Submitted by Marco Antonio Oliveira da Silva (maosilva@ipen.br) on 2017-11-09T11:20:17Z No. of bitstreams: 0 / Made available in DSpace on 2017-11-09T11:20:17Z (GMT). No. of bitstreams: 0 / Em 1958 a Comissão Internacional de Proteção Radiológica (CIPR) propôs a primeira filosofia do sistema de limitação de dose, introduzindo os Limites Anuais Máximos Permissíveis (LAMP). O grande avanço da era nuclear nas últimas décadas impôs novos paradigmas e a necessidade de atualização da filosofia em questão. O presente trabalho tem por objetivo apresentar uma análise da evolução da filosofia do sistema de limitação de dose, desde a década de 50 até os dias atuais. A primeira mudança de paradigma se deu com a criação dos Limites Anuais Máximos Admissíveis (LAMA), ainda vigentes. Por meio de um cuidadoso estudo das publicações do Organismo Internacional de Energia Atômica (OIEA) e das recomendações da CIPR, foi possível evidenciar e detalhar o processo de evolução dos LAMA ao longo das últimas décadas. A pesquisa aborda momentos-chaves que impulsionaram mudanças na filosofia do sistema de limitações de dose como, por exemplo, a crise internacional do petróleo e suas implicações no desenvolvimento da área nuclear. A comparação entre as diversas publicações das duas entidades (OIEA e CIPR) permitiu um estudo aprofundado desde o surgimento dessas filosofias até suas últimas publicações. Os resultados deste estudo apontam importantes informações que constam em publicações da CIPR, hoje consideradas \"superseded\", que não são encontradas nas publicações atuais. O OIEA, que elabora suas recomendações baseado na filosofia da CIPR, também não aborda as referidas informações. Por meio da presente pesquisa, foi possível evidenciar e detalhar valiosas informações que se perderam durante o processo de atualização das publicações e edição de recomendações de ambas as entidades. Este trabalho se propõe a apresentar essas informações, que foram estudadas em profundidade, discutindo seu real valor, propondo à comunidade internacional novas reflexões sobre a importância e a possibilidade de reintroduzir as informações perdidas em futuras publicações. / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energéticas e Nucleares - IPEN-CNEN/SP

Les accidents graves résultant de la fusion d’un coeur de réacteur nucléaire doivent être anticipés pour améliorer l’efficacité de leur mitigation. De tels accidents sont survenus à TMI-2 (1979) et à Fukushima (2011). Suite à un accident de perte de refroidissement, l’échauffement du coeur et l’oxydation de la gaine de combustible suivie d’un renoyage (injection d’eau) peuvent entraîner l’effondrement des barres de combustible et la formation d’un lit de débris dans le coeur. La vapeur produite lors du renoyage peut activer l’oxydation exothermique du Zircaloy, entraînant la fusion partielle des matériaux. Cette évolution engendre des zones à porosité réduite limitant la pénétration de l’eau et/ou des zones imperméables. Dans cette situation, l’efficacité de l’injection d’eau dans le coeur pour arrêter la progression de la dégradation et empêcher la fusion du coeur du réacteur peut être considérablement réduite. Dans ce cadre, l’IRSN a lancé le programme PEARL visant à étudier la thermohydraulique du renoyage des lits de débris chauds entourés d’une zone plus perméable simulant la présence de zones intactes ou moins endommagées dans le coeur. Dans cette thèse, les expériences PEARL ont été modélisées et simulées avec ICARE/CATHARE pour évaluer l’évolution d’un renoyage d’un lit de débris surchauffé entouré d’un bypass de perméabilité plus grande. La thermohydraulique du processus a été analysée et l’effet de différents paramètres (géométrie, conditions aux limites) sur le comportement de renoyage a été évalué. Sous certaines conditions, l’entraînement de l’eau dans le bypass a été identifié et évalué. Un modèle analytique a été mis au point ensuite pour étudier de façon approfondie le renoyage d’un milieu poreux hétérogène surchauffé composé de deux lits de débris de perméabilité et de porosité différentes et pour décrire l’entraînement de l’eau dans le bypass. Ce modèle calcule les principales variables caractérisant le processus de renoyage, telles que la vitesse du front de trempe, le taux de conversion eau-vapeur et le débit d’eau entraîné dans le bypass.Il fournit de bons résultats qualitatifs et quantitatifs concernant la redistribution du débit d’eau par rapport aux résultats expérimentaux. Ce modèle a plusieurs avantages. Il est écrit sous une forme plutôt générale incluant les termes de correction de Forchheimer et les termes croisés non nuls dans l’équation de Darcy-Forchheimer généralisée. Les différentes options des équations de quantité de mouvement proposées, y compris les changements dans les corrélations et les lois de frottement interfacial, peuvent être testées facilement. La comparaison des calculs avec les résultats expérimentaux indique qu’il est nécessaire d’inclure une loi de frottement interfacial pour obtenir de bonnes prédictions. L’extrapolation à l’échelle du réacteur est simple et des calculs ont été effectués pour évaluer l’impact des paramètres géométriques du lit de débris (granulométrie, porosité, dimensions) ainsi que les conditions thermiques et hydrauliques (température, pression, débit d’injection). Ainsi, le modèle est très utile pour estimer le temps de trempe total et latempérature maximale qui pourraient être atteinte dans le lit de débris à grande échelle. Cela permet d’évaluer la probabilité de réussite du renoyage d’un lit de débris chauds formé lors d’un scénario accidentel hypothétique. / Severe accidents arising from the fusion of a nuclear reactor core must be anticipated to enhance the efficiency of their mitigation. Such accidents have occurred at TMI-2 (1979) and Fukushima (2011). Following a loss of coolant accident, core heating and oxidation of the fuel cladding followed by reflooding (injection of water) may lead to the collapse of fuel rods and formation of porous debris bed in the core. Steam produced upon reflooding may activate the exothermic oxidation of Zircaloy leading to partial melting of materials. Such evolution generates zones with reduced porosity limiting coolant penetration and/or impermeable blocked zones. In this situation, the efficiency of injecting water into the core to stop the progress of degradation and prevent the reactor core melting may be significantly reduced. In this scope, IRSN launched PEARL program to investigate the thermal hydraulics of reflooding of hot debris beds surrounded by a more permeable zone simulating the presence of intact or less damaged zones in the core. The PEARL experiments were modeled and simulated using ICARE/CATHARE code to assess the evolution of a bottom reflooding of a superheated