Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document Nucl. Fusion 39 2137-2664, is reviewed. Recent theoretical and experimental research has made important advances in both understanding and control of MHD stability in tokamak plasmas. Sawteeth are anticipated in the ITER baseline ELMy H-mode scenario, but the tools exist to avoid or control them through localized current drive or fast ion generation. Active control of other MHD instabilities will most likely be also required in ITER. Extrapolation from existing experiments indicates that stabilization of neoclassical tearing modes by highly localized feedback-controlled current drive should be possible in ITER. Resistive wall modes are a key issue for S128 Chapter 3: MHD stability, operational limits and disruptions advanced scenarios, but again, existing experiments indicate that these modes can be stabilized by a combination of plasma rotation and direct feedback control with non-axisymmetric coils. Reduction of error fields is a requirement for avoiding non-rotating magnetic island formation and for maintaining plasma rotation to help stabilize resistive wall modes. Recent experiments have shown the feasibility of reducing error fields to an acceptable level by means of non-axisymmetric coils, possibly controlled by feedback. The MHD stability limits associated with advanced scenarios are becoming well understood theoretically, and can be extended by tailoring of the pressure and current density profiles as well as by other techniques mentioned here. There have been significant advances also in the control of disruptions, most notably by injection of massive quantities of gas, leading to reduced halo current fractions and a larger fraction of the total thermal and magnetic energy dissipated by radiation. These advances in disruption control are supported by the development of means to predict impending disruption, most notably using neural networks. In addition to these advances in means to control or ameliorate the consequences of MHD instabilities, there has been significant progress in improving physics understanding and modelling. This progress has been in areas including the mechanisms governing NTM growth and seeding, in understanding the damping controlling RWM stability and in modelling RWM feedback schemes. For disruptions there has been continued progress on the instability mechanisms that underlie various classes of disruption, on the detailed modelling of halo currents and forces and in refining predictions of quench rates and disruption power loads. Overall the studies reviewed in this chapter demonstrate that MHD instabilities can be controlled, avoided or ameliorated to the extent that they should not compromise ITER operation, though they will necessarily impose a range of constraints.

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.

Recent results on investigations of Alfvén eigenmodes, fast ion confinement and fast ion diagnostics in JT-60U are presented. It was found that toroidicity induced Alfvén eigenmodes (TAEs) were stable in negative shear discharges with a large density gradient at the internal transport barrier (ITB). If the density gradient was small at the ITB, multiple TAEs appeared around the q = 2 surface (pitch minimum) and showed a large frequency chirping (∆f ≈ 80 kHz). In low-q positive shear discharges, the location of the TAEs changed from outside to inside the q = 1 surface, owing to a temporal change of the q profile. A significant depression of the megaelectronvolt ion population was observed only with high-n (n up to 14) multiple TAEs inside the q = 1 surface. Non-circular triangularity induced Alfvén eigenmodes were observed for the first time. Considerable depression of the triton burnup was observed in negative shear discharges. Orbit following Monte Carlo simulations indicated that ripple loss was responsible for the enhanced triton losses. The fast ion stored energies in ICRF heated negative shear discharges were comparable to those of positive shear plasmas. Tail ion temperatures in high (second to fourth) harmonic ICRF heating experiments were first analysed with an MeV neutral particle analyser. The behaviour of MeV ions produced by ICRF heating was studied with gamma ray diagnostics. A scintillating fibre detector system for detecting the 14 MeV neutron emission was developed for the triton burnup studies. Ion cyclotron emission measurements discriminating between parallel and perpendicular components of the electric field were carried out for the first time.

A tearing mode with m = 3 and n = 2, destabilized in the steady state high-β p H-mode discharges with edge localized mode (ELM), was completely stabilized by local heating and current drive using the 110 GHz first harmonic O-mode electron cyclotron (EC) wave. Here, m and n are poloidal and toroidal mode numbers, respectively. The optimum EC wave injection angle was determined by identifying the mode location from an electron temperature perturbation profile and a safety factor profile. The optimum injection angle was also determined by scanning a steerable mirror during a discharge. In a typical discharge where the tearing mode is completely stabilized, the ratio of the electron cyclotron heating power to the total heating power is 0.17, and the ratio of the EC driven current to the total plasma current is 0.02. Stored energy and neutron emission rate were higher for the case with EC wave injection than that without EC wave injection, which suggests that the reduction of the stored energy and the neutron emission rate was recovered by the tearing mode stabilization.

This paper reports results on the progress in steady-state high-βp ELMy H-mode discharges in JT-60U. A fusion triple product, nD(0)τETi(0), of 3.1 × 1020 m−3 s keV under full non-inductive current drive has been achieved at Ip = 1.8 MA, which extends the record value of the fusion triple product under full non-inductive current drive by 50%. A high-beta plasma with βN ∼ 2.7 has been sustained for 7.4 s (∼60τE), with the duration determined only by the facility limits, such as the capability of the poloidal field coils and the upper limit on the duration of injection of neutral beams. Destabilization of neoclassical tearing modes (NTMs) has been avoided with good reproducibility by tailoring the current and pressure profiles. On the other hand, a real-time NTM stabilization system has been developed where detection of the centre of the magnetic island and optimization of the injection angle of the electron cyclotron wave are done in real time. By applying this system, a 3/2 NTM has been completely stabilized in a high-beta region (βp ∼ 1.2, βN ∼ 1.5), and the beta value and confinement enhancement factor have been improved by the stabilization.