L’analyse de sûreté des réacteurs nucléaires nécessite la modélisation fine des phénomènes y survenant et plus spécifiquement ceux permettant d’assurer l’intégrité des barrières de confinement. Les outils de modélisation et codes actuels favorisent une analyse fine du système réacteur par discipline dédiée, et couplée avec des modèles simplifiés. Néanmoins, le développement depuis plusieurs années d’une approche dite « Best Estimate », basée sur des calculs multiphysiques et multi-échelle, est en cours de réalisation. Cette approche permettra d’accéder au suivi et à l’analyse détaillée de problèmes complexes tels que l’étude des Réacteurs nucléaires en situation standard et accidentelle. Dans cette approche, les phénomènes physiques sont simulés aussi précisément que possible (selon la connaissance actuelle) par les modèles couplés. Par exemple, des codes disciplinaires existent et permettent la modélisation précise de la neutronique, de la thermohydraulique du cœur du réacteur ou de la thermohydraulique sur l'ensemble du système, de la thermomécanique du combustible ou des structures. Une approche « Best Estimate » consiste à coupler ces modèles afin de réaliser une modélisation globale et précise du système de réacteur nucléaire. Cette approche nécessite de bien définir les modèles qui sont utilisés afin de préciser exactement leurs limites, et donc préciser les incertitudes des résultats des modèles couplés afin de les assumer et de les optimiser.C’est dans ce contexte de travail que s’inscrit cette thèse. Elle consiste dans le développement d'un couplage multiphysique et multi-échelle « Best Estimate » afin d'obtenir une analyse précise des Réacteurs à Eau Légère en situations normale et accidentelle. Elle a consisté principalement en l’analyse des modèles et de leurs interactions et à la mise en œuvre d'un algorithme de couplage multiphysique entre une neutronique et une thermohydraulique exprimées à l'échelle du réacteur, ainsi qu’avec une thermomécanique fine à l'échelle élémentaire du crayon combustible. En outre, un travail spécifique a été effectué afin de préparer ou d'améliorer l’accés à l'information physique locale nécessaire à la mise en œuvre de modélisations couplées multi-échelles, à l'échelle du combustible. / The safety analysis of nuclear power plants requires a deep understanding of underlying key physical phenomena that determine the integrity of the physical containment barriers. At the present time, cutting edge models focus on a single aspect (discipline) of the physical system coupled with rough models of the other aspects needed to simulate the global system. But, safety analyses can be carried out based on Multiphysics and Multiscales modelling. This Best Effort approach would give a full and accurate (High Fidelity) comprehension of the reactor core under standard and accidental situations. In this approach, the physical phenomena are simulated as accurately as possible (according to present knowledge) by coupled models in the most efficient way. For example, codes exists that are accurate modellings of Neutronics, or modellings of thermal fluid mechanics inside the core, or modellings of thermal fluid mechanics over the whole system, or modellings of thermal mechanics of the fuel pin or over the whole device structure. A Best Estimate approach would couple these models in order to realize a global and accurate modelling of the Nuclear reactor. This approach requires to define well the models that are used in order to exactly specify their limits, and hence, specify uncertainties of the coupled model results in order to assume and optimize them.It is in this context that this PhD thesis work is being under taken. It consists in the development of a Multi-physics and multi-scale Best Estimate modelling in order to obtain an accurate analysis of Pressurized Water Reactor under standard and accidental operating situations. It mainly involves the understanding of each model and their interactions, followed by the implementation of multiphysics algorithms coupling Neutronics and Thermohydraulics at reactor scale to an accurate Thermomechanics at the elementary scale of the fuel pin. In addition, a work project has been carried out in order to prepare or improve the access to the local physical informations that are needed for the implementation of multiscale coupling scheme, at the elementary scale of the fuel pin.

