Influence of surface conditions on hydrogen trapping in tungsten

Investigations of hydrogen isotopes and helium retention in plasma facing components (PFC) that are exposed to various plasma conditions are important for future fusion devices such as ITER and DEMO. Due to its favorable physical properties, tungsten (W) has been chosen as the plasma-facing material of the ITER divertor. In the deuterium/tritium (D/T) phase of ITER, W PFC will experience incident particles flux composed of hydrogen isotopes (HI), helium (He), impurities and neutrons. In particular, it has been found that He significantly affects W PFC near surface, with the formation of dislocation loops, bubbles, or even W-fuzz. Furthermore, since oxygen is a typical main impurity in tokamak vacuum vessels, W PFC can get oxidized due to the high surface temperatures during ITER operation. Such surface and bulk modifications of W PFC may significantly affect the retention and other physical properties of W leading to an increase of the T inventory in the tokamak. Also, the modification of retention properties of PFC may impact plasma edge physics due to the outgassing of cold HI molecules or He atoms. In this PHD thesis, laboratory experiments involving ion implantation and thermal desorption spectrometry (TDS) technique were used to investigate the fundamental retention properties of HI in W PFC due to: - The presence of an oxide layer grown on the polycrystalline W surfaces in ITER relevant conditions, - An impact of ITER relevant He irradiation on the W surface and near-surface layer. TDS measurements were coupled with observations of modification occurring on the surface, in the bulk and in the structure of various PCW samples. (author)