Dedicated sensors are widely used throughout many industries to monitor everyday operations, maintain safety, and report performance characteristics. In order to adopt a more sustainable solution, much research is being applied to self-powered sensing, implementing solutions which harvest wasted ambient energy sources to power these dedicated sensors. The adoption of not only wireless sensor nodes, but also self-powered capabilities in the nuclear energy process is critical as it can address issues in the overall safety and longevity of nuclear power. The removal of wires for data and power transmission can greatly reduce the cost of both installation and upkeep of power plants, while self-powered capabilities can further reduce effort and money spent in replacing batteries, and importantly may enable sensors to work even in losses to power across the plant, increasing plant safety. This thesis outlines three harvesting opportunities in the nuclear energy process from: thermal, vibration, and radiation sources in the main structure of the power plant, and from thermal and radiation energy from spent fuel in dry cask storage. Thermal energy harvesters for the primary and secondary coolant loops are outlined, and experimental analysis done on their longevity in high-radiation environments is discussed. A vibrational energy harvester for large rotating plant machine vibration is designed, prototyped, and tested, and a model is produced to describe its motion and energy output. Finally, an introduction to the design of a gamma radiation and thermal energy harvester for spent nuclear fuel canisters is discussed, and further research steps are suggested. / Master of Science / In this work multiple energy harvesters are investigated aimed at collecting wasted ambient energy to locally power sensor nodes in nuclear power plants, and in spent nuclear fuel canisters. Locally self-powered, wireless sensors can increase safety and reliability throughout the nuclear process. To address this a thermal energy harvester is tested in a radiation rich environment, and its performance before and after irradiation is analyzed. A vibrational energy harvester designed for use on large rotating machinery is discussed, manufactured, and tested, and a mathematical model describing it is produced. Finally, an introduction to harvesting radiation and heat given off from spent nuclear fuel in dry cask canister storage is investigated. Power capabilities for each design are considered, and the impact of such energy harvesting for wireless sensor nodes on the longevity, safety, and reliability of nuclear power plants is discussed.

Les activités nucléaires génèrent des déchets radioactifs classés selon leur niveau d’activité et la durée de vie des radioéléments présents. La garantie d’un classement et d’une gestion optimale nécessite une caractérisation précise. Les déchets de moyenne et haute activité, contenant des radioéléments à vie très longue, seront stockés en profondeur pendant plusieurs centaines de milliers d’années, à l’issue desquelles il est nécessaire de pouvoir garantir l’absence de risques pour l’homme et l’environnement, non seulement sur le plan radiologique, mais aussi en ce qui concerne des éléments stables, toxiques du point de vue chimique. Cette thèse concerne la caractérisation par activation neutronique de ces éléments toxiques, ainsi que celle des matières nucléaires présentes dans les colis. Elle a été réalisée dans le cadre d’une collaboration entre le Laboratoire de Mesures Nucléaires du CEA Cadarache, en France, et l’institut de Gestion des Déchets Radioactifs et de Sûreté des Réacteurs du centre de recherche FZJ (Forschungszentrum Jülich), en Allemagne. La première étude a consisté à valider le modèle numérique de la cellule d’activation neutronique MEDINA (FZJ) avec le code de transport Monte Carlo MCNP. Les rayonnements gamma prompts de capture radiative d’échantillons contenant des éléments d’intérêt (béryllium, aluminium, chlore, cuivre, sélénium, strontium et tantale) ont été mesurés et comparés aux simulations avec diverses bases de données nucléaires, permettant d’aboutir à un accord satisfaisant et validant le schéma de calcul en vue des études suivantes. Ensuite, la mesure des rayonnements gamma retardés de fissions induites sur les isotopes 235U et 239Pu a été étudiée pour des fûts de 225 L contenant des enrobés bitumineux ou une matrice béton, représentatifs de déchets produits en France et en Allemagne. Les rendements d’émission des rayonnements gamma retardés de fission d’intérêt, cohérents avec ceux publiés dans la littérature, ont été déterminés à partir des mesures d’échantillons métalliques d’uranium et de plutonium dans la cellule d’activation neutronique REGAIN du LMN. Le signal utile a ensuite été extrapolé par simulation MCNP pour une répartition homogène d’isotopes 239Pu ou 235U dans les matrices considérées, en utilisant le modèle numérique de MEDINA. Des signaux faibles, de l’ordre de 100 coups par gramme d’isotope 239Pu ou 235U, ont été obtenus. Pour le colis d’enrobés bitumineux, le niveau d’irradiation gamma très élevé, dû à une activité en 137Cs de l’ordre de 1 TBq par fût, nécessiterait l’utilisation d’une collimation et/ou d’écrans pour éviter la saturation de l’électronique de mesure, rendant indétectables les rayonnements gamma retardés de fission. Les colis de déchets