Model of real time ITER plasma position, shape and current control on base of DINA code

Poloidal field magnetic control in tokama k has the essential functions of stabilizing the vertical instability inherent in elongated cross-sections, and in keeping the plasma shape, position and current nearly reference even in presence of internal plasma disturbances such as sawteeth, minor disruptions, ELMs and other transients. Poloidal magnetic control comprises in adjusting of the tokamak equilibrium by use of the currents in the PF coils and must maintain the specified plasma current Ip ref , shape and position parameters during the full scenario including plasma current ramp up and shut down stages. Control of the plasma shape in ITER in the divertor configurations is realized by controlling the size of the gaps (g1-g6) between the separatrix and the plasma-facing components at six selected locations, as shown in Fig. 1 where a scheme of real time control of plasma position, shape and plasma current for ITER is presented. For that a real time reconstruction of plasma equilibrium (similar to real time XLOC [1]) with use of magnetic diagnostics (measurements of flux loops [l, poloidal magnetic fields Bp, currents in PF coils IPF and plasma current Ip) is necessary to estimate plasma shape and include it in feedback control.