Calculating or predicting the total in-core nuclear heating is a difficult tast. Full-core models can be constructed in a Monte Carlo code, such as MCNP6 or TRIPOLI4, and will allow an analyst to calculate the prompt-gamma heating at any given in-core location; however, such codes are generally unable to track the activated or fission-product isotopes and therefore the delayed-gamma sources can't be included in such a model. Some analysts have coupled Monte Carlo transport codes to burnup codes in an effort to include delayed-gamma sources, but the solutions tend to be reactor specific, time-independent and a lot of work. New ideas are required to calculate the total time-dependent in-core nuclear heating. Within this report, two new models have been derived: the nuclear heating equation, and the coupled neutron and nuclear heating point kinetics (NHPK) equations. These models can be used to calculate the time and position-dependent in-core heating. The nuclear heating equations are generalized expressions of the nuclear heating in a volume of interest, within an arbitrary geometry; these equations use Monte Carlo tallies as coefficients and treat the geometry's scalar neutron flux within as the independent variable. The NHPK model describes the nuclear heating in a volume of interest, within a critical assembly by coupling nuclear heating to the famous neutron point kinetics equations. A SCK-CEN gamma thermometer (GT) was commissioned in a materials testing reactor (MTR), the McMaster Nuclear Reactor (MNR), to measure the dynamic in-core nuclear heating in two locations. The nuclear heating equation was used to calculate self-heating of the SCK-CEN GT by neutron capture reactions. This calculation used CapGam and IAEA PGAA prompt-gamma emission data; delayed-particle emission data from NuDat 2.6 was also employed. Analysis of the GT's signal resulted in a quantitative description of the dynamic delayed-gamma heating in MNR, and provided the coefficients for the NHPK model. The NHPK model is capable of reproducing the measured time-dependent nuclear heating, and therefore should also be capable of predicting in-core nuclear heating as a function of reactor power. / Dissertation / Doctor of Philosophy (PhD)

La volonté d'améliorer l'efficacité énergétique des réacteurs nucléaires a motivé de nouvelles solutions dans leur conception. L'une d'elles est l’utilisation d’un réflecteur lourd dans les réacteurs de génération III+ et les futurs réacteurs de génération IV. Lorsque la matière est traversée par des rayons γ, les excitations induites entraînent une élévation de sa température. Ce processus, appelé échauffement γ, est responsable de plus de 90% de la production de chaleur dans la région hors combustible d'un réacteur nucléaire. C’est également le cas dans le réflecteur. Pour simuler l'effet de l’échauffement γ en fonction de la composition du combustible, il faut disposer de données précises sur les γ prompts émis par les différents fragments produits dans le processus de fission. En 2012, une campagne d’expériences inédite, EXILL, a été menée au réacteur de recherche de l'ILL. Un grand nombre de détecteurs HPGe a été placé autour d’une cible fissile et a mesuré les rayons γ émis par la cible alors qu’elle était irradiée par un faisceau intense de neutrons froids. Dans ce travail, nous avons analysé les données obtenues avec des cibles ²³⁵U. Elles nous ont permis d’étudier la désexcitation de plusieurs fragments de fission et plus globalement le processus de fission induite par des neutrons. Dans un premier temps, nous avons utilisé la méthode standard d'analyse par coïncidence γ-γ-γ. Nous avons pu filtrer les données expérimentales, identifier les transitions γ dans des fragments bien produits et calculer leur intensité relative. Les problèmes que nous avons rencontrés concernent le bruit de fond. Les résultats obtenus dépendent de ce choix et présentent donc des problèmes de reproductibilité. Nous avons développé et testé une nouvelle méthodologie d'analyse. Son principe est un balayage des portes de coïncidence selon trois directions, ce qui permet de trouver le bruit de fond le mieux adapté. L'idée principale était finalement de passer d'une méthode "spectroscopique", dont le but est de trouver de nouvelles transitions et des états excités dans un noyau, à une méthode "spectrométrique", qui nous permet d'obtenir plus précisément l’intensité de transitions γ connues, avec une meilleure estimation de leur incertitude. Cela nous a amené à développer un logiciel d'analyse semi-automatique d'ajustement des pics. Divers schémas de calcul de l'intensité des transitions γ ont été également élaborés pour tenir compte des contaminations possibles, selon leur emplacement dans la matrice de coïncidence et leur intensité. La méthode standard et la nouvelle méthode d'analyse ont été comparées pour l'analyse du ¹⁴²Ba. Dans ce travail, nous avons également comparé nos résultats sur quelques noyaux, tel que le ¹⁰⁰Zr, avec des simulations réalisées avec le code FIFRELIN. Ce dernier est un code Monte-Carlo qui simule le processus de fission et la désexcitation des fragments de fission. FIFRELIN utilise plusieurs modèles différents pour décrire ces processus. Nous avons testé le comportement des différents modèles, trouvé les valeurs optimales des paramètres de simulation et testé comment ces configurations reproduisaient les résultats expérimentaux. FIFRELIN n'a pas été en mesure de reproduire simultanément les intensités des transitions γ émises par les fragments de ¹⁰⁰Zr et la multiplicité de neutrons prompts moyennée sur tous les fragments de fission. Cependant, avec des paramètres modifiés, FIFRELIN a fourni localement une multiplicité de neutrons prompts correcte pour les fragments de masse atomique A=100 et des intensités de transition γ bien reproduites pour le noyau de ¹⁰⁰Zr. Nous avons également comparé nos résultats expérimentaux sur les fragments de ¹⁰⁰Zr provenant du processus ²³⁵U(n,f) avec les autres données expérimentales disponibles provenant des expériences sur ²⁴⁸Cm(sf) et ²⁵²Cf(sf), et une autre expérience sur ²³⁵U(n,f). / The desire to improve the fuel efficiency of nuclear reactors has motivated new solutions in their design. One of them is the heavy reflector used in the generation III+ and in the future generation IV reactors. γ-rays passing