The purpose of this study is to evaluate the effects and importance of different cross–section representations and energy group structures for steady state and transient analysis. More energy groups may be more accurate, but the calculation becomes much more expensive, hence a balance between accuracy and calculation effort must be find. This study is aimed at comparing a multi–group energy dynamics code, MGT (Multi–group TINTE) with TINTE (TIme Dependent Neutronics and TEmperatures). TINTE’s original version (version 204d) only distinguishes between two energy group structures, namely thermal and fast region with a polynomial reconstruction of cross–sections pre–calculated as a function of different conditions and temperatures. MGT is a TINTE derivative that has been developed, allowing a variable number of broad energy groups. The MGT code will be benchmarked against the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark: the PBMR–400 core design. This comparative study reveals the variations in the results when using two different methods for cross–section generation and multi–group energy structure. Inputs and results received from PBMR (Pty) Ltd. were used to do the comparison. A comparison was done between two–group TINTE and the equivalent two energy groups in MGT as well as between 4, 6 and 8 energy groups in MGT with the different cross–section generation methods, namely inline spectrum– and tabulated cross–section method. The characteristics that are compared are reactor power, moderation– and maximum fuel temperatures and k–effective (only steady state case). This study revealed that a balance between accuracy and calculation effort can be met by using a 4–group energy group structure. A larger part of the available increase in accuracy can be obtained with 4–groups, at the cost of only a small increase in CPU time. The changing of the group structures in the steady state case from 2 to 8 groups has a greater influence on the variation in the results than the cross–section generation method that was used to obtain the results. In the case of a transient calculation, the cross–section generation method has a greater influence on the variation in the results than on the steady state case and has a similar effect to the number of energy groups. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

The purpose of this study is to evaluate the effects and importance of different cross–section representations and energy group structures for steady state and transient analysis. More energy groups may be more accurate, but the calculation becomes much more expensive, hence a balance between accuracy and calculation effort must be find. This study is aimed at comparing a multi–group energy dynamics code, MGT (Multi–group TINTE) with TINTE (TIme Dependent Neutronics and TEmperatures). TINTE’s original version (version 204d) only distinguishes between two energy group structures, namely thermal and fast region with a polynomial reconstruction of cross–sections pre–calculated as a function of different conditions and temperatures. MGT is a TINTE derivative that has been developed, allowing a variable number of broad energy groups. The MGT code will be benchmarked against the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark: the PBMR–400 core design. This comparative study reveals the variations in the results when using two different methods for cross–section generation and multi–group energy structure. Inputs and results received from PBMR (Pty) Ltd. were used to do the comparison. A comparison was done between two–group TINTE and the equivalent two energy groups in MGT as well as between 4, 6 and 8 energy groups in MGT with the different cross–section generation methods, namely inline spectrum– and tabulated cross–section method. The characteristics that are compared are reactor power, moderation– and maximum fuel temperatures and k–effective (only steady state case). This study revealed that a balance between accuracy and calculation effort can be met by using a 4–group energy group structure. A larger part of the available increase in accuracy can be obtained with 4–groups, at the cost of only a small increase in CPU time. The changing of the group structures in the steady state case from 2 to 8 groups has a greater influence on the variation in the results than the cross–section generation method that was used to obtain the results. In the case of a transient calculation, the cross–section generation method has a greater influence on the variation in the results than on the steady state case and has a similar effect to the number of energy groups. / Thesis (M.Ing. (Nuclear Engineering))--North-West University, Potchefstroom Campus, 2011.

Les critères de sureté qui régissent le développement de coeurs de réacteurs dequatrième génération implique l’usage d’outils de calcul neutronique performants. Unepremière partie de la thèse reprend toutes les étapes de modélisation neutronique desréacteurs rapides actuellement d’usage dans le code de référence ECCO. La capacité desmodèles à décrire le phénomène d’autoprotection, à représenter les fuites neutroniques auniveau d’un réseau d’assemblages combustibles et à générer des sections macroscopiquesreprésentatives est appréciée sur le domaine des réacteurs rapides innovants respectant lescritères de quatrième génération. La deuxième partie de ce mémoire se consacre à lamodélisation des coeurs rapides avec réflecteur acier. Ces derniers nécessitent ledéveloppement de méthodes avancées de condensation et d’homogénéisation. Plusieursméthodes sont proposées et confrontées sur un problème de modélisation typique : le coeurZONA2B du réacteur maquette MASURCA. / Due to safety rules running on fourth generation reactors’ core development,neutronics simulation tools have to be as accurate as never before. First part of this reportenumerates every step of fast reactor’s neutronics simulation implemented in currentreference code: ECCO. Considering the field of fast reactors that meet criteria of fourthgeneration, ability of models to describe self-shielding phenomenon, to simulate neutronsleakage in a lattice of fuel assemblies and to produce representative macroscopic sections isevaluated. The second part of this thesis is dedicated to the simulation of fast reactors’ corewith steel reflector. These require the development of advanced methods of condensationand homogenization. Several methods are proposed and compared on a typical case: theZONA2B core of MASURCA reactor.

