ou, en cas de rupture de celle-ci, à l'extérieur. Ces deux scénarios sont couramment appelés Interaction Combustible-Réfrigérant en- et hors- cuve et se distinguent de par les conditions du réacteur lors de l’accident : pression du système, degré de sous refroidissement de l’eau, etc. L'interaction entre le combustible fondu et le liquide de refroidissement peut évoluer vers une détonation thermique appelée «explosion de vapeur» qui peut endommager le réacteur, voire compromettre l'intégrité du confinement. Des expériences récentes ont montré que la composition du combustible a un effet majeur sur l’apparition et le rendement d’une telle explosion. En particulier, des comportements différents ont été observés entre un matériau simulant, l'alumine, qui explose très facilement, et diverses compositions de corium prototypique (80 m. % UO2, 20% m.% ZrO2). Cet «effet matériau» a suscité un intérêt nouveau pour les analyses post-expériences des débris issus de l’ICR afin de déterminer les mécanismes qui interviennent au cours de ces phénomènes extrêmement rapides. La thèse est organisée en neuf chapitres. Le chapitre 1 constitue une introduction générale et présente le contexte d’un accident grave d’un réacteur nucléaire. Quelques exemples d’accidents graves (Three Miles Island 1979, Tchernobyl 1986 et Fukushima 2011) sont brièvement abordés. Le chapitre 2 résume les aspects théoriques de l'interaction combustible-réfrigérant. Il est divisé en quatre parties correspondant aux quatre étapes généralement rencontrées lors du mécanisme d’ICR i) Prémélange - le combustible fondu, versé dans l'eau, se fragmente en gouttelettes grossières qui s’isolent d’un film de vapeur. ii) Déclenchement – le film de vapeur entourant les gouttes de combustible est déstabilisé, permettant ainsi la fragmentation fine du combustible. iii) Propagation - la fragmentation du combustible se propage à l’ensemble du prémélange, augmentant ainsi la surface de contact entre le combustible fondu et l’eau. Ceci conduit à une production intense de vapeur à grande échelle. iv) Expansion (explosion) - l'énergie thermique transférée du combustible à l'eau est transformée en travail mécanique de la vapeur.[...] / This work has been performed under co-tutelle supervision between Charles University in Prague (Czech Republic) and Strasbourg University (France). It also profited from the background and cooperation of Institute of Inorganic Chemistry Academy of Science of the Czech Republic and French Commission for Atomic and Alternative energies (CEA Cadarache). Results of the work contribute to the OECD/NEA project Serena 2 (Program on Steam Explosion Resolution for Nuclear Applications).Presented thesis can be classed in the scientific field of nuclear safety and material science. It is aimed on the socalled “molten nuclear Fuel – Coolant Interaction” (FCI) that belongs among the recent issues of the nuclear reactorsevere accident R&D. During the nuclear reactor melt down accident the melted reactor load can interact with the coolant (light water). This interaction can be located inside the vessel or outside in the case of vessel break-up. These two scenarios are commonly called in- and ex-vessel FCI and they differ in the conditions such as initial pressure of the system, water sub-cooling etc. The Molten fuel – coolant interaction can progress into thermal detonation called “steam explosion” that can challenge the reactor or containment integrity.Recent experiments have shown that the melt composition has a major effect on the occurrence and yield of such explosion. In particular, different behaviors have been observed between simulant material (alumina), which has important explosion efficiency, and some prototypic corium compositions (80 w. % UO2, 20% w. % ZrO2). This “material effect” has launched a new interest in the post-test analyses of FCI debris in order to estimate the processes occurring during these extremely rapid phenomena. The thesis is organized in nine chapters. The chapter 1 gives the general introduction and context of the nuclear reactor accident. Major nuclear accidents (Three Miles Island 1979, Chernobyl 1986 and Fukushima 2011) are briefly described. The chapter 2 summarizes the theoretical aspects of the fuel – coolant interaction. It is divided in four thematic fields according to the FCI progression. In general, FCI has four stages: i) Premixing – hot melt