The effect of the poloidal divertor neutralizer geometry on the scrape-off plasma temperature and density

Abstract

The poloidal magnetic divertor operating in the cool-high density plasma mode represents a good solution to the tokamak first wall erosion and impurity transport problems. It also, in general, improves the energy confinement time in the beam-heated discharges. In the existing experiments, the neutralizer plates are oriented close to normal to the flux surfaces in the poloidal plane, but the planned high-power experiments require that the plate be sharply angled to reduce the power flux density. We have studied the effect of changing the plate angle on the radial plasma temperature and density profiles at the plate. We find that some plate orientations shift the electron flux profile relative to the power flux profile, resulting in unfavorable temperature profiles. The question of existence of a steady-state solution was also addressed. For a steady state to exist, the radial neutral fluxes have to exactly cancel the radial diffusive ion fluxes. Depending on the diffusion coefficients, the plasma density may have to increase with radius to satisfy this condition. The implications of these conclusions for the INTOR, and CIT experiments is also discussed.