through matter cause its excitation and temperature rise. It is a process called γ-heating, and it is responsible for more than 90% of the heat production in the non-fuel region of the nuclear reactor. This is also the case of the heavy reflector. To simulate the γ-heating effect in every state of the nuclear reactor it is necessary to have precise data on the prompt γ-rays emitted by different fission fragments produced in the course of the nuclear chain reaction. In 2012, at the research reactor of the ILL, an innovative experiment, called EXILL, was conducted. It produced a large amount of useful data on the de-excitation of the fission fragments. A large number of HPGe detectors were used to study the neutron induced fission process by measuring the emitted γ-rays. Fissile targets were irradiated by an intense cold neutron beam. In this work we analyzed the ²³⁵U targets. We studied several fission fragments and more generally the fission process by using high-resolution γ-ray spectroscopy. At the beginning, we used the standard γ-γ-γ coincidence analysis method. We were able to filter experimental data, identify the well produced γ-rays, and calculate their relative intensities. The problems we have encountered are related to the background. The results obtained with this method were background dependent and thus presented some problems with reproducibility. We therefore developed and tested a new analysis methodology. Its crucial feature is a coincidence gates scanning in three directions which helps to find the most suitable background. The idea was to move from a “spectroscopic” method, which main purpose is finding new transitions and excited states in a nucleus, to a “spectrometric” method, which allows us to obtain more precise γ-ray intensities. We developed a semi-automatic analysis software which facilitates fitting of the chosen γ-ray peak, the contamination and the background. Various γ-ray intensity calculation schemes were derived to take into account different contamination strengths and placements. The results of the analysis with the new technique are reproducible and more reliable. The standard and the new analysis method were compared in the ¹⁴²Ba analysis. In this work, we also compared our experimental results on some nuclei, such as ¹⁰⁰Zr, with the simulation results performed with the FIFRELIN code. It is a Monte-Carlo code which simulates the fission process and the de-excitation of the fission fragments. It uses various models to describe these processes. We were able to test the behavior of different models implemented in FIFRELIN to find the optimal simulation parameter values and to test how well these setups reproduce the experimental results. FIFRELIN was unable to simultaneously reproduce the γ-ray intensities of ¹⁰⁰Zr and the prompt-neutron multiplicity averaged over all fission fragments. However, with modified simulation parameters, FIFRELIN locally provided correct prompt-neutron multiplicity for the fission fragment with the atomic mass A=100 and well reproduced γ-ray intensities of ¹⁰⁰Zr. We also compared our experimental results on ¹⁰⁰Zr coming from the ²³⁵U(n,f) process with the other available experimental data coming from the experiments on ²⁴⁸Cm(sf) and ²⁵²Cf(sf), and another experiment on ²³⁵U(n,f).

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 travaux réalisés durant cette thèse rentrent dans cette problématique. Ils concernent d’une part le développement d’un schéma de calculs et d’évaluation des échauffements nucléaires générés dans le réacteur MARIA en utilisant les codes français de transport neutronique TRIPOLI-4 © et APOLLO-2. Les travaux dans ce volet ont concerné principalement les calculs des échauffements photoniques induits par les rayonnements gammas essentiellement. D’autre part des travaux expérimentaux ont été conduits durant cette thèse. Ils ont concerné la mesure des échauffements nucléaires dans des emplacements spécifiques du réacteur MARIA. Cela a permis une première validation des schémas de calcul adoptés. Des comparaisons C/E ont été effectuées. Elles sont présentées et discutées dans cette thèse. Cela a permis d’émettre des recommandations quant aux techniques de mesure des échauffements nucléaires dans le réacteur MARIA et les moyens de modélisation qui peuvent être associés. Les comparaisons calculs-expérience font l’objet du cinquième. Les écarts relevés entre les résultats des modélisations et les mesures des échauffements nucléaires pour différentes configurations de mesures (au moyen de GT et de calorimètre mono cellule KAROLINA) permettent de dégager grâce à ces premiers travaux de thèse des recommandations pertinentes pour les travaux futurs. / This thesis work presents a calculation scheme which enables evaluation of heat generation from nuclear reactions in the MARIA nuclear reactor by use French computational codes TRIPOLI-4 © (TRIPOLI-4 is a registered trademark of CEA) and Apollo-2. Particular attention was devoted to the heat induced by gamma radiation. The thesis also presents measurements of nuclear heating in selected locations inside MARIA MTR reactor. This allows reaching first steps of validation and qualification of computer calculations. Research and analysis presented in the thesis allow one to compare the results obtained by using proposed calculation scheme with the experimental measurement methods. Finally, further works and perspectives were proposed on the development of the calculations and experimental measurements of nuclear heating in nuclear reactors.Qualifying the calculations was possible by performing especially dedicated 7-day core measurement campaigns. Nuclear heating measurements were performed with gamma thermometers and specially designed KAROLINA calorimeter. All measurement devices used were mounted in a dedicated probe, designed and built for this purpose, which allowed for the adjustment of instruments position inside the MARIA core. The main scientific hypothesis of this work is that currently available Monte Carlo simulations of neutron and gamma transport can be used to correct and accurate calculations of prompt nuclear heating in nuclear reactor, whereas delayed component of nuclear heating can be determined experimentally. For this purpose new calculation scheme and improvements in nuclear heating measurements were implemented.