is poured in water and fragmented in coarse droplets surrounded by steam filmii) Triggering – steam film around melt droplets is destabilized allowing fine fragmentation iii) Propagation – the fine fragmentation propagate through the premixture increasing the melt – water interface area, which leads to large steam production iv) Expansion (explosion) – Thermal energy transferred from the melt to water is changed into mechanical workof the steam.The chapter 3 summarizes the research conducted in different experimental facilities using nonradioactive simulant or radioactive prototypic materials. The chapter 4 shows the results of thermodynamic calculations, by which thepossible chemici reactions between melts and water/steam at high temperatures were modeled. Second part presentsthe results of 1D calculations of radiation heat transfer from FCI materials to water/steam. The chapter 5 describes the material analyses of non-radioactive simulant debris coming from MISTEE experimental research program (KTH, Sweden) and PREMIX, ECO facilities (FZK, Germany). The chapters 6 to 8 describe the material analyses of radioactive prototypic debris coming from KROTOS research program (CEA, France). The KROTOS KS2 test used melt composition 70 w. % UO2 and 30 w. % ZrO2, the KS4 test 80 w. % UO2 and 20 w. % ZrO2, the last KS5 test used suboxidized melt 80.1 w. % UO2 and 11.4 w. % ZrO2 and 8.5 w. % metallic Zr. The chapter 9 concludes the work and presents future perspectives.

Severe accidents in a Light Water Reactor (LWR) have been a subject of the research for the last three decades. The research in this area aims to further understanding of the inherent physical phenomena and reduce the uncertainties surrounding their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors. The research is also focusing on evaluation of the proposed accident management schemes for mitigating the consequences of such accidents. During a hypothetical severe accident, whatever the scenario, there is likelihood that the core material will be relocated and accumulated in the lower plenum in the form of a debris bed or a melt pool. Physical phenomena involved in a severe accident progression are complex. The interactions of core debris or melt with the reactor structures depend very much on the debris bed or melt pool thermal hydraulics. That is why predictions of heat transfer during melt pool formation in the reactor lower head are important for the safety assessment. The main purpose of the present study is to advance a method for describing turbulent natural convection heat transfer of a melt pool, and to develop a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of Core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. Given the insights gained from Computational Fluid Dynamics (CFD) simulations, a physics-based model and computationally-efficient tools are developed for multi-dimensional simulations of transient thermal-hydraulic phenomena in the lower plenum of a Boiling Water Reactor (BWR) during the late phase of an in-vessel core melt progression. A model is developed for the core debris bed heat up and formation of a melt pool in the lower head of the reactor vessel, and implemented in a commercial CFD code. To describe the natural convection heat transfer inside the volumetrically decay-heated melt pool, we advanced the Effective Convectivity Conductivity Model (ECCM), which was previously developed and implemented in the MVITA code. In the present study, natural convection heat transfer is accounted for by only the Effective Convectivity Model (ECM). The heat transport and interactions are represented through an energy-conservation formulation. The ECM then enables simulations of heat transfer of a high Rayleigh melt pool in 3D large dimension geometry. In order to describe the phase-change heat transfer associated with core debris, a temperature-based enthalpy formulation is employed in the ECM (the phase-change ECM or so called the PECM). The PECM is capable to represent possible convection heat transfer in a mushy zone. The simple approach of the PECM method allows implementing different models of the fluid velocity in a mushy zone for a non-eutectic mixture. The developed models are validated by a dual approach, i.e., against the existing experimental data and the CFD simulation results. The ECM and PECM methods are applied to predict thermal loads