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 concept for a compact DEMO reactor named 'SlimCS' is presented. Distinctive features of the concept are low aspect ratio (A = 2.6) and use of a reduced-size centre solenoid (CS) which has the function of plasma shaping rather than poloidal flux supply. The reduced-size CS enables us to introduce a thin toroidal field coil system which contributes to reducing the weight and perhaps lessening the construction cost. Low-A has merits of vertical stability for high elongation (κ) and high normalized beta (β N ), which leads to a high power density with reasonable physics requirements. This is because high κ facilitates high n GW (because of an increase in I p ), which allows efficient use of the capacity of high β N . From an engineering aspect, low-A may ensure ease in designing blanket modules robust to electromagnetic forces acting on disruptions. Thus, a superconducting low-A tokamak reactor such as SlimCS can be a promising DEMO concept with physics and engineering advantages.

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Halo currents (Ih) and their toroidal peaking factors (TPFs) have been measured in JT-60U by Rogowski coil type halo current sensors. The maximum value of TPF × Ih/Ip0 was 0.52 in the operational range of Ip0 = 0.7-1.8 MA, BT = 2.2-3.5 T, including ITER design parameters of κ> 1.6 and q95 = 3, which was lower than that of the maximum value of the ITER database (0.75). This value of TPF × Ih/Ip0 correlated with the driving forces of the halo currents generated by the vertical shift velocity of the plasma. The magnitude of the halo currents tended to decrease with the increase in stored energy just before the energy quench and with the line integrated electron density at the time of maximum halo current. A termination technique in which the current channel remains stationary demonstrated how to avoid halo current generation. Strong pulsed gas puffing during vertical displacement events was effective in reducing the halo currents.

Vertical displacement events (VDEs) during plasma current quench (Irho quench) are one of the serious problems encountered in designing tokamak fusion reactors, owing to the generation of enormously high electromagnetic forces on the vacuum vessel and in-vessel components, but they have been passively and actively avoided in JT-60U. In JT-60U `slow IP quench` is ended with very fast plasma current termination (final Irho termination), and the halo current is frequently measured at this final Irho termination. VDEs make the final Irho termination severe by increasing the halo current and the electromagnetic force. A strong dependence of VDE growth rate on the initial vertical position of the plasma current centre (ZJ) has been clarified experimentally, and a neutral point of ZJ for VDE has been found at ~15 cm above the midplane of the vacuum vessel. According to these measurements, VDE has been avoided by the selection of ZJ at the start of Irho quench (passive control) and by the control of ZJ during Irho quench (active control) eventually obtained owing to the small deviation of ZJ in real time calculations from its actual value. Furthermore, passive avoidance of VDEs by the injection of a neon ice pellet has been demonstrated