The fuel cladding is an essential component in the defence-in-depth strategy for nuclear safety. Its integrity and durability are therefore critical for maintaining acceptable safety conditions. However, the integrity of the cladding can be compromised during normal operation due to corrosion and hydriding. To ensure a sufficient level of safety, design and safety criteria have been established to limit oxidation and hydriding. EDF has various multiphysics software tools at its disposal to ensure that these criteria are met. One such tool, CYRANO3, uses oxide thickness measurements from the beginning of the French nuclear industry to model corrosion and hydriding. This study aims to improve CYRANO3 by expanding its validation database and improving its models. The first part of the study focuses on improving the CYRANO3 database by providing a more comprehensive understanding of normal corrosion in a pressurized water reactor, allowing the models to be recalibrated to better represent actual corrosion behaviour.  In the second part, a deeper analysis is conducted to improve the models and increase knowledge of the parameters that influence corrosion. This analysis highlights the significance of temperature and power as input parameters, which will affect the accuracy of CYRANO3 results. Additionally, this study has identified areas for further improvement, including modifications to the implemented corrosion models and a better understanding of the assumptions made about input data. / Brunnkapslingen utgör den första barriären i principen för försvarsdjup. Därför är dess integritet och hållbarhet en prioritet för att upprätthålla acceptabla säkerhetsförhållanden. Integriteten hos brunnskapslingen kan hotas under normal drift genom korrosion och hydridisering. Därför har design- och säkerhetskriterier fastställts för att säkerställa en tillfredsställande säkerhetsnivå genom att begränsa oxidation och hydridisering. EDF har flera multiphysics-program tillgängliga för att säkerställa att dessa kriterier uppfylls. CYRANO3, det program som används i denna studie, baserar sina korrosion- och hydridiseringmodeller på mätningar av oxidthickness som har registrerats sedan starten av den franska kärnkraftsindustrin. Kvaliteten och sanningshalten hos dessa modeller måste testas och uppdateras med den nya erfarenheten av den franska flottan. Denna studie handlar om att förbättra CYRANO3 genom att utöka dess valideringsdatabas och förbättra dess modeller. I den första delen förbättrar förbättringen av CYRANO3-databasen en mer fullständig bild av normal korrosion i en tryckvattenreaktor. Denna utökning gör att modellerna kan recalibreras för att mer exakt återspegla den faktiska korrosionsbeteendet. I den andra delen görs en djupare analys för att förbättra modellerna och öka kunskapen om de parametrar som påverkar korrosionen. Denna analys visar på vikten av temperaturen och kraften som indataparametrar, vilka kommer att definiera kvaliteten på CYRANO3:s resultat. Dessutom har denna studie öppnat upp för nya områden för förbättring, bland annat genom korrigeringar av de implementerade korrosionsmodellerna och en bättre kunskap om antagandena om indatadatan.

