Afin d’évaluer les performances du combustible irradié en situation de stockage géologique, des recherches sont menées sur le comportement à long terme des combustibles irradiés (UOx et MOx) en conditions environnementales se rapprochant de celles du site de stockage français. L’objectif de cette thèse est de déterminer si la géochimie de la couche géologique d'argilites du Callovo-Oxfordien (COx) et la corrosion des conteneurs en acier (produisant du fer et de l'hydrogène) ont un impact sur la dissolution oxydante de la matrice (U,Pu)O2 sous radiolyse alpha de l’eau.Des expériences de lixiviation ont été réalisées avec des pastilles de UO2 dopées en émetteurs alpha (Pu) et du combustible MOx MIMAS (non irradié ou irradié en réacteur) afin de mettre en évidence l’influence de l’eau du COx et de la présence de fer métallique sur la dissolution oxydante de ces différents matériaux induite par la radiolyse de l’eau. Les résultats indiquent un effet inhibiteur de l’eau du COx sur la dissolution oxydante de la matrice UO2. D’autre part en présence de fer, deux régimes différents sont observés. Sous irradiation alpha dominante telle que celle attendue en stockage géologique, la dissolution oxydante de la matrice UO2 et du combustible MOx est très fortement inhibée du fait de la consommation des espèces radiolytiques oxydantes par le fer en solution avec précipitation d’hydroxydes de Fe(III) à la surface des pastilles. En revanche, sous forte irradiation beta/gamma comme dans le cas du combustible irradié, les traceurs de l’altération indiquent que celle-ci se poursuit en présence de fer tandis que la concentration en uranium en solution est contrôlée par la solubilité de UO2(am,hyd). Ceci est expliqué par le déplacement du front redox de la surface du combustible vers la solution homogène ne protégeant plus le combustible. Les modèles géochimique (code CHESS) et de transport réactif (code HYTEC) développés représentent correctement les principaux résultats et mécanismes mis en jeu. / To assess the performance of direct disposal of spent fuel in a nuclear waste repository, researches are performed on the long-term behavior of spent fuel (UOx and MOx) under environmental conditions close to those of the French disposal site. The objective of this study is to determine whether the geochemistry of the Callovian-Oxfordian (COx) clay geological formation and the steel overpack corrosion (producing iron and hydrogen) have an impact on the oxidative dissolution of the (U,Pu)O2 matrix under alpha radiolysis of water.Leaching experiments have been performed with UO2 pellets doped with alpha emitters (Pu) and MIMAS MOx fuel (un-irradiated or spent fuel) to study the effect of the COx groundwater and of the presence of metallic iron upon the oxidative dissolution of these materials induced by the radiolysis of water. Results indicate an inhibiting effect of the COx water on the oxidative dissolution. In the presence of iron, two different behaviors are observed. Under alpha irradiation as the one expected in the geological disposal, the alteration of UO2 matrix and MOx fuel is very strongly inhibited because of the consumption of radiolytic oxidative species by iron in solution leading to the precipitation of Fe(III)-hydroxides on the pellets surface. On the contrary, under a strong beta/gamma irradiation field, alteration tracers indicate that the oxidative dissolution goes on and that uranium concentration in solution is controlled by the solubility of UO2(am,hyd). This is explained by the shifting of the redox front from the fuel surface to the bulk solution not protecting the fuel anymore. The developed geochemical (CHESS) and reactive transport (HYTEC) models correctly represent the main results and occurring mechanisms.

