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Nuclear fusion is considered a clean energy source: it emits no CO2 and leaves little radioactive waste. It is important to start paving the path toward nuclear fusion whilst simultaneously moving away from fossil fuels and carbon emissions. One of the challenges of nuclear fusion is the lack of tritium, which, together with deuterium makes up its fuel. This research is focused on utilizing one current method of nuclear fission technology, namely molten salt reactors, to generate at least the initial loads of tritium for the first fusion reactors. Current research is primarily focused on providing tritium during the nuclear fusion reaction. However, it is also necessary to have a tritium supply whenever we start up a nuclear fusion reactor. The largest source of tritium is the CANDU nuclear fission reactor. A typical 500 MW CANDU produces 130 g of tritium annually as a biproduct of power generation. However, a future commercial fusion power plant is expected to consume 300 g of tritium per day to produce 800 MW. Thus, this research explores the possibility of breeding tritium in other fission reactors, in particular molten salt reactors (MSR). MCNP4C was used to simulate a simple Molten Salt Reactor setting with 61 molten salt fuel channels and applying a molten salt blanket to study how the presence of specific elements in the blanket affects tritium production, as well as criticality. The study relies on nuclear data from the National Nuclear Data Center (NNDC), and Oak Ridge National Laboratory (ORNL) as benchmark to verify the accuracy of the results. The calculated output of tritium is 325 g/year for a 100 MW (th) reactor, which is considered a positive outcome that opens the door for more research in this direction. / Thesis / Master of Applied Science (MASc)

京都大学 / 新制・課程博士 / 博士(理学) / 甲第23701号 / 理博第4791号 / 新制||理||1686(附属図書館) / 京都大学大学院理学研究科物理学・宇宙物理学専攻 / (主査)教授 永江 知文, 准教授 成木 恵, 教授 中家 剛 / 学位規則第4条第1項該当 / Doctor of Science / Kyoto University / DFAM

No / This work evaluated the in vitro transdermal iontophoretic delivery of tritiated estradiol from ultradeformable liposomes compared with saturated aqueous solution (control). Effects of current density and application time on tritium exchange with water were also determined. Penetration studies used three Protocols. Protocol I involved occluded passive steady state estradiol penetration from ultradeformable liposomes and control. The effect of current densities on drug penetration rates was also assessed (Protocol II). In Protocol III, three consecutive stages of drug penetration (first passive, iontophoresis and second passive) through the same human epidermal membranes were monitored. Such an experimental design investigated the possible effect of high current density (0.8 mA/cm2) on skin integrity. The tritium exchange study showed that extent of exchange correlated well with current density and time of application, with some shielding of estradiol by the liposomal structure. Liposomes enhanced estradiol passive penetration after occlusion. Protocol II showed that estradiol flux increased linearly with current density, although being delivered against electroosmotic flow. In Protocol III, reduction in flux of the second passive stage to near that of the first reflected a reversibility of the structural changes induced in skin by current.

Tritium generation presents a significant hazard during operation of nuclear reactors and necessitates safety precautions in the case of a combustion incident. Accurate kinetic models should inform these safety precautions, however, reliable tritium data do not exist to generate such models. This work focuses on laminar flame speed, which is an important component of kinetic models. Current estimates based on established kinetic theories and experimental measurements of the other isotopes of hydrogen predict tritium to have a flame speed approximately 70% that of standard protium over a broad range of stoichiometries at one atmosphere in air. These estimates are based solely on isotopic mass differences and do not account for radioactive decay, which, in the case of tritium, is energetic enough to cause significant radiolysis reactions and potentially alter the radical pool for combustion. Simulations of a protium flame present compelling evidence that hydrogen flames are controlled by preferential radical diffusion from the rigorous flame region towards the unburned gases and not by heat conduction and dissociation of stable molecules. These flames rely on very low radical concentrations at the initiation region of the flame and the chemistry may be altered by a slight increase in radicals due to radioactive decay. This work also presents an experimental method suitable for measuring these radioactivity effects on tritium flame speed utilizing direct measurements of a flame propagating through a transparent tube. Measurements of protium and deuterium flame speeds using this method have proven highly repeatable and consistent with literature values while consuming much less reactant than other potential methods.

