Tecnécio-99m (99mTc), o produto de decaimento do molibdênio-99 (99Mo), é um dos radioisótopos mais utilizados na medicina nuclear, abrangendo cerca de 80% de todos os procedimentos de radiodiagnóstico médico pelo mundo. Atualmente o Brasil necessita de uma quantidade de aproximadamente 450 Ci de 99Mo por semana. Devido à crise e à escassez em seu fornecimento que vem sendo observada no cenário mundial desde 2008, o IPEN decidiu desenvolver um projeto próprio para produção de 99Mo através da fissão do urânio-235. O objetivo deste trabalho de dissertação foi desenvolver cálculos neutrônicos e temo-hidráulicos para avaliar a segurança operacional de um dispositivo para produção de 99Mo a ser irradiado no núcleo do reator IEA-R1. Neste dispositivo serão alojados dez alvos do tipo dispersão de UAlx-Al com baixo enriquecimento de urânio (LEU) e densidade de 2,889 gU/cm³. Para o cálculo neutrônico foram utilizados os programas computacionais HAMMER-TECHNION e CITATION e as temperaturas máximas atingidas nos alvos foram calculadas com o código MTRCR-IEAR1. Os cálculos demonstram que a irradiação do dispositivo deverá ocorrer sem consequências adversas à operação do reator. A quantidade total de 99Mo foi calculada com o programa SCALE e considerando que o tempo necessário para o processamento químico e recuperação do 99Mo será de cinco dias após a irradiação, teremos disponível para distribuição uma atividade de 99Mo de 176 Ci para 3 dias de irradiação, 236 Ci para 5 dias de irradiação e 272 Ci para 7 dias de irradiação dos alvos. / Technetium-99m (99mTc), the product of radioactive decay of molybdenum-99 (99Mo), is one of the most widely used radioisotope in nuclear medicine, covering approximately 80% of all radiodiagnosis procedures in the world. Nowadays, Brazil requires an amount of about 450 Ci of 99Mo per week. Due to the crisis and the shortage of 99Mo supply chain that has been observed on the world since 2008, IPEN/CNEN-SP decided to develop a project to produce 99Mo through fission of uranium-235. The objective of this dissertation was the development of neutronic and thermal-hydraulic calculations to evaluate the operational safety of a device for 99Mo production to be irradiated in the IEA-R1 reactor core at 5 MW. In this device will be placed ten targets of UAlx-Al dispersion fuel with low enriched uranium (LEU) and density of 2.889 gU/cm³. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION and CITATION and the maximum temperatures reached in the targets were calculated with the code MTRCR-IEAR1. The analysis demonstrated that the device irradiation will occur without adverse consequences to the operation of the reactor. The total amount of 99Mo was calculated with the program SCALE and considering that the time needed for the chemical processing and recovering of the 99Mo will be five days after the irradiation, we have that the 99Mo activity available for distribution will be 176 Ci for 3 days of irradiation, 236 Ci for 5 days of irradiation and 272 Ci for 7 days of targets irradiation.