debris bed surrounded by a bypass of larger permeability. The thermal hydraulics of the quenching process has been analyzed and the effect of each of the initial conditions on the reflooding behavior was assessed. The effect of pressure was investigated and related to the entrainment of injected water at quench front level into the bypass. An analytical model was then developed to investigate thoroughly the reflooding of a superheated heterogeneous porous medium, composed of two layers of contrasting permeability and porosity, and to describe the water entrainment in the bypass. This model computes the main variables characterizing the reflooding process such as quench front velocity, water-to-steam conversion ratio, and the flow rate of water entrained in the bypass. It provides good qualitative and quantitative results for the two-phase flow redistribution as compared to experimental results. This model has several advantages. It is written in a rather general form including the Forchheimer correction terms and non-zero cross-terms in the generalized Darcy-Forchheimer momentum equation. Variations of proposed momentum equations including changes in correlations andinterfacial friction laws can be tested easily and efficiently. Comparison of the calculations against experimental results indicated that it is necessary to include an interfacial friction law to obtain good predictions. This model allows performing fast evaluations of the efficiency of cooling bycomputing the fraction of the injected flow rate that participates in cooling. Upscaling to the reactor scale is straightforward and calculations were performed to assess the impact of geometric parameters of the debris bed (particle size, porosity, dimensions) as well as thermal hydraulic conditions (temperature, pressure, injection flow rate) on the reflooding process. Thus the model is very useful to estimate the total quenching time and the maximum temperature that could be reached by the hot debris bed at large scales. This allows assessing the probability of a successful quenching of a hot debris bed formed during a hypothetical accidental scenario.

[EN] In case of a severe accident in a NPP fission products are released from the degraded fuel and may reach the environment if their confinement is lost and/or bypassed. Given the high radio-toxic nature of nuclear aerosols for environment and population, their unrestricted release should be absolutely avoided. One particular situation is the core meltdown sequence with steam generator tube rupture (SGTR). The containment bypass turns this sequence into an indispensable scenario to model when assessing PWR risk. As a result, a significant database on the aerosol behavior in the secondary side of the steam generator (SG) has been developed within the international projects EU-SGTR, ARTIST and ARTIST-2. The role played by the break stage is particularly significant since it might be responsible for a good fraction of the total mass retained and for the shift of the particle size distribution towards smaller diameters. This awoke the interest in the effect of variables such as the particle nature, the breach type (size and shape) and the tubes vibration on the particle retention within the breach stage of a dry steam generator. Those aspects have been experimentally investigated in the first part of this thesis. Two experimental campaigns, CAAT2 and SET, were conducted in order to explore the potential influence of the particle nature on their retention. Moreover, the effect of the breach size and shape has been investigated in the CAAT2 campaign while the SET experiments were devoted to the tube vibration characterization and the effect of the vibration on the particle retention. The tests conducted highlighted several key insights: the strong effect of particle nature in the secondary side capability to scrub the particle-laden gas; the confirmation of the high retention efficiency when using compact particles and the significant one when using agglomerates; the similarities between guillotine and fish-mouth breaches in terms of efficiency, but their noticeable different deposition patterns; and the secondary effect of the breach size. Finally, the tube vibration is not as significant as the particle nature effect on the net deposition. The second part of the thesis is focused on the fraction of particles susceptible of leaving the containment in case of a severe accident regardless of the SGTR sequence. Accidents like Fukushima highlighted the importance of relying on efficient mitigation systems capable of reducing any release to the environment as much as possible. Although many reactors worldwide had installed filtered containment venting systems (FCVS) the interest in FCVS and even other mitigation systems has become of outstanding importance in nuclear safety. This is the frame of the PASSAM project in which an experimental sound database is being built to explore potential enhancement of existing source term mitigation devices and demonstrate the ability of innovative systems to achieve even larger source term attenuation. As a matter of fact, particle agglomeration processes via the propagation of acoustic vibrations through a gas could be applied for a better decontamination. High-intensity acoustic fields applied to an aerosol induce interaction effects among suspended particles, giving rise to successive collisions and agglomerations, resulting in larger particles that can be more easily removed or precipitated. The mitigative system acoustic agglomerator was built-up and tested in the AAA experimental campaign. The tests were conducted under a constant ultrasonic field with aerosols of different nature and size with different gas mass flow rates. The results pointed out two main insights: the small acoustic-agglomeration effect and the key effect of the gas mass flow rate and the aggregation state of the former particles in the agglomeration process. This research is the first approximation on the application of the ultrasonic chamber as an innovative system for the source term mitigation. / [ES] Durante un accidente severo en una central nuclear los productos de fisión liberados como consecuencia de la degradación del combustible podrían llegar a la atmósfera si se pierde la hermeticidad de la contención o si encuentran vías alternativas (bypass) para salir. Dada la radio-toxicidad del término fuente, las centrales nucleares deben contar con medios y medidas técnicas de seguridad para contener estos productos. En un reactor PWR, un caso particular de secuencia accidental donde los productos de fisión tienen acceso directo a la atmósfera, es aquella en el que además de la fusión de núcleo existe rotura de tubos del generador de vapor (secuencia SGTR). En este caso, es de vital