La connaissance actuelle de la distribution angulaire de la fission induite par neutrons est limitée à une énergie maximum de 15~MeV, avec de grands écarts autour de 14~MeV. Seulement 238U et 232Th ont été étudiés jusqu'à 100 MeV et un seul jeu de données existe. Nous avons réalisé une expérience à n_TOF au CERN pour mesurer les distributions angulaires de fragments de fission jusqu'à 1~GeV pour les isotopes 232Th, 235U , 238U , 237Np.L'expérience a été réalisée à l'aide d'un dispositif expérimental à base de compteurs à avalanche à plaques parallèles (PPAC). La méthode basée sur la détection des 2 fragments en coïncidence permet d'identifier sans ambiguïté la fission des autres réactions, notamment dans le domaine de spallation. Au-dessous de 10 MeV nos résultats sont cohérents avec les données existantes. Par exemple, dans le cas de 232Th , en dessous de 10 MeV ils montrent clairement la variation d'anisotropie se produisant dans les résonances vibrationnelles (1.6 MeV) correspondant à des états de transition de J et K donnés (spin total et sa projection sur l'axe de fission), et après l'ouverture de la deuxième chance de fission (7 MeV). Ils apportent une meilleure précision autour de la troisième chance de fission (14 MeV). Aux énergies intermédiaires, au-dessus de 20 MeV nous avons constaté une anisotropie significative mais bien inférieure à l'unique résultat antérieur. Notre résultat est en accord avec la systématique en fissilité du système composite et avec un modèle incluant les phénomènes essentiels, en particulier le preéquilibre. Dans le cadre de cette comparaison l'anisotropie plus grande que pour la fission induite par protons s'explique parfaitement. J'ai par ailleurs exploré et simulé les expériences de criticité qui permettent de tester la précision des données nucléaires. La section efficace de fission de 237Np induite par neutrons avait été mesurée sur l'installation n_TOF au CERN. Par rapport aux résultats antérieurs la section efficace de fission n_TOF était apparue plus élevée de 6-7% au-delà du seuil de fission. Pour vérifier la pertinence des données de n_TOF, nous avons simulé une expérience de criticité effectuée à Los Alamos avec une sphère contenant 6 kg de 237Np. Cette sphère est entourée par de l'uranium hautement enrichi en 235U de façon à approcher la criticité avec des neutrons rapides. La simulation prédit un facteur de multiplication keff en meilleur accord avec l'expérience (l'écart de -0.75% est réduit à +0.25%) quand on remplace la section efficace de fission de 237Np des bibliothèques évaluées par celle de n_TOF. Nous avons également exploré d'autres effets pouvant expliquer l'écart qui existait entre la mesure de criticité et sa prédiction par les simulations, en particulier nous avons testé la section inélastique de 235U et la multiplicité de neutrons de fission de 237Np. Dans les 2 cas la modification requise pour réconcilier l'écart de criticité n'est pas en accord avec les mesures. Des mesures de taux de fission dans des flux de neutrons dont le spectre est connu indiquent également que la section de fission du 237Np pourrait être plus grande de 4 à 5% par rapport à ce qui était admis aujourd'hui. / The present knowledge of angular distributions of neutron-induced fission is limited to a maximal energy of 15 MeV, with large discrepancies around 14 MeV. Only 238U and 232Th have been investigated up to 100 MeV in a single experiment. The n_TOF Collaboration performed the fission cross section measurement of several actinides (232Th, 235U, 238U, 234U, 237Np) at the n_TOF facility using an experimental set-up made of Parallel Plate Avalanche Counters (PPAC), extending the energy domain of the incident neutron above hundreds of MeV. The method based on the detection of the 2 fragments in coincidence allowed to clearly disentangle the fission reactions among other types of reactions occurring in the spallation domain. I will show the methods we used to reconstruct the full angular resolution by the tracking of fission fragments. Below 10 MeV our results are consistent with existing data. For example in the case of 232Th, below 10 MeV the results show clearly the variation occurring at the first (1 MeV) and second (7 MeV) chance fission, corresponding to transition states of given J and K (total spin and its projection on the fission axis), and a much more accurate energy dependence at the 3rd chance threshold (14 MeV) has been obtained. In the spallation domain, above 30 MeV we confirm the high anisotropy revealed in 232Th by the single existing data set. I'll discuss the implications of this finding, related to the low anisotropy exhibited in proton-induced fission. I also explore the critical experiments which is valuable checks of nuclear data. The 237Np neutron-induced fission cross section has recently been measured in a large energy range (from eV to GeV) at the n TOF facility at CERN. When compared to previous measurements, the n TOF fission cross section appears to be higher by 5-7 % beyond the fission threshold. To check the relevance of n TOF data, we simulate a criticality experiment performed at Los Alamos with a 6 kg sphere of 237Np. This sphere was surrounded by enriched uranium 235U so as to approach criticality with fast neutrons. The simulation predicts a multiplication factor keff in better agreement with the experiment (the deviation of 750 pcm is reduced to 250 pcm) when we replace the ENDF/B- VII.0 evaluation of the 237Np fission cross section by the n TOF data. We also explore the hypothesis of deficiencies of the inelastic cross section in 235U which has been invoked by some authors to explain the deviation of 750 pcm. The large distortion that should be applied to the inelastic cross sections in order to reconcile the critical experiment with its simulation is incompatible with existing measurements. Also we show that the nubar of 237Np can hardly be incriminated because of the high accuracy of the existing data. Fission rate ratios or averaged fission cross sections measured in several fast neutron fields seem to give contradictory results on the validation of the 237Np cross section but at least one of the benchmark experiments, where the active deposits have been well calibrated for the number of atoms, favors the n TOF data set. These outcomes support the hypothesis of a higher fission cross section of 237Np.