El Generation IV International Forum (GIF) [1] es un programa internacional dedicado a apoyar, coordinar y dirigir las iniciativas de investigación y desarrollo encaminados a implementar las soluciones tecnológicas que caracterizarán a la siguiente generación de reactores nucleares. Estos reactores se caracterizaran por una gestión más eficiente del combustible nuclear, un incremento en las exigencias de seguridad y una alta competitividad económica. Con tales objetivos, GIF propuso una serie de diseños potencialmente capaces de alcanzarlos. Estos diseños son tecnológicamente muy distintos a las plantas nucleares comerciales actuales al utilizar neutrones de espectro rápido y consecuentemente refrigeración por metales líquidos. Estos nuevos diseños requieren el desarrollo y validación de herramientas computacionales capaces de simular el comportamiento de la planta tanto en fase estacionaria como en transitoria y por tanto sean aplicables en los procesos de diseño y licitación de dichas plantas. El objetivo de esta tesis es el de adaptar los códigos computacionales actuales aplicados a la simulación de reactores refrigerados por agua a reactores rápidos refrigerados por metales líquidos, tales como el sodio o el plomo y el desarrollo de modelos capaces de simular de una manera consistente el comportamientos de los sistemas ante determinados eventos que constituyen la base de diseño de la planta Para ello se adaptaran dichos códigos a la fenomenología específica de estos reactores, se desarrollaran modelos termo-hidráulicos y neutrónicos tanto unidimensionales como tridimensionales de los diseños propuestos y se validarán los resultados para demostrar su aplicabilidad. El trabajo incluye la implementación de correlaciones específicas para habilitar los códigos para el cálculo de la condiciones termo-hidráulicas de los refrigerantes así como la adaptación de los esquemas de acoplamiento termo-hidráulico-neutrónicos existentes a esta nueva tecnología. / Lázaro Chueca, A. (2014). Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors [Tesis doctoral no publicada]. Universitat Politècnica de València. https://doi.org/10.4995/Thesis/10251/39353 / TESIS

The demand for computational tools that are capable to reliably predict the behavior of a nuclear reactor core in a variety of static and dynamic conditions does inevitably require a proper qualification of these tools for the intended purposes. One of the qualification methods is the verification of the code in question. Hereby, the correct implementation of the applied model as well as its flawless implementation in the code are scrutinized. The present work concerns with benchmarking as a substantial part of the verification of the three-dimensional, multigroup neutron kinetics model employed in the transient code POLCA-T. The benchmarking is done by solving some specified and widely used space-time kinetics benchmark problems and comparing the results to those of other, established and well-proven spatial kinetics codes. It is shown that the obtained results are accurate and consistent with corresponding solutions of other codes. In addition, a sensitivity analysis is carried out with the objective to study the sensitivity of the POLCA-T neutronics to variations in different numerical options. It is demonstrated that the model is numerically stable and provide reproducible results for a wide range of various numerical settings. Thus, the model is shown to be rather insensitive to significant variations in input, for example. The other consequence of this analysis is that, depending on the treated transient, the computing costs can be reduced by, for instance, employing larger time-steps during the time-integration process or using a reduced number of iterations. Based on the outcome of this study, one can finally conclude that the POLCA-T neutron kinetics is modeled and implemented correctly and thus, the model is fully capable to perform the assigned tasks.

