Chapter 8: Plasma operation and control

The ITER plasma control system has the same functional scope as the control systems in present tokamaks. These are plasma operation scenario sequencing, plasma basic control (magnetic and kinetic), plasma advanced control (control of RWMs, NTMs, ELMs, error fields, etc) and plasma fast shutdown. This chapter considers only plasma initiation and plasma basic control. This chapter describes the progress achieved in these areas in the tokamak experiments since the ITER Physics Basis (1999 Nucl. Fusion 39 2577) was written and the results of assessment of ITER to provide the plasma initiation and basic control. This assessment was done for the present ITER design (15 MA machine) at a more detailed level than it was done for the ITER design 1998 (21 MA machine) described in the ITER Physics Basis (1999 Nucl. Fusion 39 2577). The experiments on plasma initiation performed in DIII-D and JT-60U, as well as the theoretical studies performed for ITER, have demonstrated that, within specified assumptions on the plasma confinement and the impurity influx, ITER can produce plasma initiation in a low toroidal electric field (0.3 V m−1), if it is assisted by about 2 MW of ECRF heating. The plasma basic control includes control of the plasma current, position and shape—the plasma magnetic control, as well as control of other plasma global parameters or their profiles—the plasma performance control. The magnetic control is based on more reliable and simpler models of the control objects than those available at present for the plasma kinetic control. Moreover the real time diagnostics used for the magnetic control in many cases are more precise than those used for the kinetic control. Because of these reasons, the plasma magnetic control was developed for modern tokamaks and assessed for ITER better than the kinetic control. However, significant progress has been achieved in the plasma performance control during the last few years. Although the physics basis of plasma operation and control is similar in ITER and present tokamaks, there is a principal qualitative difference. To minimize its cost, ITER has been designed with small margins in many plasma and engineering parameters. These small margins result in a significantly narrower operational space compared with present tokamaks. Furthermore, ITER operation is expensive and component damage resulting from purely operational errors might lead to a high and avoidable repair cost. These factors make it judicious to use validated plasma diagnostics and employ simulators to ‘pre-test’ the combined ITER operation and control systems. Understanding of how to do this type of pre-test validation is now developed in present day experiments. This research push should provide us with fully functional simulators before the first ITER operation. PACS numbers: 28.52.−s, 52.55.Fa, 52.55.−s (Some figures in this article are in colour only in the electronic version)