The purpose of this thesis is to examine the accuracy of the RAPID (Real-Time Analysis for Particle Transport and In-situ Detection) eigenvalue algorithm based on a few problems from the ICSBEP (International Criticality Safety Benchmark Evaluation Project) Handbook. RAPID is developed based on the MRT (Multi-Stage Response-Function Transport) methodology and it uses the fission matrix (FM) method for performing eigenvalue calculations. RAPID has already been benchmarked based on several real-world problems including spent fuel pools and casks, and reactor cores. This thesis examines the accuracy of the RAPID eigenvalue algorithm for modeling the physics of problems with unique geometric configurations. Four problems were selected from the ICSBEP Handbook; these problems differ by their unique configurations which can effectively examine the capability of the RAPID code system. For each problem, a reference Serpent Monte Carlo calculation has been performed. Using the same Serpent model in the pRAPID (pre- and post-processing for RAPID) utility code, a series of fixed-source Serpent calculations are performed to determine spatially-dependent FM coefficients. RAPID calculations are performed using these FM coefficients to obtain the axially-dependent, pin-wise fission density distribution and system eigenvalue for each problem. It is demonstrated that the eigenvalues calculated by RAPID and Serpent agree with the experimental data within the given experimental uncertainty. Further, the detailed 3-D pin-wise fission density distribution obtained by RAPID agrees with the reference prediction by Serpent which itself has converged to less than 1% weighted uncertainty. While achieving accurate results, RAPID calculations are significantly faster than the reference Serpent calculations, with a calculation time speed-up of between 4x and 34x demonstrated in this thesis. In addition to examining the accuracy of the RAPID algorithm, this thesis provides useful information on the use of the FM method for simulation of nuclear systems. / Master of Science / In the modeling and simulation of nuclear systems, two parameters are of key importance: the system eigenvalue and the fission distribution. The system eigenvalue, known as kef f , is the ratio of neutron production from fission in the current neutron generation compared with the absorption and leakage of neutrons from the system in the previous neutron generation. When this ratio is equal to one, the system is critical and is a self-sustaining chain reaction. Knowledge of the fission distribution is important in the nuclear power industry, as it enables engineers to determine the best reactor core assembly configuration to maintain an even power distribution. Several methods have been developed over the years to effectively solve for a nuclear systems fission distribution and system eigenvalue. Aspects of both Monte Carlo and deterministic transport methods have been combined into RAPID’s MRT methodology. It is capable of accurately determining the system eigenvalue and fission distribution in real time. This thesis examines the accuracy of the RAPID algorithm using four unique problems from the ICSBEP handbook. These problems help us to test the limits of the FM method in RAPID through the modeling of small, unique geometric configurations not seen in large, uniformly configured power reactor cores and spent fuel pools. For comparison, each problem is modeled using the Serpent Monte Carlo code, an accurate code meant to serve as the industry standard for determination of the fission distribution of each problem. This model is then used to generate a set of FM coefficients for use in RAPID calculations. It is demonstrated that the eigenvalues calculated by RAPID and Serpent agree with the experimental data within the given experimental uncertainty. The fission distribution obtained by RAPID is also in agreement with the Serpent reference model. Finally, the RAPID eigenvalue calculation is significantly faster than the corresponding Serpent reference model, with speed-ups ranging from 4x to 34x demonstrated.

A análise de segurança de criticalidade nuclear é uma atividade mandatória sob o ponto de vista de licenciamento de uma instalação que manipula qualquer quantidade de material físsil. Este trabalho apresenta uma metodologia de análise para uma instalação fabril que processa e estoca ligas de urânio enriquecido. Trata-se da verificação da instalação para que se evite qualquer evento de acidente nuclear, seja por um equipamento inseguro, seja por um arranjo inseguro dos materiais. Todo o ciclo do urânio, principalmente as instalações envolvidas na fabricação do combustível nuclear, é avaliado quanto à segurança contra a criticalidade nuclear. A disponibilidade de informações relacionadas à segurança das instalações para combustíveis de reatores de potência comerciais (PWR e BWR, das siglas em inglês para reator de água pressurizada e reator de água fervente, respectivamente) utilizando o dióxido de urânio (UO2) com baixo nível de enriquecimento são amplamente compartilhadas. No entanto, informações sobre parâmetros seguros de criticalidade nuclear voltadas para combustíveis à base de ligas de urânio com nível de enriquecimento médio (até 20%) são raras na literatura. Dessa forma, o trabalho proposto visa suprir essa carência ao desenvolver um método de análise de criticalidade voltada para uma instalação destinada à fabricação de combustível nuclear, utilizando ligas metálicas urânio com 20% de enriquecimento. / Nuclear Criticality Safety analysis is a mandatory licensing activity for a facility that handles a certain amount of fissile material. This work presents an analysis methodology for a plant which processes and stores uranium alloys enriched. It is the verification of the facility in order to avoid any nuclear accident event, either by unsafe equipament or by an unsafe arrangement of materials. The whole uranium cycle, mainly the facilities involved in manufacturing of nuclear fuel is evaluated for safety against nuclear criticality. The availability of information related to facilities safety for fuel of commercial power reactors facilites (PWR and BWR, Pressurized Water Reactor and Boiling Water Reactor respectively) using the mixed oxide of uranium (UO2) with low enrichment level are widely shared. However, information of safe parameters focused on the nuclear criticality of uranium alloys based fuels with average enrichment level (up to 20%) are scarse in the literature. Thus, the proposed work aims to fill this need by developing a criticality analysis method focused on a facility dedicated to the manufacture of nuclear fuel using uranium alloys with 20% degree of enrichment.

