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.

Recent QH-mode research on DIII-D [J. L. Luxon et al., Plasma Physics and Controlled Nuclear Fusion Research 1996 (International Atomic Energy Agency, Vienna, 1987), Vol. I, p. 159] has used the peeling-ballooning modes model of edge magnetohydrodynamic stability as a working hypothesis to organize the data; several predictions of this theory are consistent with the experimental results. Current ramping results indicate that QH modes operate near the edge current limit set by peeling modes. This operating point explains why QH mode is easier to get at lower plasma currents. Power scans have shown a saturation of edge pressure with increasing power input. This allows QH-mode plasmas to remain stable to edge localized modes (ELMs) to the highest powers used in DIII-D. At present, the mechanism for this saturation is unknown; if the edge harmonic oscillation (EHO) is playing a role here, the physics is not a simple amplitude dependence. The increase in edge stability with plasma triangularity predicted by the peeling-ballooning theory is consistent with the substantial improvement in pedestal pressure achieved by changing the plasma shape from a single null divertor to a high triangularity double null. Detailed ELITE calculations for the high triangularity plasmas have demonstrated that the plasma operating point is marginally stable to peeling-ballooning modes. Comparison of ELMing, coinjected and quiescent, counterinjected discharges with the same shape, current, toroidal field, electron density, and electron temperature indicates that the edge radial electric field or the edge toroidal rotation are also playing a role in edge stability. The EHO produces electron, main ion, and impurity particle transport at the plasma edge which is more rapid than that produced by ELMs under similar conditions. The EHO also decreases the edge rotation while producing little change in the edge electron and ion temperatures. Other edge electromagnetic modes also produce particle transport; this includes the incoherent, broadband activity seen at high triangularity. Pedestal values of ν* and βT bracketing, those required for International Experimental Thermonuclear Reactor [Nucl. Fusion 39, 2137 (1999)] have been achieved in DIII-D, demonstrating the QH-mode edge densities are sufficient for future devices.

We report the results of the first experiments on the DIII-D tokamak to examine the dependence of the transport and stability characteristics of ITER hybrid scenario plasmas on the toroidal flow (or rotation) of the plasma. With the new DIII-D capability to independently vary the neutral beam torque and power, the central rotation has been reduced by as much as a factor of 4.6 compared with discharges with unidirectional beams. Although energy confinement decreases and the m/n = 3/2 NTM amplitude increases for low rotation speed, the fusion performance figure of merit, , still exceeds the value required on ITER for Qfus = 10. These observations provide optimism about the projections of the hybrid scenario to low rotation plasmas in ITER, but they also indicate the need for a better understanding of the physics of toroidal rotation in order to project present-day results to future experiments.

Plasma discharges with negative triangularity (δ = −0.4) shape have been created in the DIII-D tokamak with significant normalized beta (βN = 2.7) and confinement characteristic of the high confinement mode (H98y2 = 1.2) despite the absence of an edge pressure pedestal and no edge localized modes (ELMs). These inner-wall-limited plasmas have similar global performance as a positive triangularity (δ = +0.4) ELMing H-mode discharge with the same plasma current, elongation and cross-sectional area. For cases both of dominant electron cyclotron heating with Te/Ti > 1 and dominant neutral beam injection heating with Te/Ti = 1, turbulent fluctuations over radii 0.5 < ρ < 0.9 were reduced by 10-50% in the negative triangularity shape compared to the matching positive triangularity shape, depending on radius and conditions.

The first suppression of the important and deleterious m=2/n= 1 neoclassical tearing mode (NTM) is reported using electron cyclotron current drive (ECCD) to replace the "missing" bootstrap current in the island 0-point. Experiments on the DIII-D tokamak verify that maximum shrinkage of the m=2/n=l island occurs when the ECCD location coincides with the q=2 surface. The DIII-D plasma control system is put into "search and suppress" mode to make small changes in the toroidal field to find and lock onto the optimum position, based on real time measurements of dBddt, for complete m=2/n=l NTM suppression by ECCD. The requirements on the ECCD for complete island suppression are well modeled by the modified Rutherford equation for the DIII-D plasma conditions. ...Recently several tokamaks have demonstrated the suppression of the m = 3/n = 2 NTM using unmodulated ECCD positioned at the island location. On ASDEX Upgrade, co-ECCD was verified to be more effective at NTM stabilization than counter-ECCD or pure heating alone [7-91. These experiments used a programmed sweep of the toroidal magnetic field to ensure that the current drive layer matched the mode location at some time during the ECCD pulse. On JT-GOU, suppression of the m = 3/n = 2 mode was also achieved for 1.5 s in steady conditions using ECCD