Coordenação de Aperfeiçoamento de Pessoal de Nível Superior / Nesta pesquisa foram estudados os efeitos de aditivos químicos no processo de cimentação de rejeitos radioativos. Estes aditivos são utilizados para melhorar propriedades da pasta e do produto solidificado contendo estes rejeitos. Há uma grande variedade destes materiais no mercado, porém eles são frequentemente alterados ou simplesmente excluídos, portanto é essencial conhecer os aditivos disponíveis comercialmente e seus efeitos. Os testes foram realizados com uma solução simulando os rejeitos do concentrado do evaporador gerado em reatores nucleares do tipo PWR (Pressurized Water Reactor). Para a sua cimentação foram utilizadas duas formulações, A e C, incorporando-se maior ou menor quantidade de rejeito, respectivamente. Acrescentaram-se aditivos químicos de dois fabricantes diferentes (S e H), sendo eles: aceleradores e retardadores de pega e superfluidificantes. Os experimentos foram organizados seguindo o planejamento fatorial 23 para quantificar os efeitos das formulações, dos aditivos, de suas quantidades e de seus fabricantes nas propriedades da pasta e do produto cimentado. Os parâmetros avaliados foram: a viscosidade, o tempo de pega, as densidades da pasta e do produto e a resistência à compressão. Os superfluidificantes forneceram os melhores resultados de viscosidade (trabalhabilidade), densidade e tempo de pega. Os aceleradores de pega apresentaram os maiores valores de resistência à compressão, aos 28 dias de idade, e os menores valores de densidade. Para o parâmetro tempo de pega, o uso destes aceleradores e retardadores de pega seria recomendado somente para casos específicos. Todos os resultados de resistência à compressão atenderam aos requisitos dos órgãos reguladores, com exceção da mistura com a menor quantidade de superfluidificante do fabricante S. A utilização do superfluidificante permite a cimentação em sistemas de misturas mais simples e de menor custo, diferentemente do uso de retardador e acelerador de pega que necessitam de sistemas de mistura específicos. Apresentam pastas mais homogêneas e maior qualidade dos produtos solidificados. Os maiores valores de densidade foram obtidos utilizando o Superfluidificante, pois o seu uso tornou a pasta mais fluída e com melhor trabalhabilidade, diminuindo assim o aparecimento de poros e imperfeições. De modo geral, o uso do superfluidificante apresentou melhores resultados nos parâmetros avaliados. / In this research it has been studied the effects of chemical admixtures in the cementation process of radioactive wastes. These additives are used to improve the properties of waste cementation process, both of the paste and of the solidified product. However there are a large variety of these materials that are frequently changed or taken out of the market. Then it is essential to know the commercially available materials and their effects. The tests were carried out with a solution simulating the evaporator concentrate waste coming from PWR nuclear reactors (Pressurized Water Reactor). It was cemented using two formulations, A and C, incorporating higher or lower amount of waste, respectively. It was added chemical admixtures from two different manufacturers (S and H), which were: accelerators, set retarders and superplasticizers. The experiments were organized by a factorial design 23 to quantify the effects of formulations, admixtures, their quantity and manufacturer in properties of the paste and products. The measured parameters were: the viscosity, the setting time, the paste and product density and the compressive strength. The superplasticizers provided the best results of viscosity (workability), density and setting time. The accelerators showed the highest values of compressive strength, at age of 28 days, and the lowest values of density. In relation to the setting time, set retarders and accelerators would be recommended only for specific cases. All results of compressive strength met the requirements of regulators, except for the mixture with the lower amount of superplasticizers from the manufacturer S. The use of superplasticizers allows the cementation systems of mixtures simpler and lower cost, differently from the use of set retarders and accelerators which require particular mixing systems. Presented paste more homogeneous and better quality products solidified. The highest density values were obtained using Superfluidificante, since their use became more fluid and the pastes with better workability, thereby reducing the appearance of pores and imperfections. In general, the use of superfluidificante showed better results in all evaluated parameters.

O presente trabalho consiste no desenvolvimento de um programa computacional para simulação de transitórios provocados por paradas de bombas de circulação de refrigerante do núcleo de reatores nucleares e o acoplamento do modelo a circuitos hidráulicos típicos de reatores PWR. O programa fornece a variação temporal da vazão, rotação, torque elétrico e hidráulico e altura dinâmica das bombas para diversas situações operacionais ou acidentais. Pode ser utilizado para qualquer combinação de bombas ativas ou inativas. A verificação do modelo foi feita através de testes realizados com dados da usina nuclear de Angra-I e os resultados podem ser considerados satisfatórios quando comparados àqueles apresentados no relatório final de Angra, FSAR.

