Tungsten is a candidate plasma facing material for next generation magnetic fusion devices such as ITER and there are major operational and safety issues associated with hydrogen (tritium) retention in plasma facing components. An ion gun was used to simulate plasmamaterial interactions under various conditions in order to study hydrogen retention characteristics of tungsten thus enabling better predictions of hydrogen retention in ITER. Thermal Desorption Spectroscopy (TDS) was used to measure deuterium retention from ion irradiation while modelling of TDS spectra with the Tritium Migration Analysis Program (TMAP) was used to provide information about the trapping mechanisms involved in deuterium retention in tungsten. X-ray Photoelectron Spectroscopy (XPS) and Secondary Ion Mass Spectrometry (SIMS) were used to determine the depth resolved composition of specimens used for irradiation experiments. Carbon and oxygen atoms will be among the most common contaminants within ITER. C and O contamination in polycrystalline tungsten (PCW) specimens even at low levels (~0.1%) was shown to reduce deuterium retention by preventing diffusion of deuterium into the bulk of the specimen. This diffusion barrier was also responsible for the inhibition of blister formation during irradiations at 500 K. These observations may provide possible mitigation techniques for iii problems associated with tritium retention and mechanical damage to plasma facing components caused by hydrogen implantation. Deuterium trapping in PCW and single crystal tungsten (SCW) was studied as a function of ion energy and substrate temperature. Deuterium retention was shown to decrease with decreasing ion energy below 100 eV/D+. Irradiation of tungsten specimens with 10 eV/D+ ions was shown to retain up to an order of magnitude less deuterium than irradiation with 500 eV/D+ ions. Furthermore, the retention mechanism for deuterium was shown to be consistent across the entire energy range studied (10-500 eV) with the shallow penetration depth of low energy ions being the major factor in the reduction in retention. A change in retention mechanism was observed as tungsten temperature during irradiation was increased from 300 to 500 K. Modelling of deuterium retention in 300 and 500 K SCW specimens revealed that two traps, 1.0 and 1.3 eV, are involved in retention for irradiations performed at 300K while a single 2.1 eV trap is present for 500 K irradiations. Experiments suggest that the 2.1 eV trap is created during irradiation of tungsten at 500 K and this process also involves the annihilation of the 1.3 and 1.0 eV traps.

Tungsten is a candidate plasma facing material for next generation magnetic fusion devices such as ITER and there are major operational and safety issues associated with hydrogen (tritium) retention in plasma facing components. An ion gun was used to simulate plasmamaterial interactions under various conditions in order to study hydrogen retention characteristics of tungsten thus enabling better predictions of hydrogen retention in ITER. Thermal Desorption Spectroscopy (TDS) was used to measure deuterium retention from ion irradiation while modelling of TDS spectra with the Tritium Migration Analysis Program (TMAP) was used to provide information about the trapping mechanisms involved in deuterium retention in tungsten. X-ray Photoelectron Spectroscopy (XPS) and Secondary Ion Mass Spectrometry (SIMS) were used to determine the depth resolved composition of specimens used for irradiation experiments. Carbon and oxygen atoms will be among the most common contaminants within ITER. C and O contamination in polycrystalline tungsten (PCW) specimens even at low levels (~0.1%) was shown to reduce deuterium retention by preventing diffusion of deuterium into the bulk of the specimen. This diffusion barrier was also responsible for the inhibition of blister formation during irradiations at 500 K. These observations may provide possible mitigation techniques for iii problems associated with tritium retention and mechanical damage to plasma facing components caused by hydrogen implantation. Deuterium trapping in PCW and single crystal tungsten (SCW) was studied as a function of ion energy and substrate temperature. Deuterium retention was shown to decrease with decreasing ion energy below 100 eV/D+. Irradiation of tungsten specimens with 10 eV/D+ ions was shown to retain up to an order of magnitude less deuterium than irradiation with 500 eV/D+ ions. Furthermore, the retention mechanism for deuterium was shown to be consistent across the entire energy range studied (10-500 eV) with the shallow penetration depth of low energy ions being the major factor in the reduction in retention. A change in retention mechanism was observed as tungsten temperature during irradiation was increased from 300 to 500 K. Modelling of deuterium retention in 300 and 500 K SCW specimens revealed that two traps, 1.0 and 1.3 eV, are involved in retention for irradiations performed at 300K while a single 2.1 eV trap is present for 500 K irradiations. Experiments suggest that the 2.1 eV trap is created during irradiation of tungsten at 500 K and this process also involves the annihilation of the 1.3 and 1.0 eV traps.