to the vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heat up and melting in the BWR lower plenum. Applying the ECM and PECM to simulations of reactor-scale melt pool heat transfer, the results of the ECM and PECM calculations show an apparent effectiveness of the developed methods that enables simulations of long term accident transients. It is also found that during severe accident progression, the cooling by water flowing inside the CRGTs plays a very important role in reducing the thermal load on the reactor vessel wall. The results of the CFD, ECM and PECM simulations suggest a potential of the CRGT cooling as an effective mitigative measure during a severe accident progression. / QC 20101119

Les Etudes Probabilistes de Sûreté (EPS) de niveau 2 en centrale nucléaire visent à identifier les séquences d’événements pouvant correspondre à la propagation d’un accident d’un endommagement du cœur jusqu’à une perte potentielle de l’intégrité de l’enceinte, et à estimer la fréquence d’apparition des différents scénarios possibles. Ces accidents sévères dépendent non seulement de défaillances matérielles ou d’erreurs humaines, mais également de l’occurrence de phénomènes physiques, tels que des explosions vapeur ou hydrogène. La prise en compte de tels phénomènes dans le cadre booléen des arbres d’événements s’avère difficile, et les méthodologies dynamiques de réalisation des EPS sont censées fournir une manière plus cohérente d’intégrer l’évolution du processus physique dans les changements de configuration discrète de la centrale au long d’un transitoire accidentel. Cette thèse décrit l’application d’une des plus récentes approches dynamiques des EPS – la Théorie de la Dynamique Probabiliste basée sur les Stimuli (SDTPD) – à différents modèles de déflagration d'hydrogène ainsi que les développements qui ont permis cette applications et les diverses améliorations et techniques qui ont été mises en oeuvre. Level-2 Probabilistic Safety Analyses (PSA) of nuclear power plants aims to identify the possible sequences of events corresponding to an accident propagation from a core damage to a potential loss of integrity of the containment, and to assess the frequency of occurrence of the different scenarios. These so-called severe accidents depend not only on hardware failures and human errors, but also on the occurrence of physical phenomena such as e.g. steam or hydrogen explosions. Handling these phenomena in the classical Boolean framework of event trees is not convenient, and dynamic methodologies to perform PSA studies are expected to provide a more consistent way of integrating the physical process evolution with the discrete changes of plant configuration along an accidental transient. This PhD Thesis presents the application of one of the most recently proposed dynamic PSA methodologies, i.e. the Stimulus-Driven Theory of Probabilistic Dynamics (SDTPD), to several models of hydrogen explosion in the containment of a plant, as well as the developed methods and improvements.

En centrales nucléaires, un accident grave est une séquence très peu probable d’événements durant laquelle des composants du réacteur sont significativement endommagés, par interactions chimiques et/ou fusion, à cause de très hautes températures. Cela peut mener à des rejets radiotoxiques dans l’enceinte et à une entrée d’air dans le réacteur. Dans ce contexte, ce travail de thèse mené chez EDF R&D visait à modéliser la détérioration du gainage combustible, en alliages de zirconium, en conditions accidentelles : haute température et soit vapeur soit mélange air-vapeur. L’objectif final était d’améliorer la simulation par le code MAAP de l’oxydation du gainage et de la production d’hydrogène, en particulier pendant un renoyage avec de l’eau. Dû à l’épaississement progressif d’une couche de ZrO2 dense et protectrice, la cinétique d’oxydation du Zr en vapeur à hautes températures est généralement (sous-)parabolique. Cependant, à certaines températures, cette couche d’oxyde peut se fissurer, devenant poreuse et non protectrice. Par ce processus de « breakaway », la cinétique d’oxydation devient plus linéaire. De plus, l’augmentation de température peut mener les matériaux du réacteur à fondre et à se relocaliser dans le fond de cuve dont la rupture peut induire une entrée d’air dans le réacteur. Dans ce cas, l’oxygène et l’azote réagissent avec les gaines