bétonnés produits en Allemagne présentant un niveau d’activité plus faible, il a été possible d’estimer des limites de détection allant de 10 à 290 g d’isotope fissile 235U ou 239Pu, selon la raie gamma considérée, suite à la mesure du bruit de fond actif dans MEDINA avec une matrice béton maquette. Afin d’améliorer ces performances, le blindage du détecteur germanium de MEDINA a été optimisé à l’aide de simulations MCNP, montrant la possibilité de réduire les bruits de fond gamma et neutron d’un facteur 4 et 5, respectivement. La validation expérimentale de l’efficacité du blindage a été effectuée à partir de configuration simples à implémenter dans MEDINA, confirmant les facteurs de réduction attendus. Un blindage du détecteur optimal permettrait d’améliorer les limites de détection et aussi d’utiliser une source de neutrons d’intensité supérieure, comme un générateur de neutron à haut flux ou un accélérateur linéaire d’électrons avec une cible de conversion appropriée. / Nuclear activities produce radioactive wastes classified following their radioactive level and decay time. An accurate characterization is necessary for efficient classification and management. Medium and high level wastes containing long lived radioactive isotopes will be stored in deep geological storage for hundreds of thousands years. At the end of this period, it is essential to ensure that the wastes do not represent any risk for humans and environment, not only from radioactive point of view, but also from stable toxic chemicals. This PhD thesis concerns the characterization of toxic chemicals and nuclear material in radioactive waste, by using neutron activation analysis, in the frame of collaboration between the Nuclear Measurement Laboratory of CEA Cadarache, France, and the Institute of Nuclear Waste Management and Reactor Safety of the research center, FZJ (Forschungszentrum Jülich GmbH), Germany. The first study is about the validation of the numerical model of the neutron activation cell MEDINA (FZJ), using MCNP Monte Carlo transport code. Simulations and measurements of prompt capture gamma rays from small samples measured in MEDINA have been compared for a number of elements of interest (beryllium, aluminum, chlorine, copper, selenium, strontium, and tantalum). The comparison was performed using different nuclear databases, resulting in satisfactory agreement and validating simulation in view of following studies. Then, the feasibility of fission delayed gamma-ray measurements of 239Pu and 235U in 225 L waste drums has been studied, considering bituminized or concrete matrixes representative of wastes produced in France and Germany. The delayed gamma emission yields were first determined from uranium and plutonium metallic samples measurements in REGAIN, the neutron activation cell of LMN, showing satisfactory consistency with published data. The useful delayed gamma signals of 239Pu and 235U, homogeneously distributed in the 225 L matrixes, were then determined by MCNP simulations using MEDINA numerical model. Weak signals of about one hundred counts per gram of 239Pu or 235U after 7200 s irradiation were obtained. Because of the high gamma emission in the bituminized waste produced in France (about 1 TBq of 137Cs per drum), the use of collimator and/or shielding is mandatory to avoid electronic saturation, making fission delayed gamma rays undetectable. However, German concrete drums being of lower activity, their corresponding active background was measured in MEDINA with a concrete mock-up, leading to detection limits between 10 and 290 g of 235U or239Pu, depending on the delayed gamma line. In order to improve these performances, the shielding of MEDINA germanium detector was optimized using MCNP calculations, resulting in gamma and neutron background reduction factors of 4 and 5, respectively. The experimental validation of the shielding efficiency was performed by implementing easy-to-build configurations in MEDINA, which confirmed the expected background reduction factors predicted by MCNP. Thanks to an optimized detector shielding, it will also be possible to use a higher neutron emission source, like a high flux neutron generator or an electron LINAC with appropriate conversion targets, in view to further reduce detection limits.

Nuclear power plants provide about 11 percent of the world's electricity production. For fission process is uranium fuels used with varying percentage of enrichment 235U for most of nuclear reactors. Uranium reserves are reducing and their mining cost increases. Therefore, the thorium fuel is discussed as revolution fuel for current and future nuclear power plants. This diploma thesis deals with possibility of thorium fuel utilization at various types of nuclear reactors with a focus on light water reactors. The practical part of the thesis is focused on simulation and calculations of various uranium dioxide and thorium dioxide layers at the fuel rods. Model of WWER 440 reactor was developed for the calculations with the addition of thorium fuel. The model simulates burning out of fuel for 5 years, with monitoring of fuel behavior and tracking changes of each material. The thesis tries to define the suitable ratio and parameters of layers combination of uranium and thorium fuel. For these ratios and parameters the thesis tries to give sufficient amount of computational analyzes.