La nécessité de simulations précises d’un réacteur nucléaire et spécialement dans des cas de cœurs et de configurations de combustible complexes, a imposé un usage accru de Codes Neutroniques Stochastiques (CNS). De plus, une demande a émergé pour des CNS à capacité inhérente d’estimation en continu de la variation de la composition isotopique du cœur ainsi qu’à couplage thermo-hydraulique optimisé. Des capacités supplémentaires sont exigées pour ces codes au vu de leur utilisation pour l’étude de nouveaux concepts de réacteur comme les Réacteurs Conduits par Accélérateur (RCA). Plus précisément, le réacteur hybride comprenant un réacteur nucléaire conventionnel et un accélérateur, nécessite l’analyse des deux composantes (réacteur – accélérateur) par un outil capable de couvrir le spectre énergétique neutronique extrêmement étendu qui caractérise ce système hybride.Ce travail présente les principales caractéristiques et capacités du nouveau CNS ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback) développé en collaboration du NCSR Demokritos (Grèce) avec CNRS/IDRIS et UPMC (France) et couvrant autant que possible les exigences exposées ci-dessus. ANET est basé sur la version ouverte du code PHE GEANT3.21 et est destiné à effectuer des analyses de cœurs de réacteurs conventionnels de génération II et III ainsi que des RCA. ANET est construit avec la capacité inhérentea) d’effectuer des calculs d’évolution du combustibleb) de simuler le processus de spallation dans le cas des RCAtout en tenant compte de la thermo-hydraulique du système.La version actuelle d’ANET utilise les trois estimateurs standard Monte Carlo pour le calcul du facteur de multiplication neutronique effectif (keff), soit l’estimateur de collision, celui d’absorption et celui de longueur de trace. Pour ce qui est du calcul du débit de fluence neutronique et des taux de réaction, les estimateurs de collision et de longueur de trace sont implémentés dans ANET suivant la procédure standard Monte Carlo. Pour ce qui concerne les calculs d’évolution (par exemple la consommation du combustible), une approche purement stochastique est implémentée dans ANET. A noter que la procédure usuelle consiste à coupler le code neutronique stochastique avec un code déterministe qui calcule la consommation du combustible. Pour les besoins d’analyse des RCA, le module INCL/ABLA a été incorporé dans ANET de façon à ce que le processus de spallation soit simulé par le code. La capacité d’ANET de simuler des configurations classiques a été démontrée en utilisant des résultats de mesures et des simulations de vérification effectuées en utilisant d’autres codes bien établis, ainsi qu’il est montré par la suite.Des données provenant de plusieurs installations et des analyses de problèmes-type internationaux ont été utilisés pour vérifier et valider les capacités d’ANET.Pour conclure, les résultats obtenus lors des comparaisons avec des mesures ou avec des simulations effectuées en utilisant d’autres codes neutroniques stochastiques ou déterministes, montrent qu’ANET possède la capacité de calculer correctement d’importants paramètres de systèmes critiques ou sous-critiques. Par ailleurs, l’application préliminaire d’ANET à des problèmes dépendant du temps fournit des résultats encourageants. ANET produit des estimations de consommation de combustible raisonnables, compte tenu que des incertitudes dans ce domaine sont souvent de l’ordre de 20% ou plus. Finalement, les performances du code dans le cas de KUCA montrent qu’ANET peut analyser des RCA de façon satisfaisante. / The necessity for precise simulations of a nuclear reactor especially in case of complex core and fuel configurations has imposed the increasing use of Monte Carlo (MC) neutronics codes. Besides, a demand of additional stochastic codes’ inherent capabilities has emerged regarding mainly the simulation of the temporal variations in the core isotopic composition as well as the incorporation of the T-H feedback. In addition to the above, the design of innovative nuclear reactor concepts, such as the Accelerator Driven System (ADSs), imposed extra requirements of simulation capabilities. More specifically, the combination of an accelerator and a nuclear reactor in the ADS requires the simulation of both subsystems for an integrated system analysis. Therefore a need arises for more advanced simulation tools, able to cover the broad neutrons energy spectrum involved in these systems.This work presents the main features and capabilities of the new MC neutronics code ANET (Advanced Neutronics with Evolution and Thermal hydraulic feedback), being developed in NCSR Demokritos (Greece) in cooperation with CNRS/IDRIS and UPMC (France) and intending to meet as effectively as possible the above described modelling requirements. ANET is based on the open-source version of the HEP code GEANT3.21 and is targeting to the creation of an enhanced computational tool in the field of reactor analysis, capable of simulating both GEN II/III reactors and ADSs. ANET is structured with inherent capability of (a) performing burnup calculations and (b) simulating the spallation process in the