Les programmes d’irradiations technologiques menés dans les réacteurs expérimentaux sont d’une importance cruciale pour le soutien du parc électronucléaire actuel en termes d’étude et d’anticipation du comportement sous irradiation des combustibles et des matériaux de structures. Ces programmes permettent d’améliorer la sûreté des réacteurs actuels et également d’étudier les matériaux pour les nouveaux concepts de réacteurs.Les conditions d’irradiations des matériaux dans les réacteurs expérimentaux doivent être représentatives de celles des réacteurs de puissance. Un des principaux intérêts des réacteurs d'irradiations technologiques (Material Testing Reactors, MTRs) est de pouvoir y mener des irradiations instrumentées en ajustant les paramètres expérimentaux, en particulier le flux neutronique et la température. La maîtrise du paramètre température d’un dispositif irradié dans un réacteur expérimental nécessite la connaissance de l'échauffement nucléaire (terme source) dû au dépôt d'énergie des photons et des neutrons interagissant dans le dispositif. La bonne évaluation de cet échauffement est une donnée clé pour les études thermiques de dimensionnement et de sûreté du dispositif.L'objectif de cette thèse est d'améliorer les méthodes d’évaluation de l'échauffement nucléaire en réacteur. Ce travail consiste en l’élaboration d'un schéma de calcul complet innovant, couplé neutron-photon (permettant d’obtenir la contribution des neutrons, des gamma prompts et des gamma de décroissance), fondé principalement sur le code de transport Monte-Carlo TRIPOLI-4 (à 3-dimensions et à énergie continue). Une validation expérimentale du schéma a été effectuée en s’appuyant sur les mesures de calorimétrie réalisées dans le réacteur OSIRIS (CEA Saclay). Des études de sensibilité ont également été menées pour établir l’impact de différents paramètres sur les calculs d’échauffement nucléaire, dont les données nucléaires. Cela a permis de définir le schéma de calcul définitif pour reproduire au plus près la réalité des irradiations technologiques. Le travail de thèse débouche sur un outil opérationnel et prédictif pour l'estimation de l'échauffement nucléaire répondant aux besoins de l’expérimentation en réacteur de recherche et qui peut être étendu plus largement dans des réacteurs de puissance. / Technological irradiation programs carried out in experimental reactors are crucial for the support of the current nuclear fleet in terms of study and anticipation of the behavior under irradiation of fuels and structural materials. These programs make it possible to improve the safety of the current reactors and also to study materials for the new concepts of reactors.Irradiation conditions of materials in experimental reactors must be representative of those of nuclear power plants (NPPs). One of the main advantages of material testing reactors (MTRs) is to be able to carry out instrumented irradiations by adjusting experimental parameters, in particular the neutron flux and the temperature. The control of the parameter temperature of a device irradiated in an experimental reactor requires the knowledge of the nuclear heating (source term) due to the deposition of energy of the photons and the neutrons interacting in the device. A relevant evaluation of this heating is a key data for the thermal studies of design and safety of devices. The objective of this thesis is to improve the methods of the evaluation of nuclear heating in reactors. This work consists of the development of an innovating and complete coupled neutron-photon calculation scheme (allowing to obtain the contribution of neutrons, prompt gamma and decay gamma), mainly based on the TRIPOLI-4 Monte Carlo transport code (with 3-dimensions and continuous energy). An experimental validation of the calculation scheme has been performed, based on calorimetry measurements carried out in the OSIRIS reactor (CEA Saclay). Sensitivity studies have been undertaken to establish the impact of various parameters on nuclear heating calculations (in particular nuclear data) and to fix the final calculation scheme to be closer to the technological irradiation aspects. The thesis work leads to an operational and predictive tool for the nuclear heating estimation, meeting the experimentation needs of research reactors and can be extended more generally to NPPs.