L’hydrolyse contrôlée et progressive est une des solutions technologiques pour traiter le sodium tritié présent dans divers composants d’un Réacteur à Neutrons Rapides. Une étude expérimentale a été réalisée pour analyser et comprendre les phénomènes physico-chimiques mis en jeu lors de cette hydrolyse, fortement exothermique, et étudier l’influence des paramètres opératoires sur la répartition HT /HTO au sein de l’effluent liquide et de l’effluent gazeux générés. Les deux facteurs prédominants sont l’activité totale du sodium traité et le flux énergétique (J/s) dégagée par la réaction. Un modèle phénoménologique de l’hydrolyse de sodium tritié est proposé pour synthétiser les connaissances acquises et servir d’aide à la prédiction de la composition en tritium dans les effluents générés en vue de leur traitement. / Within the framework of the decommissioning of fast reactors, several processes are under investigation regarding sodium disposal. One of them rests on the implementation of the sodium-water reaction (SWR), in a controlled and progressive way, to remove residual sodium containing impurities such as sodium hydrides, sodium oxides and tritiated sodium hydrides. Such a hydrolysis releases some amount of energy and produces a liquid effluent, composed of a solution of soda, and a gaseous effluent, composed of hydrogen, steam and an inert gas. The tritium, originally into the sodium as a soluble (T-) or precipitate form (NaT), will be distributed between the liquid and gaseous effluent, and according to two chemical forms, the tritium hydride HT and the tritiated water HTO. HTO being 10,000 times more radiotoxic than HT, a precise knowledge of the mechanisms governing the distribution of tritium is necessary in order to estimate the exhaust gas releases and design the process needed to treat the off-gas before its release into the environment. An experimental study has been carried out in order to determine precisely the phenomena involved in the hydrolysis. The influence of the experimental conditions on the tritium distribution has been tested. The results of this study leaded to a phenomenological description of the tritiated sodium hydrolysis that will help to predict the composition of the effluents, regarding tritium.

En France, 23 000 tonnes de graphites irradiés générés par le démantèlement des réacteurs nucléaires de première génération Uranium Naturel-Graphite-Gaz (UNGG) sont en attente d'une solution de gestion à long terme. Cette thèse porte sur le comportement du tritium, l'un des principaux contributeurs à l'inventaire radiologique des graphites à l'arrêt des réacteurs. Afin d'anticiper des rejets de tritium lors du démantèlement ou de la gestion des déchets, il est indispensable d'obtenir des données sur sa migration, sa localisation et son inventaire. Notre étude repose sur la simulation du tritium par implantation de l'ordre de 3 % at. de deutérium jusqu'à environ 3 μm dans un graphite nucléaire vierge. Celui-ci a ensuite subi des recuits jusqu'à 300 h et 1300 ° C sous atmosphère inerte, gaz caloporteur UNGG et gaz humide, dans le but de reproduire des conditions proches de celles rencontrées en réacteur et lors des opérations de gestion des déchets. Les profils et la répartition spatiale du deutérium ont été analysés via la réaction nucléaire 2H(3He,p)4He. Les principaux résultats montrent un relâchement thermique du deutérium se produisant selon trois régimes contrôlés par le dépiégeage de sites superficiels ou interstitiels. L'extrapolation des données au cas du tritium tend à montrer que son relâchement thermique en réacteur pourrait avoir été inférieur à 30 % et localisé à proximité des surfaces libres du graphite. L'essentiel de l'inventaire en tritium à l'arrêt des réacteurs serait retenu en profondeur dans les graphites irradiés, dont la décontamination nécessiterait alors des températures supérieures à 1300 °C, et serait plus efficace sous gaz inerte que sous gaz humide / In France, 23 000 t of irradiated graphite that will be generated by the decommissioning of the first generation Uranium Naturel-Graphite-Gaz (UNGG) nuclear reactors are waiting for a long term management solution. This work focuses on the behavior of tritium, which is one of the main contributors to the radiological inventory of graphite waste after reactor shutdown. In order to anticipate tritium release during dismantling or waste management, it is mandatory to collect data on its migration, location and inventory. Our study is based on the simulation of tritium by implantation of approximately 3 at. % of deuterium up to around 3 μm in a virgin nuclear graphite. This material was then annealed up to 300 h and 1300 °C in inert atmosphere, UNGG coolant gas and humid gas, aiming to reproduce thermal conditions close to those encountered in reactor and during waste management operations. The deuterium profiles and spatial distribution were analyzed using the nuclear reaction 2H(3He,p)4He. The main results evidence a thermal release of implanted deuterium occurring essentially through three regimes controlled by the detrapping of atomic deuterium located in superficial or interstitial sites. The extrapolation of our data to tritium suggests that its purely thermal release during reactor operations may have been lower than 30 % and would be located close to the graphite free surfaces. Consequently, most of the tritium inventory after reactor shutdown could be trapped deeply within the irradiated graphite structure. Decontamination of graphite waste should then require temperatures higher than 1300°C, and would be more efficient in dry inert gas than in humid gas

