This paper presents the first control algorithm for the inner- and outer-strike point position for a Spherical Torus (ST) fusion experiment and the performance analysis of the controller. A liquid lithium divertor (LLD) will be installed on NSTX which is believed to provide better pumping than lithium coatings on carbon PFCs. The shape of the plasma dictates the pumping rate of the lithium by channelling the plasma to LLD, where the strike point location is the most important shape parameter. Simulations show that the density reduction depends on the proximity of the strike point to LLD. Experiments were performed to study the dynamics of the strike point, design a new controller to change the location of the strike point to the desired location and stabilize it. The most effective poloidal field (PF) coils in changing inner- and outer-strike points were identified using equilibrium code. The PF coil inputs were changed in a step fashion between various set points and the step response of the strike point position was obtained. From the analysis of the step responses, proportional–integral–derivative controllers for the strike points were obtained and the controller was tuned experimentally for better performance. The strike controller was extended to include the outer-strike point on the inner plate to accommodate the desired low outer-strike points for the experiment with the aim of achieving 'snowflake' divertor configuration in NSTX.
Plasma boundary shape control and real-time equilibrium reconstruction on NSTX-U
A model-based strike point sweeping technique applied to JET