Independent thesis Advanced level (degree of Master (Two Years)), 20 credits / 30 HE credits
Retention will be one of the key issues in ITER. Retention is associated to the effects of plasma species being trapped in and on the tokamak wall. Trapping of plasma species inside the wall material is a problem of material science, which mainly deals with material structure and various type lattice defects. Reten- tion is a crucial factor in such eects: mechanical wall degradation, fuel man- agement, contamination assessment. In this study tungsten W, as a propective material for ITER design, is the material of our interest.
Retention in non-radioactive environment exists as an outcome of presence of innate defects, such as dislocations and grain boundaries (GBs) formed during manufacturing. However in tokamaks retention must be considered in pres- ence of neutron irradiation, because the neutron irradiation is a catalysator of retention gain in tungsten. Highly energetic neutron (14 MeV) ux penetrates tungsten bulk causing damage to the tungsten lattice. The result of this is numerous self-interstitial (SIA) and vacancy clusters of various con gurations built into tungsten bulk, which increase plasma trapping. The knowledge about retention mechanisms will lead to better understand-ing of embrittlement, wear of wall materials and TDS (thermal desorption spectroscopy) data.
Numerical simulations is great means to observe and investigate irradiation process and its dynamics at atomic scales, which is impossible to be observed experimentally. In this study collision/displacements cascades were simulated, which provided us with information about dynamics of a neutron-tungsten collisions and let observe the formation of irradiation defects. Numerical sim-ulations were as well used in construction of various formations of defects, investigation of trapping mechanisms and interactions with implanted plasma species.
The main objective in this study was to provide a comparison between the two origins of defects: pre-existing and n-induced defects. The interest was to explore, which type is a dominating type of defects in terms of imposing e ect on plasma species. At rst it was found that energetically both types have the same eect on plasma species. Secondly it was found that spatial range of in-uence is not determined by the origin of defect. Distinct defects would rather couple themselves to defects of other origin in terms of interaction range: GB with vacancy cluster, edge dislocation with SIA cluster. And the last objec- tive was to determine, whether there is any distinction between H and He in term of interaction with defects. Generally H and He would always give the same, highly identical data, however there were one distinct occasion, where substantial dierence between H and He was observed in interaction patterns. The work consists of three parts: characterization of the primary damage state in terms of morphology and size distributions of irradiation defects, character-ization of extended lattice defects such as dislocations and grain boundaries and their interaction with H & He atoms, characterization of n-induced defects and their interaction with H & He atoms.