importancia la evaluación del riesgo del suceso, objetivo de los proyectos internacionales EU-SGTR, ARTIST y ARTIST-2. Particularmente significativa es la "etapa de rotura" (break stage) del generador de vapor (SG), que es responsable de la retención de una fracción importante de partículas y de la evolución de su distribución a tamaños más pequeños. Estos motivos despertaron el interés hacia la propia retención de las partículas sobre los tubos y el efecto de variables como la naturaleza de la partícula, el tipo de rotura y la vibración de tubos sobre la retención en la etapa de rotura en condiciones secas; aspectos en los que se centra la primera parte de esta tesis. Con el objetivo de estudiar las cuestiones señaladas se han llevado a cabo dos campañas experimentales, CAAT2 y SET, con materiales enmarcados en el posible espectro de los aerosoles nucleares. La primera de ellas se centró en explorar la influencia potencial de la naturaleza de la partícula y el efecto del tipo de rotura de los tubos (forma y tamaño) sobre la retención de aerosoles. La segunda concierne la caracterización de la vibración de los tubos y el estudio de su efecto en la eficiencia de retención de partículas. Las pruebas realizadas resaltan varias ideas clave: el fuerte efecto de la naturaleza de la partícula sobre la retención en el lado secundario del SG; la alta eficiencia de retención cuando las partículas son compactas y la significativa retención cuando están aglomeradas; las pequeñas diferencias en eficiencia neta entre distintos tipos de rotura (guillotina vs. boca de pez) que resultan notables sobre los patrones de deposición, y el efecto secundario del tamaño de la rotura. Finalmente los resultados revelaron que frente a la naturaleza de la partícula, la vibración de tubos juega un papel secundario en la eficiencia de la retención. La segunda parte de este trabajo se centra en la fracción de partículas que es susceptible de alcanzar la contención en caso de accidente severo. Accidentes como el de Fukushima ponen de manifiesto la necesidad de tecnologías capaces de evitar las indeseadas consecuencias de la emisión de material radiactivo al medio ambiente. Esta es la dirección de investigación del proyecto PASSAM (7º Programa Marco de EURATOM) que está construyendo una base de datos experimental para el desarrollo de sistemas innovadores y la mejora de los sistemas de venteo filtrado de la contención que ya existen. Entre estos sistemas se encuentran las cámaras de ultrasonidos donde las ondas acústicas facilitan la aglomeración y el crecimiento de partículas, resultando sistemas potenciales para su mitigación. La campaña experimental AAA ha constituido una primera aproximación para la aplicación de las cámaras de ultrasonidos como sistemas innovadores para la mitigación del término fuente en la contención. El sistema de mitigación de aglomeración acústica (MSAA) se construyó y ha sido probado durante los experimentos AAA. Los resultados obtenidos ponen de manifiesto el leve efecto del campo acústico sobre el crecimiento de las partículas. Además, tanto el flujo másico de gas portador como la naturaleza de la partícula son claves en el proceso de aglomeración. / [CAT] En cas d'accident sever d'una central nuclear els productes de fissió resultants del combustible degradat podrien assolir l'atmosfera si es perd la hermeticitat de la contenció o si troben un camí alternatiu que l'evitin. Donada la naturalesa radio-tòxica dels aerosols nuclears ha d'evitar-se per tots els mitjans que surtin a l'exterior. En un reactor PWR, un cas particular d'accident és en el qual a més de la fusió de nucli existeix trencament de tubs del generador de vapor. En aquest cas, l'alliberament de material radioactiu cap al medi ambient fa que l'escenari sigui indispensable de modelar en l'avaluació del risc d'aquest reactor. Aquesta és la raó dels projectes internacionals EU-SGTR, ARTIST i ARTIST-2, gràcies als quals s'ha construït una extensa base de dades sobre el comportament dels aerosols en el circuit secundari del generador de vapor (Steam Generator, SG). Particularment significativa és l'etapa de trencament, que és responsable de la retenció d'una fracció important de partícules i de modificar la seva distribució cap a les mides més petites. Aquests motius van despertar l'interès vers l'efecte de variables com la naturalesa de la partícula, el tipus de trencament i la vibració de tubs sobre la retenció de partícules sobre els tubs en condicions seques a l'etapa de trencament del SG. Aquests són els aspectes en els quals es centra la primera part d'aquesta tesi. Dues campanyes experimentals, CAAT2 i SET, s'han dut a terme amb diferents materials, tots ells emmarcats dins del possible rang dels aerosols nuclears. La primera d'elles es va centrar a explorar la influència potencial de la naturalesa de la partícula i l'efecte del tipus de trencament (forma i grandària) sobre la retenció d'aerosols en els tubs. La segona va seguir per la caracterització en termes de vibració dels tubs i el seu efecte en l'eficiència de retenció de partícules. Les proves realitzades ressalten diverses idees clau: el fort efecte de la naturalesa de la partícula sobre la retenció en el costat secundari del SG; l'alta eficiència de retenció quan les partícules són compactes i la també significativa retenció quan són aglomerats; les petites diferències en eficiència entre diferents tipus de trencament (guillotina vs. boca de peix), però notables sobre els patrons de deposició, i l'efecte secundari de la grandària de trencament. Finalment van revelar que enfront de la naturalesa de la partícula, la vibració de tubs juga un paper secundari en l'eficiència de retenció del feix de tubs. La segona part d'aquesta tesi es centra en la fracció de partícules que en cas d'accident sever, amb o sense seqüència SGTR, és susceptible d'aconseguir la contenció. Accidents com Fukushima posen de manifest la necessitat de tecnologia capaç de cobrir les indesitjades conseqüències de l'emissió de material radioactiu al medi. Aquesta és la raó del projecte PASSAM (7é Programa Marc d'EURATOM) que està construint una base de dades experimental per al desenvolupament de sistemes innovadors i millorar els sistemes de venteig filtrat que ja existeixen de la contenció. Les ones d'ultrasons faciliten l'aglomeració de partícules i resulten sistemes potencials per a la seva mitigació. S'ha realitzat una primera aproximació per a l'aplicació de les càmeres d'ultrasons com a sistemes innovadors per a la mitigació del terme font en la contenció. El sistema de mitigació d'aglomeració acústica (MSAA) es va construir i ha estat provat durant la campanya experimental AAA. Els experiments duts a terme en la planta PECA-MSAA del LASS. Els resultats obtinguts posen de manifest dues idees: el sistema MSAA és efectiu en la reducció de la massa de partícules i tant el flux màssic de gas portador com la naturalesa de la partícula són claus en l'eficiència de retenció del sistema. / Delgado Tardáguila, R. (2016). Experimental investigations on nuclear aerosols in a severe accident [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/63243 / TESIS

