No abstract

An integrated plasma profile control strategy, ARTAEMIS, is being developed for extrapolating present-day advanced tokamak (AT) scenarios to steady-state operation. The approach is based on semi-empirical modelling and was initially explored on JET (Moreau et al 2008 Nucl. Fusion 48 106001). This paper deals with the general applicability of this strategy for simultaneous magnetic and kinetic control on various tokamaks. The determination of the device-specific, control-oriented models that are needed to compute optimal controller matrices for a given operation scenario is discussed. The methodology is generic and can be applied to any device, with different sets of heating and current drive actuators, controlled variables and profiles. The system identification algorithms take advantage of the large ratio between the magnetic and thermal diffusion time scales and have been recently applied to both JT-60U and DIII-D data. On JT-60U, an existing series of high bootstrap current (∼70%), 0.9 MA non-inductive AT discharges was used. The actuators consisted of four groups of neutral beam injectors aimed at perpendicular injection (on-axis and off-axis), and co-current tangential injection (also on-axis and off-axis). On DIII-D, dedicated system identification experiments were carried out in the loop voltage (V ext) control mode (as opposed to current control) to avoid feedback in the response data from the primary circuit. The reference plasma state was that of a 0.9 MA AT scenario which had been optimized to combine non-inductive current fractions near unity with 3.5 < βN < 3.9, bootstrap current fractions larger than 65% and H 98(y,2) = 1.5. Actuators other than V ext were co-current, counter-current and balanced neutral beam injection, and electron cyclotron current drive. Power and loop voltage modulations resulted in dynamic variations of the plasma current between 0.7 and 1.2 MA. It is concluded that the response of essential plasma parameter profiles to specific actuators of a given device can be satisfactorily identified from a small set of experiments. This provides, for control purposes, a readily available alternative to first-principles plasma modelling.

No abstract

In a magnetic fusion reactor, the achievement of a certain type of plasma current profiles, which are compatible with magnetohydrodynamic stability at high plasma pressure, is key to enable high fusion gain and non-inductive sustainment of the plasma current for steady-state operation. The approach taken toward establishing such plasma current profiles at the DIII-D tokamak is to create the desired profile during the plasma current ramp-up and early flattop phases. The evolution in time of the current profile is related to the evolution of the poloidal flux, which is modeled in normalized cylindrical coordinates using a partial differential equation usually referred to as the magnetic diffusion equation. The control problem is formulated as an open-loop, finite-time, optimal control problem for a nonlinear distributed parameter system, and is approached using extremum seeking. Simulation results, which demonstrate the accuracy of the considered model and the efficiency of the proposed controller, are presented.

Key plasma physics and real-time control elements needed for robustly stable operation of high fusion power discharges in ITER have been demonstrated in US fusion research. Optimization of the current density profile has enabled passively stable operation without n " 1 tearing modes in discharges simulating ITER's baseline scenario with zero external torque. Stable rampdown of the discharge has been achieved with ITER-like scaled current ramp rates, while maintaining an X-point configuration. Significant advances have been made toward real-time prediction of disruptions: machine learning techniques for prediction of disruptions have achieved 90% accuracy in offline analysis, and direct probing of ideal and resistive plasma stability using 3D magnetic perturbations has shown a rising plasma response before the onset of a tearing mode. Active stability control contributes to prevention of disruptions, including direct stabilization of resistive-wall kink modes in high-β discharges, forced rotation of magnetic islands to prevent wall locking, and localized heating/current drive to shrink the islands. These elements are being integrated into stable operating scenarios and a new event-handling system for off-normal events in order to develop the physics basis and techniques for robust control in ITER.