Au cours des prochaines années, neuf centrales nucléaires de type UNGG (Uranium Naturel Graphite Gaz) devront être démantelées en France. Ces centrales utilisent le graphite comme modérateur et réflecteur de neutrons. Pendant leur exploitation, celui-ci est activé. Leur démantèlement conduira à 23000 tonnes de déchets de graphite irradiés à gérer. Ce travail focalise sur deux radionucléides contenus dans ces déchets : le 36Cl et le 3H. Le 36Cl a l'une des demi-vies les plus longues (301 000 ans). Par contre, le 3H a une demi-vie plus courte (12 ans), mais contribue beaucoup à l'activité initiale des déchets. Différentes données expérimentales suggèrent que le 36Cl et le 3H sont piégés à différents endroits du graphite, comme les boucles de dislocation, les surfaces ou les joints de grains. Le seul mécanisme de migration des radionucléides est le relâchement. Pour cette raison, il est important de comprendre quels sont les pièges et les différentes conditions du relâchement.Le graphite UNGG a une structure complexe, hétérogène et multi-échelle qui diffère du monocristal idéal du graphite. Cependant, pour comprendre les données macroscopiques, les études théoriques à l'échelle nanoscopique et microscopique sont des outils importants, même si elles reposent sur des modèles plus simples. Dans cette thèse, une approche multi-échelle a été utilisée afin d’étudier les interactions des radionucléides avec le graphite ainsi que les mécanismes de diffusion et de piégeage à l'échelle du nm-μm.Les interactions du 3H et du 36Cl avec différents défauts du graphite ont été étudiées dans le cadre de la théorie fonctionnelle de la densité (DFT). L'hydrogène forme une liaison covalente avec le graphite massique ainsi qu'avec ses surfaces (001), (100) et (110). Plusieurs reconstructions de surface ont été explorées. Les résultats montrent que les hypothèses existantes sur le piégeage de l'hydrogène doivent être affinées. Le comportement du Cl est plus complexe. Sa chimisorption est observée sur les surfaces (100) et (110). Cependant, sur la surface (001), le Cl interagit par transfert de charge. Le Cl2 n'interagit que par interactions de van der Waals avec celle-ci. Le Cl2 se dissocie dans le graphite massique.Les diffusions du H et du Cl dans le graphite irradié ont été étudiées en effectuant des simulations de dynamique moléculaire. Les résultats ab initio ont été utilisés pour développer des potentiels de type « bond order » afin de modéliser l'interaction des radionucléides avec la matrice de graphite, qui possède des contributions à court et à long portée. Pour le Cl, un nouveau potentiel a été paramétré qui reproduit toutes les données obtenues au niveau DFT. Pour les interactions 3H-graphite, les potentiels AIREBO/M, pour les interactions C-H, et LCBOP, pour les interactions C-C, ont été utilisés.Pour évaluer l'influence de la structure complexe du graphite UNGG sur le comportement des radionucléides, plusieurs modèles atomiques ont été utilisés pour rendre compte de cette diversité, tels que les surfaces, les joints de grains et les nanopores.Pour le Cl, des simulations d'irradiation ont été réalisées pour une gamme d’énergie allant de 1 à 10 keV et une gamme de température de 200 à 500ºC. Les dépendances à la température et à la direction d'irradiation ont été étudiées. D’une façon générale, les dommages causés par l'irradiation perpendiculaire aux surfaces augmentent avec la température. L'irradiation à des angles d’incidence <90º aux surfaces peut causer plus ou moins de dommages par rapport à l'irradiation perpendiculaire selon le type de surface.Les diffusions du H et du Cl montrent que tous les bords de cristallites avec des liaisons pendantes sont des pièges. Pour le Cl, la diffusion dans le graphite nanoporeux a révélé deux emplacements préférés: les bords des cristallites où le Cl forme une liaison covalente et les coins des microfissures où le Cl interagit par transfert de charge. / In the upcoming years, nine nuclear UNGG (Uranium Naturel Graphite Gaz) power plants will have to be dismantled in France. In these power plants, nuclear graphite was used as a neutron moderator and reflector, and was activated during operation. The dismantlement will lead to 23000 tons of irradiated graphite waste, which will have to be managed. The graphite is classified as a nuclear waste containing radionuclides with low activity and long half-life. Two radionuclides are the focus of this work: 36Cl and 3H. 36Cl has one of the longest half-lives (about 301000 years) among the waste's radionuclides. 