ADS analysis, while taking T-H feedback into account.The current ANET version utilizes the three standard Monte Carlo estimators for the neutron multiplication factor (keff) calculation, i.e. the collision estimator, the absorption estimator and the track-length estimator. Regarding the simulation of neutron fluence and reaction rates, the collision and the track-length estimators are implemented in ANET following the standard Monte Carlo procedure. For the burnup calculations ANET attempts to apply a pure Monte Carlo approach, adopting the typical procedure followed in stochastic codes. With respect to code improvements for the ADS analysis, so far ANET has incorporated the INCL/ABLA code so that the spallation process can be inherently simulated. The ANET reliability in typical computations was tested using observational data and parallel simulations by different codes as described in the following chapters.Various installations and international benchmarks were considered suitable for the verification and validation of all the previously mentioned features incorporated in the new code ANET. The obtained results are compared with experimental data from the nuclear infrastructures and with computations performed by well-established stochastic or deterministic neutronics codes and show satisfactory agreement with both measurements and independent computations, verifying thus ANET’s ability to successfully simulate important parameters of critical and subcritical systems. Also, the preliminary ANET application for dynamic analysis is encouraging since it indicates the code capability to inherently provide a reasonable prediction for the core inventory evolution taking into account the uncertainties of the order of 20% and even higher that are traditionally expected in core inventory evolution calculations. Lastly, the code performance in the KUCA case was found satisfactory demonstrating thus inherent capability of analyzing ADSs.

Perhaps surprisingly, the largest experience in operating nuclear power plants has been in nuclear naval propulsion, particularly submarines. This accumulated experience may become the basis of a proposed new generation of compact nuclear power plant designs. In an effort to de-carbonise commercial freight shipping, there is growing interest in the possibility of using nuclear propulsion systems. Reactor cores for such an application would need to be fundamentally different from land-based power generation systems, which require regular refueling, and from reactors used in military submarines, as the fuel used could not conceivably be as highly enriched. Nuclear-powered propulsion would allow ships to operate with low fuel costs, long refueling intervals, and minimal emissions; however, currently such systems remain largely confined to military vessels. This research project undertakes computational modeling of possible soluble-boron-free (SBF) reactor core designs for this application, with a view to informing design decisions in terms of choices of fuel composition, materials, core geometry and layout. Computational modeling using appropriate reactor physics (e.g. WIMS, MONK, Serpent and PANTHER), thermal-hydraulics etc. codes (e.g. COBRA-EN) is used for this project. With an emphasis on reactor physics, this study investigates possible fuel assembly and core designs for civil marine propulsion applications. In particular, it explores the feasibility of using uranium/thorium-rich fuel in a compact, long-life reactor and seek optimal choices and designs of the fuel composition, reactivity control, assembly geometry, and core loading in order to meet the operational needs of a marine propulsion reactor. In this reactor physics and 3D coupled neutronics/thermal-hydraulics study, we attempt to design a civil marine reactor core that fulfills the objective of providing at least 15 effective full-power-years (EFPY) life at 333 MWth. In order to unleash the benefit of thorium in a long life core, the micro-heterogeneous ThO2-UO2 duplex fuel is well-positioned to be utilized in our proposed civil marine core. Unfortunately, A limited number of studies of duplex fuel are available in the public domain, but its use has never been examined in the context of a SBF environment for long-life small modular rector (SMR) core. Therefore, we assumed micro-heterogeneous ThO2-UO2 duplex fuel for our proposed marine core in order to explore its capability. For the proposed civil marine propulsion core design, this study uses 18% U-235 enriched micro-heterogeneous ThO2-UO2 duplex fuel. To provide a basis for comparison we also evaluate the performance of homogeneously mixed 15% U-235 enriched all-UO2 fuel. This research also attempts to design a high power density core with 14 EFPY while satisfying the neutronic and thermal-hydraulics safety constraints. A core with an average power density of 100 MW/m3 has been successfully designed while obtaining a core life of 14 years. The average core power density for this core is increased by ∼50% compared to the reference core design (63 MW/m3 and is equivalent to Sizewell B PWR (101.6 MW/m3 which means capital costs could be