Le tungstène est le matériau choisi pour le divertor d'ITER en raison de ses propriétés thermiques et physiques. Les densités de flux et les énergies des particules, et la température de surface pourront varier de plusieurs ordres de grandeur le long de la surface du divertor, soit respectivement 1020-1024 m2s-1, 0.1-100 eV, et 370-1370K. Exposé à de telles conditions, le tungstène peut subir de l'érosion, des fissurations et d'autres modifications de surface affectant ses propriétés thermomécaniques. Une autre préoccupation est la rétention des atomes de tritium implantés dans la surface et leur diffusion dans l'épaisseur du matériau. Nombre d'études ont porté sur la rétention et les modifications de surface induite par le plasma, mais se concentraient principalement sur l'effet de l'énergie des ions, leur fluence et la température de surface, mais très peu portaient sur l'influence du flux de plasma, avec des résultats erratiques et peu cohérents entre eux. L'objectif de cette thèse est de fournir une image cohérente du comportement du tungstène exposé à des conditions pertinentes pour les futurs réacteurs de fusion. Une investigation systématique évaluant l'impact de la densité de flux de plasma et de la température de l'exposition sur les modifications de surface et l'accumulation d'hydrogène dans tungstène a été effectuée au moyen d'expériences menées dans les dispositifs plasma linéaires, PSI-2 à Jülich, Pilot-PSI et Magnum-PSI à DIFFER, et PISCES-A à UCSD. La corrélation entre densité de flux de particules, température d'exposition, modifications de surface et rétention de l'hydrogène dans le tungstène a été étudiée pour différentes microstructures de matériau. Des échantillons de trois types de tungstène poly-cristallin (traité thermiquement à 1273 et 2273 K) et monocristallin (orientation 110) ont été exposés à des plasmas de deutérium et des températures de surface de 530 à 1170 K à deux gammes différentes de flux d'ions deutérium (~1022 et ~1024 m2s-1). Toutes les expositions ont été effectuées à la même énergie d'ions incidents de 40 eV et une fluence de particules de 1026 m2. Les échantillons ont été analysés post mortem en utilisant diverses techniques d'imagerie et d'analyse de surface (microscopie, spectroscopie de désorption thermique (TDS) et analyse par faisceau d'ions). L'augmentation du flux de particules de deux ordres de grandeur a provoqué la formation de bulles au-dessus de 700 K, températures pour lesquelles elles sont généralement absentes à faible flux. De petites cloques de respectivement quelques dizaines de nm et jusqu'à une taille latérale de 1 micron ont été détectées sur des échantillons de tungstène poly-cristallin recuit et monocristallin. Au contraire, les cloques sont absentes sur les échantillons recristallisés, sauf à faible flux et basse température où des vésicules d'environ 10 micron et des cavités sont apparues le long des joints de grains. La rétention totale de deutérium a été mesurée par TDS. A faible température d'exposition la fraction retenue était un à deux ordres de grandeur plus élevée après l'exposition à faible flux qu'à haut flux. Au contraire, une tendance opposée de la rétention totale à des températures d'exposition élevées a été observée. Par conséquent, le maximum de la rétention de deutérium totale a été observé pour une température supérieure dans le cas du flux incident de particules élevée (~850 K) comparée à l'exposition de faible flux (~650 K). Globalement les résultats sur la rétention de deutérium étaient similaires pour toutes les microstructures de tungstène étudiées. La rétention diminue à haute température et son maximum est plus bas pour des expositions à haut flux. Or, en raison du décalage de la rétention maximale vers des températures plus élevées, la quantité de deutérium piégée pour des températures supérieures à 800 K était plus élevée à haut flux qu'à faible flux, soit ~ un ordre de grandeur inférieur à la rétention maximum à faible flux / Tungsten is the selected material to be used in the ITER divertor due to its favorable thermal and physical properties. Particle flux densities and energies, and surface temperature will vary by several orders of magnitude along the divertor surface, with values in the range 1020-1024 m2s-1, 0.1-100 eV and 370-1370 K, respectively. Exposed to such conditions, tungsten may undergo erosion, cracking and other surface modifications affecting its thermal and mechanical properties. Another concern is the retention of implanted radioactive fuel atoms (tritium) in the material surface and their diffusion through the bulk. A considerable amount of studies have addressed retention and plasma induced surface modifications, focusing mainly on the effect of ion energy, ion fluence and surface temperature while very little knowledge exists on the influence of the plasma flux. These results are largely scattered and occasionally bear a lack of consistency. The aim of this thesis is to provide a coherent picture of the behavior of tungsten exposed to plasma conditions relevant for future fusion reactors. A systematic investigation assessing the impact of the plasma flux density and exposure temperature on surface modifications and hydrogen accumulation in tungsten was performed by means of experiments carried out in the linear plasma devices PSI2 at Forschungszentrum Juelich, Pilot-PSI and Magnum-PSI at DIFFER, and PISCES-A at UCSD. The correlation between the particle flux density, exposure temperature, surface modifications and hydrogen retention in tungsten was investigated for different material microstructures. Three types of polycrystalline tungsten (thermally treated at 1273 and 2273 K) and single crystal tungsten samples (110 crystal orientation) were exposed to deuterium plasmas at surface temperatures of 530-1170 K to two different ranges of deuterium ion fluxes (low and high flux: ~1022 and ~1024 m2s-1). All the exposures were performed at the same incident ion energy of 40 eV and particle fluence of ~1026 m2. The exposed samples were analyzed postmortem utilizing various surface imaging and analyses techniques (microscopy, thermal desorption spectroscopy and ion beam analysis). Increasing the particle flux by two orders of magnitude caused blister formation at temperatures above 700 K for which blistering is usually absent under low flux exposure conditions. Small blisters of several tens of nanometers and up to 1 micrometer of lateral size were detected on the annealed polycrystalline and in single crystal tungsten samples, respectively. On the contrary, blisters were absent on the recrystallized samples except for the low flux and low temperature case where large blisters of about 10 micrometer and cavities along the grain boundaries appeared. The total deuterium retention was measured by means of thermal desorption spectroscopy (TDS). In the cases with low exposure temperatures, the retained fraction of deuterium was one to two orders of magnitude higher after exposure to the low flux compared to the high flux. On the contrary, an opposite tendency of the total deuterium retention at high exposure temperatures was observed. Hence, the maximum of the total deuterium retention was observed to occur at a higher temperature in the case of high incident particle flux (~850 K) compared to low flux exposures (~650 K). Overall, experimental results on deuterium retention were similar for all the investigated tungsten microstructures. Deuterium retention decreased at high temperatures and the maximal retention was lower for high flux exposures. However, due to the shift of the maximal retention to higher temperatures, the amount of deuterium retained at temperatures above 800 K was higher at high flux rather than at low flux, being still about one order of magnitude lower than the maximal retention at low flux

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>