Abstract
This paper presents particle-in-cell simulations of the plasma behaviour in the vicinity of gaps in castellated plasma-facing components (PFCs). The point of interest was the test limiter of the TEXTOR tokamak, a PFC designed for studies of plasma-wall interactions, in particular, related to impurity transport and fuel retention.
Simulations were performed for various plasma conditions in the vicinity of the castellated surface, where the gaps can be either shaped or unshaped. It was observed that depending on plasma parameters the transport of plasma particles inside the gap can be either in potential- or geometry-dominated regimes.
The mechanisms responsible for the formation of a potential peak inside the poloidal gap and its consequences on plasma deposition profiles are discussed. A study of gap shaping was performed in order to validate its effectiveness.