significantly reduced and the economic attractiveness of the marine core commensurately improved. In addition, similar to the standard SMR core, a reference core with a power density of 63 MW/m3 has been successfully designed while obtaining a core life of ∼16 years. One of the most important points that can be drawn from these studies is that a duplex fuel lattice needs less burnable absorber than uranium-only fuel to achieve the same poison performance. The higher initial reactivity suppression and relatively smaller reactivity swing of the duplex can make the task of reactivity control through BP design in a thorium-rich core easier. It is also apparent that control rods have greater worth in a duplex core, reducing the control material requirements and thus potentially the cost of the rods. This research also analyzed the feasibility of using thorium-based duplex fuel in different cases and environments to observe whether this fuel consistently exhibit superior performance compared to the UO2 core in both the assembly and whole-core levels. The duplex fuel/core consistently exhibits superior performance in consideration of all the neutronic and TH constraints specified. It can therefore be concluded from this study that the superior performance of the thorium-based micro-heterogeneous ThO2-UO2 duplex fuel provides enhanced confidence that this fuel can be reliably used in high power density and long-life SBF marine propulsion core systems, offering neutronic advantages compared to the all-UO2 fuel. Last, but not least, considering all these factors, duplex fuel can potentially open the avenue for low-enriched uranium (LEU) SBF cores with different configurations. Motivated by growing environmental concerns and anticipated economic pressures, the overall goal of this study is to examine the technological feasibility of expanding the use of nuclear propulsion to civilian maritime shipping and to identify and propose promising candidate core designs.

Advanced computational tools are applied to simulate a nuclear power plant (NPP) control rod assembly ejection (CRE) accident. The impact of these reactivity-initiated accidents (RIAs) on core reactivity behavior, 3D power distribution and stochastic reactivity estimation are evaluated. The three tools used are: the thermal-hydraulic (TH) RELAP5 (R5) code, the neutronic (NK) PARCS (P3D) code, and the coupled version P3D/R5, with specially developed linkage using the environment code MATLAB. This study considers three different-size cores: NPP1 (2772 MWt); NPP2 (530 MWt); and NPP3 (1061 MWt). The three cores have the same general design and control rod assembly (CRA) positions, and the ejected CRA has similar worth and at the same rod ejection pace. The CRE is assessed under both hot zero power (HZP) and hot full power (HFP) conditions. The analyses indicate that RIA modeling and simulation should be carried out through a systematic coding and configuration approaches, otherwise the results will not capture the true transient behavior of the core under analysis. The simulation of one code depends on the appropriate configuration of parameters generated by the other code and on the correct determination of the TH/NK mapping weight factors for the various mesh regions in each of the models. From the design point of view, the standalone codes predict milder magnitude of power and reactivity increase compared to the coupled P3D/R5 simulation. The magnitudes of reduced peak power and reactivity become larger as the core size shrinks. The HFP simulation shows that the three NPPs have the same transient peak value, but the post-transient steady power is lower for a smaller core. The HZP analysis indicates that the transient peak is lower for the smaller core, but the post-transient power occurs at the same level. The three-dimensional (3D) power distributions are different among the HFP and HZP cases, but do not depend on the size of the core. The results indicate: i) HFP: core power increases in the area surrounding the ejected rod/bank assembly, and this increase becomes lower as the NPPs shrinks however, the power is well-distributed after the transient; and ii) HZP: the area surrounding the CRA stays hotter, but the 3D peak assembly factor becomes lower, during and after the transients, as the NPPs shrinks. These features confirm that the smaller cores yield a safer response to a given inserted reactivity compared to larger cores. A stochastic extended Kalman filter (EKF) algorithm is implemented to estimate the reactivity based on the reactor power profile, after the addition of random noise. The inverse point kinetics (IPK) deterministic method is also implemented and the results of the application of EKF and IPK are compared to the P3D/R5 simulation. The following sophisticated strategies made the EKF algorithm robust and accurate: the system is modeled by a set of continuous time nonlinear stochastic differential equations; the code uses a time step directly based on the power measured and applies that to the model for online discretization and linearization; filter tuning goes automatically up from the first time step; and the