The aim of the thesis is to analyze the state of preparedness of nuclear power plants (NPP) for long term operation (LTO) based on the IAEA SALTO (Safety Aspects of Long Term Operation) peer review service, analysis of the most significant failures, accidents and operational experience with type reactors PWR/VVER focusing on cases caused by equipment ageing and identification of major structures and components limiting life time of PWR/VVER-type nuclear power plants, and possible measures to ensure their required service life. Based on the results of the IAEA SALTO peer review service, an analysis of the main deficiencies and measures of NPPs in preparation for a safe LTO was performed, focusing on topics whose deeper knowledge is important for the future more precise determination of technical factors limiting the lifetime of NPPs. The main deficiencies and measures in the preparatory phase for LTO and the most important technical measures are summarized in chapter 4.5. The main deficiencies and the most important technical corrective measures in the area of ageing management of structures and components are discussed separately. The history of major failures and operational experience of nuclear power plants with PWR/VVER reactors from the point of view of ageing of structures and components is analyzed in chapter 6.2. The result is a statistic analysis of ageing-related events, an overview of the most significant PWR/VVER reactor failures with an impact on their service life, a statistical overview and discussion of the most important degradation mechanisms, and other important findings from the history of major failures and operational experience. Chapter 6.3 analyzes factors limiting the operation of nuclear power plants with PWR/VVER reactors with focus on structures and components potentially limiting the life of PWR/VVER reactors and possible measures to ensure their required life. In conclusion, the main reasons of permanent shut down of NPPs (actual and potential) for 40, 60 and 80 years of operation and the measures to ensure their required life are summarized.

The Institute of Resource Ecology (IRE) is one of the currently eight institutes of the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). The research activities are fully integrated into the program “Nuclear Safety Research” of the Helmholtz Association and focused on the topics “Safety of Nuclear Waste Disposal” and “Safety Research for Nuclear Reactors”. With the integration of the division of “Reactor Safety” from the former “Institute of Safety Research” nuclear research at HZDR is now mainly concentrated within this institute. In addition, various activities have been started investigating chemical and environmental aspects of processing and recycling of strategic metals, namely rare earth elements. Here, a knowledge transfer from the nuclear to the non-nuclear community, branching thermodynamics and spectroscopy, has been established. This also strengthens links to the recently established “Helmholtz-Institute Freiberg for Resource Technology”.

The system code TRACE is the latest development of the U.S. Nuclear Regulatory Commission (US NRC). TRACE, developed for the analysis of operational conditions, transients and accidents of light water reactors (LWR), is a best-estimate code with two fluid, six equation models for mass, energy, and momentum conservation, and related closure models. Since TRACE is mainly applied to LWR specific issues, the validation process related to innovative nuclear systems (liquid metal cooled systems, systems operated with supercritical water, etc.) is very limited, almost not existing. In this work, essential contribution to the validation of TRACE related to lead and lead alloy cooled systems as well as systems operated with supercritical water is provided in a consistent and corporate way. In a first step, model discrepancies of the TRACE source code were removed. This inconsistencies caused the wrong prediction of the thermo physical properties of supercritical water and lead bismuth eutectic, and hence the incorrect prediction of heat transfer relevant characteristic numbers like Reynolds or Prandtl number. In addition to the correction of the models to predict these quantities, models describing the thermo physical properties of lead and Diphyl THT (synthetic heat transfer medium) were implemented. Several experiments and numerical benchmarks were used to validate the modified TRACE version. These experiments, mainly focused on wall-to-fluid heat transfer, revealed that not only the thermo physical properties are afflicted with inconsistencies but also the heat transfer models. The models for the heat transfer to liquid metals were enhanced in a way that the code can now distinguish between pipe and bundle flow by using the right correlation. The heat transfer to supercritical water was not existing in TRACE up to now. Completely new routines were implemented to overcome that issue. The comparison of the calculations to the experiments showed, on one hand, the necessity of these changes and, on the other hand, the success of the new implemented routines and functions. The predictions using the modified TRACE version were close to the experimental data. After validating the modified TRACE version, two design studies related to the Generation IV International Forum (GIF) were investigated. In the first one, a core of a lead-cooled fast reactor (LFR) was analyzed. To include the interaction between the thermal hydraulic and the neutron kinetic due to temperature and density changes, the TRACE code was coupled to the program system ERANOS2.1. The results gained with that coupled system are in accordance with theory and helped to identify sub-assemblies with the highest loads concerning fuel and cladding temperature. The second design which was investigated was the High Performance Light Water Reactor (HPLWR). Since the design of the HPLWR is not finalized, optimization of vital parameters (power, mass flow rate, etc.) are still ongoing. Since most of the parameters are affecting each other, an uncertainty and sensitivity analysis was performed. The uncertainty analysis showed the upper and lower boundaries of selected parameters, which are of importance from the safety point of view (e.g., fuel and cladding temperature, moderator temperature). The sensitivity study identified the most relevant parameters and their influence on the whole system.