Depuis quelques années, les codes de calculs Monte Carlo évoluant qui couplent un code Monte Carlo, pour simuler le transport des neutrons, à un solveur déterministe, qui traite l'évolution des milieux dû à l'irradiation sous le flux neutronique, sont apparus. Ces codes permettent de résoudre les équations de Boltzmann et de Bateman dans des configurations complexes en trois dimensions et de s'affranchir des hypothèses multi-groupes utilisées par les solveurs déterministes. En contrepartie, l'utilisation du code Monte Carlo à chaque pas de temps requiert un temps de calcul prohibitif.Dans ce manuscrit, nous présentons une méthodologie originale évitant la répétition des simulations Monte Carlo coûteuses en temps et en les remplaçant par des perturbations. En effet, les différentes simulations Monte Carlo successives peuvent être vues comme des perturbations des concentrations isotopiques de la première simulation. Dans une première partie, nous présenterons donc cette méthode, ainsi que la méthode de perturbation utilisée: l'échantillonnage corrélé. Dans un second temps, nous mettrons en place un modèle théorique permettant d'étudier les caractéristiques de la méthode des échantillons corrélés afin de comprendre ses effets durant les calculs en évolution. Enfin, dans la troisième partie nous discuterons de l'implémentation de cette méthode dans TRIPOLI-4® en apportant quelques précisions sur le schéma de calcul qui apportera une accélération importante aux calculs en évolution. Nous commencerons par valider et optimiser le schéma de perturbation à travers l'étude de l'évolution d'une cellule de combustible de type REP. Puis cette technique sera utilisée sur un calcul d'un assemblage de type REP en début de cycle. Après avoir validé la méthode avec un calcul de référence, nous montrerons qu'elle peut accélérer les codes Monte Carlo évoluant standard de presque un ordre de grandeur. / For several years, Monte Carlo burnup/depletion codes have appeared, which couple Monte Carlo codes to simulate the neutron transport to deterministic methods, which handle the medium depletion due to the neutron flux. Solving Boltzmann and Bateman equations in such a way allows to track fine 3-dimensional effects and to get rid of multi-group hypotheses done by deterministic solvers. The counterpart is the prohibitive calculation time due to the Monte Carlo solver called at each time step.In this document we present an original methodology to avoid the repetitive and time-expensive Monte Carlo simulations, and to replace them by perturbation calculations: indeed the different burnup steps may be seen as perturbations of the isotopic concentration of an initial Monte Carlo simulation. In a first time we will present this method, and provide details on the perturbative technique used, namely the correlated sampling. In a second time we develop a theoretical model to study the features of the correlated sampling method to understand its effects on depletion calculations. In a third time the implementation of this method in the TRIPOLI-4® code will be discussed, as well as the precise calculation scheme a meme to bring important speed-up of the depletion calculation. We will begin to validate and optimize the perturbed depletion scheme with the calculation of a REP-like fuel cell depletion. Then this technique will be used to calculate the depletion of a REP-like assembly, studied at beginning of its cycle. After having validated the method with a reference calculation we will show that it can speed-up by nearly an order of magnitude standard Monte-Carlo depletion codes.

Le travail de cette thèse porte sur le développement de modèles neutroniques innovants pour le couplage avec la thermohydraulique, associant précision et temps de calcul raisonnable. Un des cas d'application principaux étant le réacteur à sel fondu, à spectre neutronique rapide et en cycle thorium MSFR (Molten Salt Fast Reactor), réacteur de 4ème génération à combustible liquide circulant, la prise en compte du mouvement des précurseurs de neutrons retardés et des phénomènes associés est nécessaire. Les études de conception de ce type de réacteur ont été le point de départ de ces développements, via le besoin d'une représentation multiphysique adaptée pour l'obtention d'une image globale et la réalisation d'études de transitoire.Dans un premier temps un couplage stationnaire a été développé, associant un modèle neutronique basé sur une approche stochastique, et un code de CFD (Computational Fluid Dynamics) résolvant les équations de Navier Stokes des écoulements turbulents ainsi que le transport des précurseurs de neutrons retardés. Ce modèle