state noise covariance matrix is updated online at each time step. It was found that the IPK reactivity has higher noise content compared to the EKF reactivity for all cases. Thus, the EKF presents superior and more accurate results. Furthermore, under a small reactivity insertion, the IPK reactivity varies widely from positive to negative values: this variation is not observed within the EKF. A sensitivity analysis for three distinct standard deviation (SD) noise measurements suggests that EKF is superior to IPK method, independent of the noise load magnitude. As the noise content increases, the error between the IPK and P3D/R5 reactivity also increases. A sensitivity analysis for five distinct carry-over effects of different random noise loads indicates that the random addition of different noise loads to the reactor power does not change the overall performance of both algorithms. / Este trabalho aplica métodos computacionais avançados para simular a ejeção de barras de controle (CRE) em uma planta térmica nuclear (NPP). São avaliados o impacto da ocorrência de acidentes iniciados por reatividade (RIAs) na reatividade total, na distribuição da potência em três dimensões (3D) e na determinação da reatividade. As ferramentas utilizadas são: o código termo-hidráulico (TH) RELAP5 (R5), o código neutrônico (NK) PARCS (P3D), a versão acoplada P3D/R5, e o ambiente computacional MATLAB. Este estudo considera três reatores nucleares de diferentes tamanhos: NPP1 (2772 MWT); NPP2 (530 MWt); e NPP3 (1061 MWt). Os três núcleos possuem projeto similar e idêntica posição dos grupos das barras de controle (CRA), além do mesmo valor de reatividade diferencial das CRA ejetadas e idêntica velocidade de ejeção. A ocorrência da CRE é avaliada sob condições de hot zero power (HZP) e de hot full power (HFP). As análises indicam que a modelagem e a simulação de RIAs devem ser realizadas sistematicamente, caso contrário os resultados não irão refletir o comportamento em regime transitório do núcleo. A simulação de um modelo em um código depende da apropriada configuração de parâmetros gerados pelo outro código e da determinação adequada do mapeamento TH/NK para as várias malhas dos modelos. Do ponto de vista de projeto, a utilização de códigos independentes resulta em cálculos de potência e reatividade conservadores em comparação com os resultados utilizando-se P3D/R5. Os picos de potência e de reatividade são menores à medida que o núcleo encolhe. A simulação em condições de HFP resulta em valores de pico de potência similares durante transitório para as três NPPs, mas a potência de pós-transitórios é menor para o menor núcleo. A análise em condições de HZP também indica que o valor máximo durante o transitório é menor para o menor núcleo, mas o pós-transitórios ocorre aos mesmos níveis de potência das demais NPPS. A distribuição de potência em 3D também apresenta resultados distintos para condições de HFP e HZP, mas tais resultados são independentes do tamanho do núcleo: i) HFP: há um aumento da potência do núcleo em torno da CRE, mas tal comportamento diminui para núcleos menores - no entanto, a potência é bem distribuída após o transitório; e ii) HZP: há aumento de potência na área do CRE, mas o pico de potência em 3D é menor durante e depois dos transitórios para núcleos menores. Tais características indicam que os núcleos menores respondem de forma mais segura quando da inserção de reatividade em comparação a reatores de maiores dimensões. O método estocástico de filtragem de Kalman estendido (EKF) foi codificado para estimar a reatividade com base no perfil de potência da NPP, após a adição de ruído aleatório. O método determinístico da cinética pontual inversa (IPK) também foi implementado e os resultados da aplicação dos algoritmos do EKF e IPK foram comparados com os resultados da simulação do P3D/R5. As seguintes estratégias, implementadas neste trabalho, possibilitaram a aplicação robusta e precisa do EKF: o sistema foi modelado por um conjunto de equações diferenciais não-lineares estocásticas de tempo contínuo; o algoritmo obtém o passo de tempo diretamente da potência medida e aplica-o ao modelo para a discretização e linearização online; o ajuste do filtro ocorre automaticamente a partir do primeiro passo de tempo; e a matriz de covariância do ruído no estado é atualizada online. Verificou-se que a reatividade calculada pelo método IPK possui maior nível de ruído quando comparada ao EKF para todos os casos estudados. Portanto, o EKF apresenta resultados superiores e mais precisos. Além disso, sob uma pequena inserção de reatividade, a reatividade calculada pelo método IPK varia consideravelmente de valores positivos para negativos: esta variação não é observada com o EKF. Uma análise de sensibilidade para três desvios padrão (SD) sugere que o algoritmo EKF é superior ao método IPK, independente da magnitude do ruído. Com o aumento da magnitude do ruído, o erro entre as reatividades calculadas pelo IPK e pelo P3D/R5 aumenta. A análise de sensibilidade para cinco ruídos aleatórios indica que a adição de ruído na potência do reator não altera o desempenho global de ambos os algoritmos.