Thesis Abstract Current Nuclear Power Plant (NPP) design does not account for Beyond Design Basis Events (BDBEs) and thus lack the provisions to effectively mitigates complete loss of AC power and total loss of heat sink. Furthermore, parametric models used in PRA studies to assess Nuclear Power Plant’s safety risk for BDBE and External Events (EE) have significant limitations and proved ineffective to provide solutions on how to mitigate in BDBE or EEs situations. The Fukushima accident is a good example where PRA assessments did not provide the necessary means to cool or contain the reactors effectively. In this thesis, Emergency Mitigation Preparedness (EMP) model and assessment is proposed. The EMP model is objective and practical in evaluating NPP’s mitigation readiness in BDBE and EEs situations and provide a practical NPP Vulnerability indicator gauge which can potentially be used in risk-informed decisions. This will aid further in the NPP to improve in areas of emergency planning, enhance site and reactor design and improve workers safety and readiness to execute effective mitigation procedures and emergency plans. / Thesis / Master of Engineering (ME)

Nukleare Brennelemente werden nach ihrem Betrieb mehrere Jahre in Nasslagerbecken gelagert, wo ihre Nachzerfallswärme durch elektrisch betriebene Kühlsysteme abgeführt wird. Bei Ausfall der Stromversorgung droht eine Überhitzung der Brennelemente und im schlimmsten Fall die Schädigung der Brennstabhüllen und der Austritt von radioaktivem Material in die Umwelt. Im Mittelpunkt der vorliegenden Dissertation steht die Untersuchung des komplexen Zusammenspiels von Strömung und Wärmetransport bei solch einem angenommenen Unfall, der zu teilweise freigelegten Brennelementen führt. Eine Auswertung des aktuellen Forschungsstandes verdeutlicht, dass die zugrundeliegenden physikalischen Prozesse zwar theoretisch verstanden sind, aber bisher keine speziellen Simulationsprogramme zur präzisen Vorhersage der Temperaturverteilung für mögliche Unfallszenarien existieren. Für die detaillierte Analyse der Vorgänge werden deshalb erstmals numerische Strömungssimulationen unter Berücksichtigung der exakten Geometrie und aller relevanten Wärmetransportmechanismen für ein teilweise freigelegtes Brennelement durchgeführt. Zur Gewährleistung eines praktikablen Rechenaufwands wird der instationäre Verdampfungsvorgang in mehrere, eigenständige Simulationen mit stationären Randbedingungen und jeweils konstantem Füllstand unterteilt. Die Validierung mit experimentellen Daten zeigt, dass dieser Ansatz bei niedriger Nachzerfallsleistung geeignet ist, um die Stabtemperaturen mit ausreichender Genauigkeit vorherzusagen. Durch eine umfassende Sensitivitätsanalyse wird darüber hinaus der Einfluss zahlreicher unsicherer Faktoren auf die Temperaturverteilung und Zusammensetzung im Brennelement untersucht, der sich rein auf Grundlage des Experiments nicht beurteilen lässt. Die Simulationsergebnisse zeigen, dass die maximale Stabtemperatur hauptsächlich vom Füllstand und der Leistung der Brennstäbe abhängt. Eine horizontal gerichtete Luftströmung oberhalb des Brennelements führt insgesamt zu einem Temperaturgefälle in Strömungsrichtung innerhalb des Brennelements. Die Ursache dafür ist ein charakteristisches Strömungsfeld, bei dem kaltes Gas an der stromabwärts gelegenen Wand des Brennelements nach unten und heißes Gas an der stromaufwärts gelegenen Wand nach oben befördert wird. Die alleinige Variation der Geschwindigkeit der Luftströmung bewirkt jedoch keine nennenswerte Änderung der maximalen Stabtemperatur. Erst durch die Verwendung realitätsnaher Randbedingungen für Geschwindigkeit, Temperatur und Zusammensetzung, die aus großskaligen Simulationen des gesamten Lagerbeckens gewonnen wurden, wird der Einfluss der Querströmung auf die Temperaturverteilung im Brennelement deutlich. Bedingt durch das Verhältnis aus Auftriebs- zu Trägheitskräften, steigt die Temperatur im Brennelement bei einer Kombination aus geringer Temperatur, geringem Dampfmassenanteil und hoher Geschwindigkeit der Querströmung signifikant an. Diese Ergebnisse ermöglichen die Ableitung gezielter Beladungsstrategien von Lagerbecken, sofern die Randbedingungen oberhalb der Brennelemente hinreichend genau bekannt sind bzw. vorhergesagt werden können. Im letzten Schritt wird eine Methode zur skalenübergreifenden Modellierung eines Lagerbeckenbereichs vorgestellt. Durch die Kopplung zweier Modellierungsansätze wird eine teilweise geometrieauflösende Simulation ermöglicht, bei der das zentrale Brennelement geometrisch aufgelöst und die benachbarten Brennelemente als poröse Körper modelliert werden. Diese Vorgehensweise verbessert die Übertragbarkeit der Ergebnisse auf ein ganzes Lagerbecken, weil die Auswertung im geometrisch aufgelösten Brennelement unabhängiger von den mit Unsicherheit behafteten Randbedingungen wird.:1 Einleitung 1 1.1 Chancen und Risiken der Kernenergienutzung 1 1.2 Randbedingungen für den Wärme- und Stofftransport im Lagerbecken 3 1.2.1 Zerfallsleistung 3 1.2.2 Brennelement-Typ und Aufbau 4 1.2.3 Wärmetransportmechanismen 6 1.2.4 Verdampfungsrate 8 1.2.5 Grenztemperaturen 9 1.3 Simulation des Wärme- und Stofftransports im Lagerbecken 10 1.3.1 Das Lagerbecken als Multiskalenproblem 10 1.3.2 Systemcodes und Codes für schwere Störfälle 12 1.3.3 CFD-Simulation mit Brennelementen als poröse Körper 13 1.3.4 Geometrieauflösende CFD-Simulation 15 1.4 Zielstellung und Aufbau der Arbeit 16 2 Modell für ein ausdampfendes Brennelement 19 2.1 Vorbetrachtungen 