Dans cette thèse, nous présentons une étude rigoureuse des barres d'erreurs et des sensibilités de paramètres neutroniques (tels le keff) aux données nucléaires de base utilisées pour les calculer. Notre étude commence au niveau fondamental, i.e. les fichiers de données ENDF et leurs incertitudes, fournies sous la forme de matrices de variance/covariance, et leur traitement. Lorsqu'un calcul méthodique et consistant des sensibilités est consenti, nous montrons qu'une approche déterministe utilisant des formalismes bien connus est suffisante pour propager les incertitudes des bases de données avec un niveau de précision équivalent à celui des meilleurs outils disponibles sur le marché, comme les codes Monte-Carlo de référence. En appliquant notre méthodologie à trois exercices proposés par l'OCDE, dans le cadre des Benchmarks UACSA, nous donnons des informations, que nous espérons utiles, sur les processus physiques et les hypothèses sous-jacents aux formalismes déterministes utilisés dans cette étude. / This dissertation presents a comprehensive study of sensitivity/uncertainty analysis for reactor performance parameters (e.g. the k-effective) to the base nuclear data from which they are computed. The analysis starts at the fundamental step, the Evaluated Nuclear Data File and the uncertainties inherently associated with the data they contain, available in the form of variance/covariance matrices. We show that when a methodical and consistent computation of sensitivity is performed, conventional deterministic formalisms can be sufficient to propagate nuclear data uncertainties with the level of accuracy obtained by the most advanced tools, such as state-of-the-art Monte Carlo codes. By applying our developed methodology to three exercises proposed by the OECD (UACSA Benchmarks), we provide insights of the underlying physical phenomena associated with the used formalisms.

Submitted by Claudinei Pracidelli (cpracide@ipen.br) on 2016-11-11T16:39:02Z No. of bitstreams: 0 / Made available in DSpace on 2016-11-11T16:39:02Z (GMT). No. of bitstreams: 0 / O reator de pesquisas IEA-R1 opera em regime de 40 h semanais à potência de 4,5 MW. Nestas condições, os cestos disponíveis para o armazenamento dos elementos combustíveis irradiados possuem menos de metade da sua capacidade inicial. Assim, nestas condições de operação, teremos apenas cerca de seis anos de capacidade para armazenamento. Considerando que a vida útil desejada do IEA-R1 seja de pelo menos mais 20 anos, será necessário aumentar a capacidade de armazenamento de combustível irradiado. Dr. Henrik Grahn, especialista da Agência Internacional de Energia Atômica sobre o armazenamento molhado (em piscinas de estocagem), ao visitar o reator IEA-R1 (setembro/2012) fez algumas recomendações. Entre elas, a concepção e instalação de cestos fabricados com aço inoxidável borado e internamente revestidos com uma película de alumínio, de modo que a corrosão dos elementos combustíveis não ocorresse. Após uma revisão da literatura sobre opções de materiais disponíveis para esse tipo de aplicação chegamos ao BoralcanTM fabricado pela 3M devido suas propriedades. Este trabalho apresenta estudos sobre a análise de criticalidade com o código computacional MCNP-5 utilizando duas bibliotecas americanas de dados nucleares avaliados (ENDF/B-VI e ENDF/B-VII) comparativamente. Estas análises demonstraram a possibilidade de dobrar a capacidade de armazenamento de elementos combustíveis, no mesmo espaço ocupado pelos cestos atuais, atendendo a demanda do reator de pesquisas IEA-R1 e também satisfazendo os requisitos de segurança da Comissão Nacional de Energia Nuclear (CNEN) e da Agência Internacional de Energia Atômica (IAEA). / Dissertação (Mestrado em Tecnologia Nuclear) / IPEN/D / Instituto de Pesquisas Energeticas e Nucleares - IPEN-CNEN/SP