A análise de segurança de criticalidade nuclear é uma atividade mandatória sob o ponto de vista de licenciamento de uma instalação que manipula qualquer quantidade de material físsil. Este trabalho apresenta uma metodologia de análise para uma instalação fabril que processa e estoca ligas de urânio enriquecido. Trata-se da verificação da instalação para que se evite qualquer evento de acidente nuclear, seja por um equipamento inseguro, seja por um arranjo inseguro dos materiais. Todo o ciclo do urânio, principalmente as instalações envolvidas na fabricação do combustível nuclear, é avaliado quanto à segurança contra a criticalidade nuclear. A disponibilidade de informações relacionadas à segurança das instalações para combustíveis de reatores de potência comerciais (PWR e BWR, das siglas em inglês para reator de água pressurizada e reator de água fervente, respectivamente) utilizando o dióxido de urânio (UO2) com baixo nível de enriquecimento são amplamente compartilhadas. No entanto, informações sobre parâmetros seguros de criticalidade nuclear voltadas para combustíveis à base de ligas de urânio com nível de enriquecimento médio (até 20%) são raras na literatura. Dessa forma, o trabalho proposto visa suprir essa carência ao desenvolver um método de análise de criticalidade voltada para uma instalação destinada à fabricação de combustível nuclear, utilizando ligas metálicas urânio com 20% de enriquecimento. / Nuclear Criticality Safety analysis is a mandatory licensing activity for a facility that handles a certain amount of fissile material. This work presents an analysis methodology for a plant which processes and stores uranium alloys enriched. It is the verification of the facility in order to avoid any nuclear accident event, either by unsafe equipament or by an unsafe arrangement of materials. The whole uranium cycle, mainly the facilities involved in manufacturing of nuclear fuel is evaluated for safety against nuclear criticality. The availability of information related to facilities safety for fuel of commercial power reactors facilites (PWR and BWR, Pressurized Water Reactor and Boiling Water Reactor respectively) using the mixed oxide of uranium (UO2) with low enrichment level are widely shared. However, information of safe parameters focused on the nuclear criticality of uranium alloys based fuels with average enrichment level (up to 20%) are scarse in the literature. Thus, the proposed work aims to fill this need by developing a criticality analysis method focused on a facility dedicated to the manufacture of nuclear fuel using uranium alloys with 20% degree of enrichment.