19 2.1.1 Strömungsform 19 2.1.2 Form des Wärmeübergangs 22 2.2 Physikalische Modellierung 23 2.2.1 Simulationsstrategie 23 2.2.2 Physikalische Modellgleichungen 24 2.2.3 Rechengebiet und Randbedingungen 27 2.3 Numerische Modellierung 32 2.3.1 Örtliche Diskretisierung 32 2.3.2 Zeitliche Diskretisierung 34 3 Sensitivitätsanalyse für ein ausdampfendes Brennelement 37 3.1 Vorgehensweise 37 3.2 Einfluss der Strahlungsmodellierung 39 3.2.1 Motivation 39 3.2.2 Bestimmung des Absorptionskoeffzienten 40 3.2.3 Einfluss der Gasstrahlung 41 3.2.4 Einfluss der numerischen Parameter 44 3.3 Einfluss unsicherer Randbedingungen 46 3.3.1 Wärmeverlust über die Isolierschicht 46 3.3.2 Verteilung des Dampfmassenstroms an der Wasseroberfläche 51 3.4 Einfluss der effektiv freigelegten Länge der Heizstäbe 56 3.5 Einfluss der Stableistung 58 4 Wechselwirkung zwischen Querüberströmung und Wärmetransport im Brennelement 63 4.1 Rechengebiet und Randbedingungen 63 4.2 Physikalische und numerische Modellierung 65 4.2.1 Physikalische Modellierung 65 4.2.2 Numerische Einstellungen 67 4.3 Ergebnisse und Diskussion 67 4.3.1 Generelles Vorgehen 67 4.3.2 Temperaturentwicklung und Strömung im Stabbereich 69 4.3.3 Temperatur und Strömung im Überströmkanal 75 5 Ansätze zur skalenübergreifenden Modellierung eines Lagerbeckens 81 5.1 Einordnung 81 5.2 Co-Simulation des Wärme- und Stoffaustauschs zwischen Einzelbrennelement und Lagerbeckenatmosphäre 81 5.2.1 Konfiguration 81 5.2.2 Einfluss der Konvektionsströmung oberhalb der Brennelemente 86 5.3 Gekoppelte Simulation eines Lagerbeckenbereichs 92 5.3.1 Motivation 92 5.3.2 Parametrierung des porösen Körpers 92 5.3.3 Vergleich der Simulationsansätze 94 5.3.4 Simulation der Brennelement-Gruppe 96 6 Zusammenfassung und Ausblick 101 Literaturverzeichnis 115 Symbol- und Abkürzungsverzeichnis 119 / After their operation, spent nuclear fuel assemblies are stored for several years in wet storage pools, where their decay heat is removed by electrically operated cooling systems. If the power supply fails, this poses the risk of overheating of the fuel assemblies and, in the worst case, damage to the fuel rod cladding and the release of radioactive material into the environment. This dissertation focuses on the investigation of the complex interaction of flow and heat transport in such an assumed accident, which leads to partially uncovered fuel assemblies. A review of the current state of research illustrates that although the underlying physical processes are theoretically understood, no specific simulation programmes exist to date to accurately predict the temperature distribution for possible accident scenarios. For the detailed analysis of the processes, numerical flow simulations taking into account the exact geometry and all relevant heat transport mechanisms are therefore carried out for a partially uncovered fuel assembly for the first time. To ensure a manageable computational effort, the transient evaporation process is subdivided into several, independent simulations with steady boundary conditions and a constant water level in each case. The validation with experimental data shows that this approach is suitable for predicting the rod temperatures with sufficient accuracy for low decay heat. A comprehensive sensitivity analysis also identifies the influence of numerous uncertain factors on the temperature distribution and composition in the fuel assembly, which cannot be assessed purely on the basis of the experiment. The simulation results show that the maximum rod temperature depends mainly on the water level and the power of the fuel rods. A horizontally directed air flow above the fuel assembly leads to an overall temperature gradient in the flow direction within the fuel assembly. This is caused by a characteristic flow field in which cold gas is transported down the downstream wall of the fuel assembly and hot gas is transported up the upstream wall. However, varying the velocity of the airflow alone does not cause a significant change in the maximum rod temperature. The influence of the crossflow on the temperature distribution in the fuel assembly only becomes clear by using realistic boundary conditions for velocity, temperature and composition, obtained from large-scale simulations of the entire storage pool. Determined by the ratio of buoyant to inertial forces, the temperature in the fuel assembly increases significantly with a combination of low temperature, low steam mass fraction and high velocity of the crossflow. These results provide information on how to best arrange fuel assemblies in spent fuel pools, provided that the boundary conditions above the fuel assemblies are known or can be predicted with sufficient accuracy. Finally, a method for modelling a larger part of the spent fuel pool is presented. The combination of two modelling approaches enables a partially geometry-resolving simulation in which the central fuel assembly is geometrically resolved and the neighbouring fuel assemblies are modelled as porous bodies. This approach improves the transferability of the results to an entire spent fuel pool, because the evaluation in the geometrically resolved fuel assembly becomes more independent from the uncertain boundary conditions.:1 Einleitung 1 1.1 Chancen und Risiken der Kernenergienutzung 1 1.2 Randbedingungen für den Wärme- und Stofftransport im Lagerbecken 3 1.2.1 Zerfallsleistung 3 1.2.2 Brennelement-Typ und Aufbau 4 1.2.3 Wärmetransportmechanismen 6 1.2.4 Verdampfungsrate 8 1.2.5 Grenztemperaturen 9 1.3 