This diploma thesis deals with storing the spent nuclear fuel and reviewing its safety. The theoretical part analyzes the processes taking place while the fuel is being used, such as fission, isotopic changes, fission gas release, cracking, swelling and densification of fuel pellet. The thesis is also focused on handling the spent fuel and on the way it makes from the reactor, through the spent fuel pool, the transportation, various kinds of storing, till the reprocessing and final deep geological repository. Furthermore, this part of the thesis briefly discusses computing code MCNP, its main characteristics, input files and using. The practical part of the work is focused on creating the model of the spent fuel pool located next to the nuclear reactor WWER 440/V213. This type was chosen, because it is the most used type of nuclear reactor in Czech Republic and Slovakia. With the help of the code MCNP, the multiplication factor of the main configurations of the fuel in the pool was calculated, and then the required safety regulations to ensure sufficient subcriticality, so its safety, were checked. Next, several analysis were performed using this model. These analyses were concerning the temperature of coolant, fuel and the use of various nuclear data libraries. In the future this model can be used to realize new analyses with new kinds of fuels, materials and data libraries.

<p>Accelerator-driven systems for transmutation of nuclear waste have been suggested as a means for dealing with spent fuel components that pose potential radiological hazard for long periods of time. While not entirely removing the need for underground waste repositories, this nuclear waste incineration technology provides a viable method for reducing both waste volumes and storage times. Potentially, the time spans could be diminished from hundreds of thousand years to merely 1.000 years or even less. A central aspect for accelerator-driven systems design is the prediction of safety parameters and fuel economy. The simulations performed rely heavily on nuclear data and especially on the precision of the neutron cross section representations of essential nuclides over a wide energy range, from the thermal to the fast energy regime. In combination with a more demanding neutron flux distribution as compared with ordinary light-water reactors, the expanded nuclear data energy regime makes exploration of the cross section sensitivity for simulations of accelerator-driven systems a necessity. This fact was observed throughout the work and a significant portion of the study is devoted to investigations of nuclear data related effects. The computer code package EA-MC, based on 3-D Monte Carlo techniques, is the main computational tool employed for the analyses presented. Directly related to the development of the code is the extensive IAEA ADS Benchmark 3.2, and an account of the results of the benchmark exercises as implemented with EA-MC is given. CERN's Energy Amplifier prototype is studied from the perspectives of neutron source types, nuclear data sensitivity and transmutation. The commissioning of the n_TOF experiment, which is a neutron cross section measurement project at CERN, is also described.</p>

Accelerator-driven systems for transmutation of nuclear waste have been suggested as a means for dealing with spent fuel components that pose potential radiological hazard for long periods of time. While not entirely removing the need for underground waste repositories, this nuclear waste incineration technology provides a viable method for reducing both waste volumes and storage times. Potentially, the time spans could be diminished from hundreds of thousand years to merely 1.000 years or even less. A central aspect for accelerator-driven systems design is the prediction of safety parameters and fuel economy. The simulations performed rely heavily on nuclear data and especially on the precision of the neutron cross section representations of essential nuclides over a wide energy range, from the thermal to the fast energy regime. In combination with a more demanding neutron flux distribution as compared with ordinary light-water reactors, the expanded nuclear data energy regime makes exploration of the cross section sensitivity for simulations of accelerator-driven systems a necessity. This fact was observed throughout the work and a significant portion of the study is devoted to investigations of nuclear data related effects. The computer code package EA-MC, based on 3-D Monte Carlo techniques, is the main computational tool employed for the analyses presented. Directly related to the development of the code is the extensive IAEA ADS Benchmark 3.2, and an account of the results of the benchmark exercises as implemented with EA-MC is given. CERN's Energy Amplifier prototype is studied from the perspectives of neutron source types, nuclear data sensitivity and transmutation. The commissioning of the n_TOF experiment, which is a neutron cross section measurement project at CERN, is also described.

This master’s thesis deals with usage of burn-up (spent) nuclear fuel in nuclear power plants of next generation – accelerator driven transmutation plants which is produced in current nuclear power plants. This system could significantly reduce the volume of dangerous long-lived radioisotopes and moreover they would be able to take advantage of its great energy potential due to fast neutron spectrum. In the introduction are listed basic knowledge and aspects of spent nuclear fuel along with its reprocessing and the possibility of further use while minimizing environmental impact. As another point detailed description of accelerator driven systems is described together with its basic components. In addition this search is followed by individual chronological enumeration of projects of global significance concerning their current development. Emphasis is placed on SAD and MYRRHA projects, which are used like base for calculations. This last, computational part, deals with the creation of the geometry of subcritical transmutation reactor driven by accelerator and subsequent evaluation which assembly is the most effective for transmutation and energy purposes along with changing of target, nuclear fuel and coolant/moderator.