pré-oxydées, successivement par oxydation du Zr (épaississant la couche de ZrO2), nitruration du Zr (formant des particules de ZrN) et oxydation du ZrN (créant de l’oxyde et relâchant de l’azote). Ces réactions auto-entretenues relancent la fissuration du gainage et de sa couche de ZrO2, induisant une hausse de sa porosité ouverte. Afin de quantifier cette porosité du gainage, un protocole expérimental innovant en deux étapes a été défini et appliqué : il consistait à soumettre des échantillons de gainage en ZIRLO® à diverses conditions accidentelles pendant plusieurs durées puis à des mesures de la porosité ouverte par porosimétrie par intrusion de mercure. Les conditions de corrosion comprenaient plusieurs températures allant de 1100 à 1500 K ainsi que de la vapeur et un mélange air-vapeur 50-50 mol%. Pour les échantillons de ZIRLO® oxydés en vapeur, sauf à 1200 et 1250 K, les transitions de cinétique n’ont pas lieu et la porosité ouverte reste négligeable au cours de l’oxydation. Cependant, pour les autres échantillons, corrodés en air-vapeur ou oxydés en vapeur à 1200 ou 1250 K, des transitions « breakaway » sont observées et les résultats de porosimétrie montrent que la porosité ouverte augmente au cours de la corrosion, proportionnellement au gain en masse. De plus, il a été mis en évidence que la distribution de tailles de pores des échantillons de ZIRLO® s’étend significativement pendant la corrosion, en particulier après « breakaway ». En effet, ces tailles vont de 60 μm à environ : 2 μm avant la transition, 50 nm juste après et 2 nm plus longtemps après. Enfin, un modèle numérique en deux étapes a été développé dans le code MAAP pour améliorer sa simulation de l’oxydation du gainage. D’abord, grâce à la proportionnalité entre porosité ouverte et gain en masse des échantillons, des corrélations de porosité ont été implémentées pour chaque condition de corrosion. Ensuite, les valeurs de porosité calculées sont utilisées pour augmenter proportionnellement la vitesse d’oxydation du gainage. Ce modèle amélioré simule ainsi non seulement les réactions chimiques des gaines en Zr (oxydation et nitruration) mais aussi leur dégradation mécanique et son impact sur leur vitesse d’oxydation. Ceci a été validé en simulant des essais QUENCH (-06, -08, -10 et -16), conduits au KIT pour étudier le comportement de gaines dans des conditions accidentelles avec un renoyage final. Ces simulations montrent un meilleur comportement thermique du gainage et une production d’hydrogène significativement plus haute et donc plus proche des valeurs expérimentales, en particulier pendant le renoyage. / In nuclear power plants, a severe accident is a very unlikely sequence of events during which components of the reactor core get significantly damaged, through chemical interactions and/or melting, because of very high temperatures. This may potentially lead to radiotoxic releases in the containment building and to air ingress in the reactor core. In that context, this thesis work led at EDF R&D aimed at modeling the deterioration of the nuclear fuel cladding, made of zirconium alloys, in accidental conditions: high temperature and either pure steam or air-steam mixture. The final objective was to improve the simulation by the MAAP code of the cladding oxidation and of the hydrogen production, in particular during a core reflooding with water. Due to the progressive thickening of a dense and protective ZrO2 layer, the oxidation kinetics of Zr in steam at high temperatures is generally (sub-)parabolic. However, at certain temperatures, this oxide layer may crack, becoming porous and not protective anymore. By this “breakaway” process, the oxidation kinetics becomes rather linear. Additionally, the temperature increase can lead core materials to melt and to relocate down to the vessel lower head whose failure may induce air ingress into the reactor core. In this event, oxygen and nitrogen both react with the pre-oxidized claddings, successively through oxidation of Zr (thickening the ZrO2 layer), nitriding of Zr (forming ZrN particles) and oxidation of ZrN (creating oxide and releasing nitrogen). These self-sustained reactions enhance the cracking of the cladding and of its ZrO2 layer, inducing a rise of its open porosity.In order to quantify this cladding porosity, an innovative two-step experimental protocol was defined and applied: it consisted in submitting ZIRLO® cladding