Simulation des Wärme- und Stofftransports im Lagerbecken 10 1.3.1 Das Lagerbecken als Multiskalenproblem 10 1.3.2 Systemcodes und Codes für schwere Störfälle 12 1.3.3 CFD-Simulation mit Brennelementen als poröse Körper 13 1.3.4 Geometrieauflösende CFD-Simulation 15 1.4 Zielstellung und Aufbau der Arbeit 16 2 Modell für ein ausdampfendes Brennelement 19 2.1 Vorbetrachtungen 19 2.1.1 Strömungsform 19 2.1.2 Form des Wärmeübergangs 22 2.2 Physikalische Modellierung 23 2.2.1 Simulationsstrategie 23 2.2.2 Physikalische Modellgleichungen 24 2.2.3 Rechengebiet und Randbedingungen 27 2.3 Numerische Modellierung 32 2.3.1 Örtliche Diskretisierung 32 2.3.2 Zeitliche Diskretisierung 34 3 Sensitivitätsanalyse für ein ausdampfendes Brennelement 37 3.1 Vorgehensweise 37 3.2 Einfluss der Strahlungsmodellierung 39 3.2.1 Motivation 39 3.2.2 Bestimmung des Absorptionskoeffzienten 40 3.2.3 Einfluss der Gasstrahlung 41 3.2.4 Einfluss der numerischen Parameter 44 3.3 Einfluss unsicherer Randbedingungen 46 3.3.1 Wärmeverlust über die Isolierschicht 46 3.3.2 Verteilung des Dampfmassenstroms an der Wasseroberfläche 51 3.4 Einfluss der effektiv freigelegten Länge der Heizstäbe 56 3.5 Einfluss der Stableistung 58 4 Wechselwirkung zwischen Querüberströmung und Wärmetransport im Brennelement 63 4.1 Rechengebiet und Randbedingungen 63 4.2 Physikalische und numerische Modellierung 65 4.2.1 Physikalische Modellierung 65 4.2.2 Numerische Einstellungen 67 4.3 Ergebnisse und Diskussion 67 4.3.1 Generelles Vorgehen 67 4.3.2 Temperaturentwicklung und Strömung im Stabbereich 69 4.3.3 Temperatur und Strömung im Überströmkanal 75 5 Ansätze zur skalenübergreifenden Modellierung eines Lagerbeckens 81 5.1 Einordnung 81 5.2 Co-Simulation des Wärme- und Stoffaustauschs zwischen Einzelbrennelement und Lagerbeckenatmosphäre 81 5.2.1 Konfiguration 81 5.2.2 Einfluss der Konvektionsströmung oberhalb der Brennelemente 86 5.3 Gekoppelte Simulation eines Lagerbeckenbereichs 92 5.3.1 Motivation 92 5.3.2 Parametrierung des porösen Körpers 92 5.3.3 Vergleich der Simulationsansätze 94 5.3.4 Simulation der Brennelement-Gruppe 96 6 Zusammenfassung und Ausblick 101 Literaturverzeichnis 115 Symbol- und Abkürzungsverzeichnis 119

Le 11 mars 2011, le Japon a subi un séisme suivi d’un tsunami aux conséquences terribles. Dans la centrale de Fukushima Dai-ichi s’est produit un accident nucléaire de niveau 7 (le plus élevé) sur l’échelle internationale, qui a marqué les esprits comme celui de Tchernobyl en 1986. Cet accident a laissé le monde en émoi face à ces nouvelles formes de menaces, d’autant que l’exploitant TEPCO n’a pas su maitriser la situation ni tirer les leçons du passé. Depuis Fukushima, l’échelle des fondamentaux en Europe et dans le monde a donc été bouleversée et la question de la sûreté et de la sécurité des centrales se pose avec une acuité renforcée, qui a nécessité de redéfinir en droit et en pratique certaines normes et principes au niveau national, européen et international en concordance avec ces nouvelles menaces extérieures, vers le plus haut niveau de sûreté. Mais les révisions entreprises nécessitent d’être plus ambitieuses. L’avenir du nucléaire implique dès lors : au niveau européen, une révision plus ambitieuse de la directive sûreté; la mise en place d’une autorité de réglementation indépendante de jure ; la définition d’un droit de la responsabilité civile harmonisé au sein de l’UE en faveur des victimes dans l’hypothèse d’un accident. Au niveau international, la gouvernance s’impose comme étant le vecteur d’une commune culture de sûreté et de sécurité nucléaires ; bien que la diversité des modèles nationaux de gestion et de contrôle de l’industrie nucléaire paraisse rendre a priori difficile l’évolution vers des règles communes. De même au niveau européen, dans ce même esprit, l’écriture d’un texte unique en droit de la réparation des dommages serait nécessaire. La révision de la Convention sûreté nucléaire est également un chantier important pour l’avenir. Dans l’immédiat, l’harmonisation concerne de nombreux domaines dont, pour l’essentiel : la gestion de crise pendant et après un accident nucléaire ; la mise en place des principes de sûreté et de sécurité les plus performants et les plus élevés, de la conception au démantèlement d’une installation ; la maîtrise d’une interaction adaptée entre sûreté et sécurité nucléaires. Il conviendra, par ailleurs, de veiller à l’intégration du public au processus décisionnel dans les domaines du nucléaire, condition nécessaire à l’acceptabilité de cette énergie. / On March 11, 2011, the Japan suffered an earthquake followed by a tsunami to the terrible consequences. In nuclear power plant Fukushima Dai-ichi happened a nuclear accident of level 7 (highest) on the international scale, which marked the spirits such as rivaled that of Chernobyl in 1986. This accident left the world agog with these new forms of threats, especially since the TEPCO operator did not master the situation or learn the lessons of the past. Since Fukushima, the fundamentals in Europe and worldwide has so upset been turned upside-down and this raises the question of safety and security of power plants with renewed acuity, which necessitated. It is imperative to redefine in law and in practice some standards and principles at the national, European and international level in accordance with these new threats to the highest level of safety. But the legal revisions need to be more ambitious. The future of nuclear power suggest therefore: at the European level: a more ambitious revision