neutronique intègre l'effet lié au transport de ces précurseurs par une reconstruction de la gerbe prompte qu'ils génèrent. Cette approche dite par gerbe considère le réacteur critique comme un système sous-critique prompt amplifiant la source de neutrons retardés.Dans un second temps, un modèle neutronique basé sur une version temporelle des matrices de fission (Transient Fission Matrix ou TFM) a été développé afin de réaliser des études de transitoires. Le modèle TFM permet, en un premier calcul des matrices avec un code stochastique (MCNP, SERPENT), de réaliser une caractérisation de l'ensemble de la réponse neutronique spatiale et temporelle du réacteur avec une précision proche de celle du calcul Monte Carlo. Dans un second temps cette information est utilisée pour les calculs de transitoires tout en gardant un temps de calcul réduit. Le modèle TFM, utilisable pour différents types de systèmes, permet également le calcul de paramètres cinétiques effectifs tels que la fraction effective de neutrons retardés ou le temps de génération effectif. Différents cas d'application ont été utilisés afin de vérifier et d'illustrer cette approche sur des calculs temporels ou de paramètres cinétiques.Enfin le modèle TFM a été implémenté dans le code de thermohydraulique OpenFOAM. Ce couplage a été testé sur un benchmark numérique à géométrie simplifiée, puis des calculs sur le MSFR ont été réalisés, pour des transitoires normaux (suivis de charge) ou accidentels (insertions de réactivité, sur-refroidissements). / In this PhD thesis, we describe the development of innovative neutronic models for their coupling with thermalhydraulics such that they combine precision and reasonable computational times. One of the main cases where this method is applied is the Molten Salt Fast Reactor (MSFR) whose combines a fast neutron spectrum with a thorium cycle. In this fourth generation reactor, the motion of the delayed neutron precursors and the associated phenomena have to be taken into account due to the liquid fuel circulation. The starting point for these developments was the preliminary design of this type of system where a dedicated multi-physical representation was needed to study the reactor performance in steady and transient conditions.As a first step, a stationary coupling was developed. A neutronic model based on a stochastic approach was associated to a CFD (Computational Fluid Dynamics) code to solve the Navier Stokes equations for turbulent flows and the transport of the delayed neutron precursors. The impact of this precursor motion is taken into account by reconstructing the prompt shower that they generate. This approach, called by shower, views the critical reactor as a prompt subcritical reactor that amplifies a source of delayed neutrons.A second step consisted in developing a neutronic model based on a time dependent version of the fission matrices (Transient Fission Matrix or TFM) so as to enable reactor transient studies. With the TFM model, an initial computation of the matrices with a stochastic code (MCNP, SERPENT) allows the characterization of the global spatial and time dependent neutronic response of the reactor with a precision close to that of a Monte Carlo calculation. The information thus obtained is then used to calculate transients, while retaining the advantage of reduced computational time. The TFM model, which can be used for various system concepts, also allows the evaluation of effective kinetic parameters such as the effective fraction of delayed neutrons or the effective generation time. The method was applied to various cases in order to verify it and demonstrate the approach for time dependent or kinetic parameter calculations.Finally, the TFM model was integrated in the OpenFOAM thermalhydraulic code. The coupling was first tested on a simple geometry numerical benchmark. Subsequently, it was applied to the MSFR to calculate normal (load-following) and accidental (reactivity insertion, over-cooling) transients.