En règle générale, les études de sûreté-criticité concernant les combustibles usés stockés, transportés ou retraités sont très conservatives et considèrent ce combustible comme neuf donc le plus réactif possible. Le « Crédit Burn-up » (CBU) est la prise en compte de l’antiréactivité du combustible irradié par rapport au combustible neuf. Une méthodologie CBU rigoureuse, développée par le CEA en collaboration avec AREVA-NC a récemment été validée et réévaluée pour les combustibles REP-UOx. Cependant, 22 réacteurs sur les 58 que compte la France utilisent également du combustible MOx. De plus en plus d’assemblages MOx irradiés doivent donc être entreposés et transportés, ce qui conduit les industriels à s’intéresser à la prise en compte du CBU pour ces applications, dans le but de pouvoir gagner des marges en terme de dimensionnement des installations. Des publications récentes et les travaux du Groupe de Travail Français sur le CBU ont souligné l’importance de la prise en compte des 15 produits de fission stables et non volatiles les plus absorbants qui sont à l’origine de la moitié de l’antiréactivité totale apportée dans les combustibles REP-MOx. C’est pourquoi, dans le but de garantir la sous-criticité de la configuration étudiée suivant les dispositions règlementaires relatives à la sûreté des installations, les biais de calcul affectant leur bilan-matière et leur effet individuel en réactivité doivent également être pris en considération dans les études de sûreté-criticité s’appuyant sur des calculs de criticité. Dans ce contexte, une revue bibliographique exhaustive a permis d’identifier les particularités des combustibles REP-MOx et une démarche rigoureuse a été suivie afin de proposer une méthodologie CBU adaptée à ces combustibles validée et physiquement représentative, permettant de prendre en compte les produits de fission et permettant d’évaluer les biais liés au bilan-matière et à l’antiréactivité des isotopes considérés. Cette démarche s’est articulée autour des études suivantes : • détermination de facteurs correctifs isotopiques permettant de garantir le conservatisme du calcul de criticité sur la base de la qualification du formulaire d’évolution DARWIN-2.3 pour les applications REP-MOx et d’une analyse des données nucléaires des produits de fission métalliques afin de déterminer l’impact des incertitudes associées sur le calcul de leur bilan matière ; • évaluation de l’antiréactivité individuelle des produits de fission sur la base des résultats d’interprétation des expériences d’oscillation des programmes CBU et MAESTRO, réalisés dans le réacteur expérimental MINERVE à Cadarache, avec le formulaire dédié PIMS développé au SPRC/LEPh avec mise à jour des schémas de calcul pour la criticité ; • élaboration de matrices de covariances réalistes associées à la capture de deux des principaux produits de fission du CBU REP-MOx : 149Sm et le 103Rh associées à l’évaluation JEFF-3.1.1 ; • détermination des biais et incertitudes « a posteriori » dus aux données nucléaires des actinides et produits de fission considérés pour deux applications industrielles (piscine d’entreposage et château de transport) par une étude de transposition réalisée avec l’outil RIB, développé au SPRC/LECy, qui a bénéficié à cette occasion de développements spécifiques et de mises à jour des données utilisées (importation des données de covariance issues de la bibliothèque COMAC V0 associée à JEFF-3.1.1 pour les isotopes 235,238U, 238,239,240,241,242Pu, 241Am et 155Gd et prise en compte des corrélations inter-réactions pour un même isotope). • évaluation de la méthodologie proposée pour deux applications industrielles (piscine d’entreposage et château de transport), démonstration de son intérêt et de sa robustesse. / Considering spent fuel management (storage, transport and reprocessing), the approach using « fresh fuel assump-tion » in criticality-safety studies results in a significant conservatism in the calculated value of the system reactivity.The concept of Burnup Credit (BUC) consists in considering the reduction of the spent fuel reactivity due to its burnup.A careful BUC methodology, developed by CEA in association with AREVA-NC was recently validated and writtenup for PWR-UOx fuels. However, 22 of 58 french reactors use MOx fuel, so more and more irradiated MOx fuelshave to be stored and transported. As a result, why industrial partners are interested in this concept is because takinginto account this BUC concept would enable for example a load increase in several fuel cycle devices. Recent publi-cations and discussions within the French BUC Working Group highlight the current interest of the BUC concept inPWR-MOx spent fuel industrial applications. In this case of PWR-MOx fuel, studies show in particular that the 15FPs selected thanks to their properties (absorbing, stable, non-gaseous) are responsible for more than a half of the totalreactivity credit and 80% of the FPs credit. That is why, in order to get a conservative and physically realistic valueof the application keff and meet the Upper Safety Limit constraint, calculation biases on these 15 FPs inventory andindividual reactivity worth should be considered in a criticality-safety approach. All of this work is supported by the use of the CEA reference calculation tools : the deterministic code APOLLO-2.8and the probabilistic code TRIPOLI-4 used by the CRISTAL V2 criticality-safety package, the DARWIN-2.3 packagefor fuel cycle applications, the JEFF-3.1.1 nuclear data library and the Integral Experiment Methodology based on thestatistical adjustment method of the nuclear data and the integral experiment representativity.The feedback on the nuclear data of the oscillation programmes BUC and MAESTRO allows to halve the prioruncertainties linked to 149Sm and 103Rh capture cross sections. The application of the developed methodology,benefiting from the CEA dedicated experimental programmes quality and better physically justified to twoapplications, representative of fuel storage and transport, shows that the introduced conservatisms represents40 % of the total Burnup Credit. On top of that, the two configurations results comparison shows that theevaluated BUC is independent from the considered application and proves the calculation route robustness.