One of the major challenges for the design and operation of next-step high-power steady-state fusion devices is to develop and validate advanced divertor solutions for handling power exhaust, while maintaining acceptable divertor target plate erosion. This requires to access divertor detachment at relatively low main plasma densities compatible with current drive and high plasma confinement. Recently a new initiative has been launched on the EAST superconducting tokamak to develop a new divertor for evaluating boundary plasma solutions applicable to the next step fusion experiments beyond ITER. In the present work, four different divertor configurations with different degrees of closure are evaluated through detailed modeling with the SOLPS plasma boundary code. The modeling results show that increasing divertor closure can significantly trap more neutrals from to with the same upstream separatrix density and hence facilitate the onset of detachment decreasing from to Moreover, with increasing divertor closure the divertor radiated power is also increased from 200 to 450 kW and the peak heat flux density at the divertor target is reduced from 9.3 to 3.8 MW m−2. Therefore, increasing divertor closure can benefit divertor operation for EAST and may be used as one of the scientific metrics for the divertor design of future fusion devices.
Effect of divertor geometry on boundary and core plasma performance in ASDEX Upgrade and JET