Tecnécio-99m (99mTc), o produto de decaimento do molibdênio-99 (99Mo), é um dos radioisótopos mais utilizados na medicina nuclear, abrangendo cerca de 80% de todos os procedimentos de radiodiagnóstico médico pelo mundo. Atualmente o Brasil necessita de uma quantidade de aproximadamente 450 Ci de 99Mo por semana. Devido à crise e à escassez em seu fornecimento que vem sendo observada no cenário mundial desde 2008, o IPEN decidiu desenvolver um projeto próprio para produção de 99Mo através da fissão do urânio-235. O objetivo deste trabalho de dissertação foi desenvolver cálculos neutrônicos e temo-hidráulicos para avaliar a segurança operacional de um dispositivo para produção de 99Mo a ser irradiado no núcleo do reator IEA-R1. Neste dispositivo serão alojados dez alvos do tipo dispersão de UAlx-Al com baixo enriquecimento de urânio (LEU) e densidade de 2,889 gU/cm³. Para o cálculo neutrônico foram utilizados os programas computacionais HAMMER-TECHNION e CITATION e as temperaturas máximas atingidas nos alvos foram calculadas com o código MTRCR-IEAR1. Os cálculos demonstram que a irradiação do dispositivo deverá ocorrer sem consequências adversas à operação do reator. A quantidade total de 99Mo foi calculada com o programa SCALE e considerando que o tempo necessário para o processamento químico e recuperação do 99Mo será de cinco dias após a irradiação, teremos disponível para distribuição uma atividade de 99Mo de 176 Ci para 3 dias de irradiação, 236 Ci para 5 dias de irradiação e 272 Ci para 7 dias de irradiação dos alvos. / Technetium-99m (99mTc), the product of radioactive decay of molybdenum-99 (99Mo), is one of the most widely used radioisotope in nuclear medicine, covering approximately 80% of all radiodiagnosis procedures in the world. Nowadays, Brazil requires an amount of about 450 Ci of 99Mo per week. Due to the crisis and the shortage of 99Mo supply chain that has been observed on the world since 2008, IPEN/CNEN-SP decided to develop a project to produce 99Mo through fission of uranium-235. The objective of this dissertation was the development of neutronic and thermal-hydraulic calculations to evaluate the operational safety of a device for 99Mo production to be irradiated in the IEA-R1 reactor core at 5 MW. In this device will be placed ten targets of UAlx-Al dispersion fuel with low enriched uranium (LEU) and density of 2.889 gU/cm³. For the neutronic calculations were utilized the computer codes HAMMER-TECHNION and CITATION and the maximum temperatures reached in the targets were calculated with the code MTRCR-IEAR1. The analysis demonstrated that the device irradiation will occur without adverse consequences to the operation of the reactor. The total amount of 99Mo was calculated with the program SCALE and considering that the time needed for the chemical processing and recovering of the 99Mo will be five days after the irradiation, we have that the 99Mo activity available for distribution will be 176 Ci for 3 days of irradiation, 236 Ci for 5 days of irradiation and 272 Ci for 7 days of targets irradiation.

Over the last two decades, there has been widespread international interest in the development of the molten salt reactor concept due to its passive safety, high coolant boiling temperature, low operational pressure, high thermal efficiency, and ease of breeding. Terrestrial Energy Incorporated (TEI) is developing a thermal-spectrum converter type molten salt reactor, called the Integral Molten Salt Reactor (IMSR-400) to be built by 2030. A physics experiment is needed in order to validate the theoretical predictions of the temperature reactivity coefficients of the IMSR-400. This thesis will determine the feasibility of conducting a subcritical experiment, utilizing a Deuterium-Deuterium Fusion Neutron Source (DD). / Thesis / Master of Science (MSc)

Dans le cadre des études neutroniques d'innovation sur les cœurs de Réacteurs à Eau légère Pressurisée (REP) de 3ème génération, la recherche de l'économie des ressources naturelles est fondamentale afin de pérenniser la filière électronucléaire. Cette étude consiste à rechercher l'économie des ressources par la conception de cœurs de réacteurs à hauts facteurs de conversion, s'appuyant sur des combustibles oxydes mixtes à base de thorium / uranium / plutonium, ainsi que d'élaborer des stratégies de multi-recyclage des matières fissiles (plutonium et 233U). La démarche s'est déroulée en quatre étapes. Deux domaines d'étude ont tout d'abord été identifiés, le premier concerne les faibles rapports de modération (RM) et un combustible ThPuO2, le second les RM standards à accrus et un combustible ThUO2. La première voie a conduit à l'étude de Réacteurs Sous-Modérés (RSM) selon les critères de production d'233U accrue et de consommation limitée de plutonium. Deux concepts ont été retenus en particulier, à partir desquels des stratégies de multi-recyclage des matières fissiles ont été élaborées. La production et le recyclage de l'233U exclusivement en RSM limitent l'économie annuelle d'Unat à 30% environ. Il a été mis en évidence que le besoin en plutonium des RSM producteurs d'233U est le facteur limitant. C'est pourquoi un dernier chapitre évalue comment la production d'233U au sein de REP, dès 2020, permet de favoriser la transition vers un cycle symbiotique REP/RSM en