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

About 60 percent of the electricity production in Belgium originates from nuclear power plants. Belgium owns 7 nuclear pressurized water reactors, which are located in two sites: 4 reactors in Doel and 3 reactors in Tihange. Together they have a capacity of approximately 5900 MWe. All these reactors use classical uranium oxide fuel assemblies. Two of them (Doel3, Tihange2) have also accepted a limited number of mixed (uranium and plutonium) oxide fuel assemblies. These mixed fuel assemblies came from the reprocessing of spent uranium oxide fuel assemblies in La Hague (France). The reprocessing of spent fuel gives birth to vitrified high-level waste, and to different isotopes of uranium and plutonium, which can be used in the manufacture of mixed oxide fuel assemblies. Each country producing radioactive waste must find a solution to dispose them safely. The internationally accepted solution is to dispose high-level radioactive waste in a deep and stable geological layer. This seems to be the most secure and environment-friendly way to get rid of the high-level radioactive waste. One of the few stable geological layers, which could accept radioactive waste in Belgium, is the Boom clay layer. Another possible layer is the Ypresian clay layer, but it is not the reference option for the moment. The Boom clay layer is quite thin (about 100 m thick) and is not at a large depth (about 240 m below the ground surface) at the proposed disposal site, beneath the SCK CEN Nuclear Research Centre in Mol. A large number of studies have already been performed on the Boom clay layer, and on the possibility of building a high-level radioactive waste repository in this geological medium. Since 1993, the Belgian government has promulgated a moratorium on the reprocessing of spent uranium oxide fuels in La Hague. Since then, spent fuel assemblies are considered as waste, and ONDRAF/NIRAS (the Belgium Agency for Radioactive Waste and Enriched Fissile Materials) has thus to deal with them as waste. This rises a number of questions on how to deal with this new kind of waste. A solution is to directly dispose these spent fuel assemblies in containers in a repository, just like the other high-level radioactive waste. This repository would be build in the Boom clay layer at a depth of about 240 m beneath the SCK CEN. One of the questions raised by this new kind of waste is: "could the direct disposal of the spent nuclear fuel assemblies lead to nuclear criticality risks in the future?". Nuclear criticality is the ability of a system to sustain a nuclear fission chain reaction. This question was not a key issue with vitrified high-level waste because these do not include fissile uranium and plutonium isotopes, which could lead to a criticality event. The spent fuel repository will be designed in order to totally avoid the occurrence of a criticality event at the closure time. But in the future history of the repository, external events could possibly affect this. These events could maybe lead to criticality inside the repository, and this has also to be avoided. This work tries to answer this question, and to determine how to avoid a long-term criticality event inside the repository. The only complete research work answering this question has been performed in the U.S. for the Yucca Mountain repository but this design is fully different from the Belgian one studied here: for example, the waste are not only spent fuel waste, and the geological layer is volcanic tuff.<p>The main achievements of this work are: <p>*A first set of in-package criticality scenarios for different design options for a Belgian spent fuel repository in the Boom clay layer. <p>*A large number of criticality calculations with different parameters (fuel type, fuel burnup, fuel enrichment, distance between the fuel assemblies, distance between the fuel rods, water fraction inside the overpack) for the different design options. <p>*A preliminary study of the effects of the spent fuel assemblies isotopic evolution with time on the multiplication factor. <p>*For the first time, a coupling between the in-package criticality scenarios and the criticality calculations has been performed. / Doctorat en sciences appliquées / info:eu-repo/semantics/nonPublished

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.

Author´s name: Bc. Stanislav Klimek School: Brno University of Technology, Faculty of Mechanical Engineering, Energy institute Title: Determining the life storage of a carbon steel cask Consultant: Prof. Ing. Oldřich Matal, CSc. Number of pages: 70 Year: 2009 The assignment of this diploma thesis is to estimate the lifetime of spent fuel container made from carbon steel grade. This container is designed for deep geological disposal of spent nuclear fuel. Basic mechanism of corrosion are described in detail in the first part. Further on, this work deals with the other specific phenomena and influences, which affect at corrosion of steel in conditions of a deep geological repository. Heat, radiation and surroundings are considered of particular importance. In the following part an estimate of the lifetime of model container is introduced, which is affected by temperature and radiation. Here recommendations for protection of container are introduced, arising from the model calculation. Finally, the relevancy of incidence of particular parameters is evaluated, which affect the corrosion.

This dissertation covers the application of heat accumulation systems in nuclear power engineering, namely in nuclear power plants. It is mainly a case of passive emergency systems, whose task is to accumulate the heat produced in the reactor’s active zone and in spent fuel pools during DBA (design-basis accidents) or beyond DBA. A particular example of heat accumulation is steam condensation after LOCA (loss of coolant accident). The primary circuit steam leakage increases containment pressure and has to be decreased by the steam condensation. This thesis deals with a theoretical substitute for ice condensers, which are used as a passive safety measure in some nuclear power plants. The substitute involves a choice of an alternative material, whose melting temperature (for heat accumulation) is closer to nuclear power plant operating temperatures. The other part of the dissertation discusses heat accumulation in spent fuel pools in case of all cooling systems failure.

The diploma thesis aim to the design of equipment for ŠKODA JS a.s., which is part of a new inspection stand (N-SIO). This equipment is an inspection column that provides the possibility to inspect spent fuel assemblies at the operation of the Temelín nuclear power plant. This master thesis is also a summary of the whole design of the new inspection stand and a description of the individual inspection components and devices. One of the parts of the thesis is also a basic strength analysis and a drawing of the main assembly of inspection equipment.

This work seeks to illustrate the potential benefits afforded by implementing aspects of fluid dynamics, especially the latest computational fluid dynamics (CFD) modeling approach, through numerical experimentation and the traditional discipline of physical experimentation to improve the calibration of the severe reactor accident analysis code, MELCOR, in one of several spent fuel pool (SFP) complete loss-of-coolant accident (LOCA) scenarios. While the scope of experimental work performed by Sandia National Laboratories (SNL) extends well beyond that which is reasonably addressed by our allotted resources and computational time in accordance with initial project allocations to complete the report, these simulated case trials produced a significant array of supplementary high-fidelity solutions and hydraulic flow-field data in support of SNL research objectives. Results contained herein show FLUENT CFD model representations of a 9x9 BWR fuel assembly in conditions corresponding to a complete loss-of-coolant accident scenario. In addition to the CFD model developments, a MATLAB based control-volume model was constructed to independently assess the 9x9 BWR fuel assembly under similar accident scenarios. The data produced from this work show that FLUENT CFD models are capable of resolving complex flow fields within a BWR fuel assembly in the realm of buoyancy-induced mass flow rates and that characteristic hydraulic parameters from such CFD simulations (or physical experiments) are reasonably employed in corresponding constitutive correlations for developing simplified numerical models of comparable solution accuracy. / text