of the directive on nuclear safety; the establishment of a regulatory body with effective independence de jure ; the definition of a liability law harmonised throughout the EU and the IAEA for victims in the event of an accident. At the international level: the governance is necessary as a vector of a common safety culture and security culture ; although the diversity of national models of management and control of the nuclear industry appears a priori difficult to move towards common rules. As well as at the European level, the writing of a single text entitled to the repair of damages would be necessary for the same reasons already stated. The revision of the Convention on nuclear safety is also as important crucial for the future. For immediate harmonization concerns many fields, for the most part: during and after a nuclear accident crisis management; the implementation of the principles of safety and security at the most efficient and highest level from the conception to the dismantling of an installation; strengthening interaction adapted between nuclear safety and nuclear security ; but also the integration of the population in the decision-making process in the areas of nuclear is mandatory for the acceptance of nuclear energy.

La probabilité et la causalité sont deux outils indispensables à la prise en compte des situations de risque social. Lesrelations causales sont le fondement des représentations à partir desquelles on peut évaluer le risque et concevoirdes actions de prévention, de mitigation ou d’indemnisation. La probabilité permet de quantifier cette évaluation et de calibrer ces actions. Dès lors, il semble non seulement naturel, mais nécessaire d’expliciter la place de la causalité et de la probabilité dans la définition d’un problème de décision en situation de risque social. C’est l’objet de cette thèse.Un tour d’horizon de la terminologie du risque et des logiques d’intervention publique dans différentes catégories de risque social nous permettent de mieux comprendre la notion et les problèmes soulevés par sa représentation. Nous approfondissons notre analyse dans le cas de la sûreté nucléaire, en examinant en détail les méthodes et doctrinesdéveloppées dans ce domaine et leur évolution au cours du temps, ce qui nous conduit à formuler différentesobservations au sujet des évaluations de risque et de sûreté.En généralisant la notion d’intervention dans les réseaux bayésiens, nous développons une forme de réseau bayésien causal qui répond à nos besoins. Nous parvenons, par son biais, à une définition du risque qui semble pertinente pour un grand nombre de situations. Nous proposons ensuite des applications simples de ce modèle à certains aspects de l’accident de Fukushima et d’autres problèmes de sûreté nucléaire. Outre certains enseignements spécifiques, ceci nous amène à souligner la nécessité d’une démarche systématique d’identification des incertitudes dans ce domaine.Étendu en direction de la théorie de la décision, notre outil débouche naturellement sur un modèle de décision dynamique dans lequel les actes causent les conséquences et sont causalement liés entre eux. Il apporte en outre une interprétation causale au cadre conceptuel de Savage et permet d’en résoudre certains paradoxes et clarifier certains aspects. Il conduit enfin à envisager la question de l’ambigüité comme incertitude concernant la structure causale d’un problème de décision, ce qui correspond à une vision courante du principe de précaution. / Probability and causality are two indispensable tools for addressing situations of social risk. Causal relations are the foundation for building risk assessment models and identifying risk prevention, mitigation and compensation measures. Probability enables us to quantify risk assessments and to calibrate intervention measures. It therefore seems not only natural, but also necessary to make the role of causality and probability explicit in the definition of decision problems in situations of social risk. Such is the aim of this thesis.By reviewing the terminology of risk and the logic of public interventions in various fields of social risk, we gain a better understanding of the notion and of the issues that one faces when trying to model it. We further elaborate our analysis in the case of nuclear safety, examining in detail how methods and policies have been developed in this field and how they have evolved through time. This leads to a number of observations concerning risk and safety assessments.Generalising the concept of intervention in a Bayesian network allows us to develop a variety of causal Bayesian networks adapted to our needs. In this framework, we propose a definition of risk which seems to be relevant for a broad range of issues. We then offer simple applications of our model to specific aspects of the Fukushima accident and other nuclear safety problems. In addition to specific lessons, the analysis leads to the conclusion that a systematic approach for identifying uncertainties is needed in this area.When applied to decision theory, our tool evolves into a dynamic decision model in which acts cause consequencesand are causally interconnected. The model provides a causal interpretation of Savage’s conceptual framework, solves some of its paradoxes and clarifies certain aspects. It leads us to considering uncertainty with regard to a problem’s causal structure as the source of ambiguity in decision-making, an interpretation which corresponds to a common understanding of the precautionary principle.