Lors de l’opération d’ITER, des flux importants d’isotopes d’hydrogène (HI) constituant le fuel interagissent avec les composants face au plasma (CFP) de la machine. Dans le cas du Tungstène (W) composant le divertor qui est la zone la plus exposée aux interactions plasma paroi, le flux incident est implanté et diffuse ensuite dans le corps du matériau entrainant un piégeage du fuel. Pour des raisons de sureté, l’inventaire de Tritium retenu dans les parois d’ITER est limité. De plus, le dégazage du fuel depuis les parois vers le plasma, lors des opérations plasma peut avoir un impact sur le contrôle global du plasma. Le but de cette thèse est d’abord de déterminer les paramètres de piégeages du fuel dans le W (énergies/températures de dépiégeage, concentrations de pièges) grâce à la modélisation de résultats expérimentaux. Ces simulations de résultats expérimentaux montrent que l’implantation d’HIs dans le W peut induire, sous certaines conditions, la formation de lacunes contenant des impuretés. En plus de ce piège induit par l’implantation d’ions, 2 pièges intrinsèques sont présents dans le W. Ces 3 pièges retiennent les HIs jusqu’à 700 K. Enfin, il est montré que le W endommagé par des ions lourds ou des neutrons contient des dislocations, des boucles de dislocations et des cavités retenant les HIs jusqu’à 1000 K.Après avoir déterminé ces paramètres de piégeages des HIs dans le W, la rétention des HIs durant l’opération d’ITER est estimée. Lors de cette opération, la température des CFP W atteint environ 1000 K. Les simulations montrent donc que la rétention dans les CFPs non endommagé est bien plus faible que dans le cas d’un CFP endommagé. / During ITER operation, important flux of Hydrogen Isotopes (HIs) constituting the fuel interact with the plasma facing components (PFC) of the machine. In the case of tungsten (W) making the divertor which is the most exposed area to the plasma wall interaction, the incident flux can be implanted and diffuse inside the bulk material inducing a trapping of the fuel. To safety issue, the tritium inventory retained in ITER’s PFC is limited. In addition, the outgassing of the fuel during plasma operation can impact the edge plasma control.The aim of this PhD project is first to determined relevant trapping parameters of the fuel in W (detrapping energies/temperatures and trap concentrations) by modelling experimental results. The simulations of experimental results shows that under specific condition, the HI implantation can induce the formation of mono-vacancies containing impurities. In addition to this induced trap, 2 intrinsic traps are present in W. This 3 traps retain HIs up to 700 K. Finally, it has been shown that the damaged W by heavy ions or neutrons contains dislocations, dislocation loops and cavities that can trap HIs up to 1000 K.After determining the fuel retention properties of W, the HIs retention during ITER operation is estimated. During this operation, the PFC temperature reaches around 1000 K so the simulations show that the damaged W retains much more HIs than the undamaged W.

Thermonuclear fusion may become an attractive future power source. The most promising of all fusion machine concepts is the tokamak. Despite decades of active research, still huge tasks remain before a fusion power plant can go online. One of these important tasks deals with the interaction between the fusion plasma and the reactor wall. This work focuses on how eroded wall materials of different origin and mass are transported in a tokamak device. Element transport can be examined by injection of certain species of unique and predetermined origin, so called tracers. Tracer experiments were conducted at the TEXTOR tokamak before its final shutdown. This offered an unique opportunity for studies of the wall and other internal components: For the first time it was possible to completely dismantle such a machine and analyse every single part of reactor wall, obtaining a detailed pattern of material migration. Main focus of this work is on the high-Z metals tungsten and molybdenum, which were introduced by WF6 and MoF6 injection into the TEXTOR tokamak in several material migration experiments. It is shown that Mo and W migrate in a similar way around the tokamak and that Mo can be used as tracer for W transport. It is further shown how other materials - medium-Z (Ni), low-Z (N-15 and F), fuel species (D) - migrate and get deposited. Finally, the outcome of dust sampling studies is discussed. It is shown that dust appearance and composition depends on origin, formation conditions and that it can originate even from remote systems like the NBI system. Furthermore, metal splashes and droplets have been found, some of them clearly indicating boiling processes. / <p>QC 20151203</p>