La conception du coeur d’un réacteur nucléaire est fortement multidisciplinaire (neutronique, thermo-hydraulique, thermomécanique du combustible, physique du cycle, etc.). Le problème est aussi de type multi-objectif (plusieurs performances) à grand nombre de dimensions (plusieurs dizaines de paramètres de conception).Les codes de calculs déterministes utilisés traditionnellement pour la caractérisation des coeurs demandant d’importantes ressources informatiques, l’approche de conception classique rend difficile l’exploration et l’optimisation de nouveaux concepts innovants. Afin de pallier ces difficultés, une nouvelle méthodologie a été développée lors de ces travaux de thèse. Ces travaux sont basés sur la mise en oeuvre et la validation de schémas de calculs neutronique et thermo-hydraulique pour disposer d’un outil de caractérisation d’un coeur de réacteur à neutrons rapides à caloporteur sodium tant du point de vue des performances neutroniques que de son comportement en transitoires accidentels.La méthodologie mise en oeuvre s’appuie sur la construction de modèles de substitution (ou métamodèles) aptes à remplacer la chaîne de calcul neutronique et thermo-hydraulique. Des méthodes mathématiques avancées pour la planification d’expériences, la construction et la validation des métamodèles permettent de remplacer cette chaîne de calcul par des modèles de régression au pouvoir de prédiction élevé.La méthode est appliquée à un concept innovant de coeur à Faible coefficient de Vidange sur un très large domaine d’étude, et à son comportement lors de transitoires thermo-hydrauliques non protégés pouvant amener à des situations incidentelles, voire accidentelles. Des analyses globales de sensibilité permettent d’identifier les paramètres de conception influents sur la conception du coeur et son comportement en transitoire. Des optimisations multicritères conduisent à des nouvelles configurations dont les performances sont parfois significativement améliorées. La validation des résultats produits au cours de ces travaux de thèse démontre la pertinence de la méthode au stade de la préconception d’un coeur de réacteur à neutrons rapides refroidi au sodium. / Nuclear reactor core design is a highly multidisciplinary task where neutronics, thermal-hydraulics, fuel thermo-mechanics and fuel cycle are involved. The problem is moreover multi-objective (several performances) and highly dimensional (several tens of design parameters).As the reference deterministic calculation codes for core characterization require important computing resources, the classical design method is not well suited to investigate and optimize new innovative core concepts. To cope with these difficulties, a new methodology has been developed in this thesis. Our work is based on the development and validation of simplified neutronics and thermal-hydraulics calculation schemes allowing the full characterization of Sodium-cooled Fast Reactor core regarding both neutronics performances and behavior during thermal hydraulic dimensioning transients.The developed methodology uses surrogate models (or metamodels) able to replace the neutronics and thermal-hydraulics calculation chain. Advanced mathematical methods for the design of experiment, building and validation of metamodels allows substituting this calculation chain by regression models with high prediction capabilities.The methodology is applied on a very large design space to a challenging core called CFV (French acronym for low void effect core) with a large gain on the sodium void effect. Global sensitivity analysis leads to identify the significant design parameters on the core design and its behavior during unprotected transient which can lead to severe accidents. Multi-objective optimizations lead to alternative core configurations with significantly improved performances. Validation results demonstrate the relevance of the methodology at the predesign stage of a Sodium-cooled Fast Reactor core.