3H has a shorter half-life (12 years), but contributes significantly to the waste’s initial activity. Previous experiments suggest that both, 36Cl and 3H, are mainly fixed at different traps in graphite, which are defective structures, such as dislocation loops, surfaces, or grain boundaries. Since the only significant migration mechanism of these radionuclides is release, it is important to understand where the traps are located and the conditions of the release.UNGG graphite has a complex heterogeneous multi-scale structure which differs substantially from an ideal monocrystal of graphite. However, in order to understand macroscopic data, theoretical studies at the nano- and microscopic scale are an important tool to explain underlying phenomena even though they rely on simpler models due to the limitations of computation power. A multi-scale approach was therefore applied to study the local interactions of the radionuclides with graphite as well as diffusion and trapping mechanisms on the nm-μm length scale.First, the interaction of 3H and 36Cl with defects in graphite was studied with density functional theory (DFT). Hydrogen interacts covalently with bulk graphite as well as with the studied surfaces (001), (100), and (110). Several surface reconstructions were investigated: arch-type reconstructions and in-plane reconstructions. The results show that the existing hypothesis on the trapping of hydrogen needs to be refined. The behavior of Cl is more complex. On the (100) and (110) surface chemisorption is observed. However, on the (001) surface a strong charge transfer interaction is observed for Cl. In contrast to that, Cl2 only interacts via weak van der Waals interactions with this surface. In bulk graphite Cl2 dissociates.The diffusion of H and Cl in irradiated graphite has been investigated by performing molecuar dynamics simulations. The ab initio results were used to develop bond order potentials to model the interaction of radionuclides and the graphite matrix, which attributes for short and long range interactions. For Cl, a new potential has been parameterized which is able to describe all aspects obtained with DFT. For the 3H-graphite interactions, the bond order potential AIREBO/M was used for C-H interactions. For C-C interactions the LCBOP potential was used.To evaluate the influence of the complex heterogeneous structure of the UNGG graphite on the radionuclide's behavior, several different atomic models were studied to account for this diversity such as surfaces, grain boundaries and nanopores.For Cl, irradiation simulations of different systems were performed up to an energy of 10 keV for the primary knock-on atom (PKA), and in a temperature range of 200 to 500ºC. The dependence on temperature and irradiation direction was investigated. In general, direct irradiation damage increases with temperature. Irradiation at incident angles <90º can create more or less damage compared to the perpendicular one depending on the surface type.Diffusion of H and Cl along surfaces shows that all crystallite edges with dangling bonds can serve as traps. For Cl, diffusion in nanoporous graphite revealed two preferred locations : First, the crystallite edges where Cl forms strong covalent; second, the corners of microcracks where Cl interacts via charge transfer.

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