samples first to various accidental conditions during several time periods and then to measurements of the open porosity through porosimetry by mercury intrusion. The tested corrosion conditions included numerous temperatures ranging from 1100 up to 1500 K as well as both pure steam and a 50-50 mol% air-steam mixture. For the ZIRLO® samples oxidized in pure steam, except at 1200 and 1250 K, the “breakaway” kinetic transitions do not occur and the open porosity remains negligible along the oxidation process. However, for all other samples, corroded in air-steam or oxidized in pure steam at 1200 or 1250 K, “breakaway” transitions are observed and the porosimetry results show that the open porosity increases along the corrosion process, proportionally to the mass gain. Moreover, it was evidenced that the pore size distribution of ZIRLO® samples significantly extends during corrosion, especially after “breakaway” transitions. Indeed, the detected pore sizes ranged from 60 μm down to around: 2 μm before the transition, 50 nm just after and 2 nm longer after. Finally, a two-step numerical model was developed in the MAAP code to improve its simulation of the cladding oxidation. First, thanks to the proportionality between open porosity and mass gain of cladding samples, porosity correlations were implemented for each tested corrosion condition. Second, the calculated porosity values are used to proportionally enhance the cladding oxidation rate. This improved model thus simulates not only chemical reactions of Zr-based claddings (oxidation and nitriding) but also their mechanical degradation and its impact on their oxidation rate. It was validated by simulating QUENCH tests (-06, -08, -10 and -16), conducted at KIT to study the behavior of claddings in accidental conditions with a final reflooding. These simulations show a better cladding thermal behavior and a hydrogen production significantly higher and so closer to experimental values, in particular during the reflooding.

Les Etudes Probabilistes de Sûreté (EPS) de niveau 2 en centrale nucléaire visent à identifier les séquences d’événements pouvant correspondre à la propagation d’un accident d’un endommagement du cœur jusqu’à une perte potentielle de l’intégrité de l’enceinte, et à estimer la fréquence d’apparition des différents scénarios possibles.<p>Ces accidents sévères dépendent non seulement de défaillances matérielles ou d’erreurs humaines, mais également de l’occurrence de phénomènes physiques, tels que des explosions vapeur ou hydrogène. La prise en compte de tels phénomènes dans le cadre booléen des arbres d’événements s’avère difficile, et les méthodologies dynamiques de réalisation des EPS sont censées fournir une manière plus cohérente d’intégrer l’évolution du processus physique dans les changements de configuration discrète de la centrale au long d’un transitoire accidentel.<p>Cette thèse décrit l’application d’une des plus récentes approches dynamiques des EPS – la Théorie de la Dynamique Probabiliste basée sur les Stimuli (SDTPD) – à différents modèles de déflagration d'hydrogène ainsi que les développements qui ont permis cette applications et les diverses améliorations et techniques qui ont été mises en oeuvre.<p><p>Level-2 Probabilistic Safety Analyses (PSA) of nuclear power plants aims to identify the possible sequences of events corresponding to an accident propagation from a core damage to a potential loss of integrity of the containment, and to assess the frequency of occurrence of the different scenarios.<p>These so-called severe accidents depend not only on hardware failures and human errors, but also on the occurrence of physical phenomena such as e.g. steam or hydrogen explosions. Handling these phenomena in the classical Boolean framework of event trees is not convenient, and dynamic methodologies to perform PSA studies are expected to provide a more consistent way of integrating the physical process evolution with the discrete changes of plant configuration along an accidental transient.<p>This PhD Thesis presents the application of one of the most recently proposed dynamic PSA methodologies, i.e. the Stimulus-Driven Theory of Probabilistic Dynamics (SDTPD), to several models of hydrogen explosion in the containment of a plant, as well as the developed methods and improvements.<p> / Doctorat en Sciences de l'ingénieur / info:eu-repo/semantics/nonPublished