relâchant la contrainte sur les inventaires de plutonium. Cette stratégie laisse présager une économie annuelle de l'ordre de 65% d'Unat par rapport à la poursuite du mono-recyclage du MOX en REP. / Within the framework of innovative neutronic conception of Pressurized Light Water Reactors (PWR) of 3rd generation, saving of natural resources is of paramount importance for sustainable nuclear energy production. This study consists in the one hand to design high Conversion Reactors exploiting mixed oxide fuels composed of thorium / uranium / plutonium, and in the other hand, to elaborate multirecycling strategies of both plutonium and 233U, in order to maximize natural resources economy. This study has two main objectives: first the design of High Conversion PWR (HCPWR) with mixed oxide fuels composed of thorium / uranium / plutonium, and secondly the setting up of multirecycling strategies of both plutonium and 233U, to better natural resources economy. The approach took place in four stages. Two ways of introducing thorium into PWR have been identified: the first is with low moderator to fuel volume ratios (MR) and ThPuO2 fuel, and the second is with standard or high MR and ThUO2 fuel. The first way led to the design of under-moderated HCPWR following the criteria of high 233U production and low plutonium consumption. This second step came up with two specific concepts, from which multirecycling strategies have been elaborated. The exclusive production and recycling of 233U inside HCPWR limits the annual economy of natural uranium to approximately 30%. It was brought to light that the strong need in plutonium in the HCPWR dedicated to 233U production is the limiting factor. That is why it was eventually proposed to study how the production of 233U within PWR (with standard MR), from 2020. It was shown that the anticipated production of 233U in dedicated PWR relaxes the constraint on plutonium inventories and favours the transition toward a symbiotic reactor fleet composed of both PWR and HCPWR loaded with thorium fuel. This strategy is more adapted and leads to an annual economy of natural uranium of about 65%.

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

A breed-and-burn molten salt reactor (BBMSR) concept is proposed to address the Generation IV fuel cycle sustainability objective in a once-through cycle with low enrichment and no reprocessing. The BBMSR uses separate fuel and coolant molten salts, with the fuel contained in assemblies of individual tubes that can be shuffled and reclad periodically to enable high burnup. In this dual-salt configuration, the BBMSR may overcome several limitations of previous breed-and-burn (B$\&$B) designs to achieve high uranium utilisation with a simple, passively safe design. A central challenge in design of the BBMSR fuel is balancing the neutronic requirement of large fuel volume fraction for B$\&$B mode with the thermal-hydraulic requirements for safe and economically competitive reactor operation. Natural convection of liquid fuel within the tubes aids heat transfer to the coolant, and a systematic approach is developed to efficiently model this complex effect. Computational fluid dynamics modelling is performed to characterise the unique physics of the system and produce a new heat transfer correlation, which is used alongside established correlations in a numerical model. A design framework is built around this numerical model to iteratively search for the limiting power density of a given fuel and channel geometry, applying several defined temperature and operational constraints. It is found that the trade-offs between power density, core pressure drop, and pumping power are lessened by directing the flow of coolant downwards through the channel. Fuel configurations that satisfy both neutronic and thermal-hydraulic objectives are identified for natural, 5$\%$ enriched, and 20$\%$ enriched uranium feed fuel. B$\&$B operation is achievable in the natural and 5$\%$ enriched versions, with power densities of 73 W/cm$^3$ and 86 W/cm$^3$, and theoretical uranium utilisations of 300 $\mathrm{MWd/kgU_{NAT}}$ and 25.5 $\mathrm{MWd/kgU_{NAT}}$, respectively. Using 20$\%$ enriched feed fuel relaxes neutronic constraints so a wider range of fuel configurations can be considered, but there is a strong inverse correlation between power density and uranium utilisation. The fuel design study demonstrates the flexibility of the BBMSR concept to operate along a spectrum of modes ranging from high fuel utilisation at moderate power density using natural uranium feed fuel, to high power density and moderate utilisation using 20$\%$ uranium enrichment.