This is a M. Eng. degree project at Uppsala University carried out at Vattenfall NuclearFuel AB. The goal of it is to present a best estimate method based on the code package CASMO/SIMULATE for the purpose of calculating the isotopic concentrations of a specified number of isotopes in a fuel sample. The calculations done with the method shall produce small deviations from reliable measured values, which characterize the accuracy of CASMO/SIMULATE, but also simplicity based onthe computing time and handling of the amount of data is an important factor in the development of the method. The development of the method has been based on a sensitivity calculation with CASMO/SIMULATE on a number of relevant parameters affecting the isotope concentrations. The proposed method has then been applied on three samples irradiated in Ringhals 4 and Ringhals 3. At last the calculated isotopic concentrations have been benchmarked against measured data from Studsvik Laboratory. The sensitivity analyzes has shown that the parameters affecting the neutron moderation are very important for calculating the isotopic concentrations. The core axial resolution is also an important factor for the samples taken from top of the rod,where the power gradient is large. The comparison of the calculated and measured values has shown that SIMULATE, in the analysed cases, simulates a lower finalburnup. This has created a need to correct the final burnup in order to get better results in terms of lower relative deviations between the measured and calculated data.

The nuclear waste currently generated in the United States is stored in spent fuel pools and dry casks throughout the country awaiting a permanent disposal solution. One efficient solution would be to remove the actinides from the waste and transmute these isotopes in a fast spectrum reactor. Currently this technology is unavailable on a commercial scale and a considerable amount of research and development is still required. An alternate solution is to reprocess and recycle the used fuel in thermal reactors, creating new fuel while reducing the amount of waste and its impact to the environment. This thesis examines the possibility of multi-recycling the transuranics (Pu, Np, Am, and Cm) in a standard pressurized water reactor (PWR). Two types of recycling strategies will be examined: one where Pu, Np, and Am are recycled (TRU-Cm) and a second where the previous isotopes as well as Cm are recycled (TRU+Cm). To offset the hardened neutron spectrum that results from the inclusion of the transuranics, a smaller fuel pin is employed to provide additional moderation. Computer simulations are used to model the in-reactor physics and long-term isotopic decay. Each fuel type is assessed based on the required U-235 enrichment, void coefficient, transuranic production/destruction, and radiotoxicity reduction as compared to a UOX and MOX assembly. It is found that the most beneficial recycling strategy is the one where all of the transuranics are recycled. The inclusion of Cm reduces the required U-235 enrichment, compared to the other multi-recycled fuel and, after a significant number of recycles, can result in the required enrichment to decrease. This fuel type also maintains a negative void coefficient for each recycle. The void coefficient of the fuel type without Cm becomes positive after the third cycle. The transmutation destruction of the two multi-recycled assemblies is less than that of a MOX assembly, but the transmutation efficiency of the multi-recycled assemblies exceeds the MOX assemblies. The radiotoxicity of both multi-recycled assemblies is significantly lower than the UOX and MOX with the TRU+Cm fuel being the lowest. When Curium is recycled only 28,000 years are required for the radiotoxicity of the waste to reach that of natural Uranium and when Cm is not recycled, the amount of time increases to 57,000 years.

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Les conduites primaires des centrales nucléaires REP (304L et 316L) sont protégées contre la corrosion par une couche d'oxyde. Ces conduites, qui forment une barrière entre le milieu primaire et l'extérieur, subissent des phénomènes liés à la corrosion sous contrainte (assisté ou non par irradiation) pouvant entrainer des dommages. L'objectif de cette étude est de comprendre les phénomènes régissant les stades initiaux de formation des couches d'oxydes sur ces alliages en tenant compte de l'orientation cristallographique des grains sous-jacents. En utilisant des techniques de caractérisation avancées comme GIXRD, spectroscopie Raman, XPS, MEB et MET (BF, HRTEM, Astar, EELS et HREELS), sur les toutes premières étapes de l'oxydation, une séquence d'oxydation a été proposée pour l'alliage 316L. / PWR primary pipes (304L & 316L) are protected against corrosion by an oxide layer. Pipes, which are forming a barrier between the PWR environment and outside, undergo stress corrosion cracking phenomena (assisted or not by irradiation) which can lead to damages. The purpose of this study is to understand the phenomena governing the first steps of the oxide layers formation on these alloys by taking acount the cristallographic orientation of the underlying grains. Using advanced caracterisations methods as GIXRD, Raman spectroscopy, XPS, SEM, TEM (BF, HRTEM, Astar, EELS and HREELS, on the very beginning of the oxidation, an oxidation process have been proposed for 316L alloy.

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