RF heating and current drive. Real-time analysis of profile data such as MSE current profile measurements and real-time identification of stability boundaries are essential components of profile control. Strong plasma rotation can stabilize resistive wall modes, as demonstrated in tokamak experiments, and rotational shear is also predicted to stabilize resistive modes. Opportunities to test these predictions are provided by configurations such as the ST, spheromak, and FRC, which have a large natural diamagnetic rotation, as well as tokamaks with rotation driven by neutral beam injection. The Electric Tokamak experiment is intended to have a very large driven rotation, approaching Alfvénic regimes where ideal stability may also be influenced. Maintaining sufficient plasma rotation, and the possible role of the RWM in damping the rotation, are important issues that can be investigated in these experiments. Feedback Control Active feedback control of MHD instabilities should allow operation beyond the “passive” stability limits. Localized RF current drive at the rational surface is predicted to reduce or eliminate neoclassical tearing mode islands. Experiments have begun in ASDEX–U and COMPASS-D with promising results, and are planned for next year in DIII–D. Routine use of such a technique in generalized plasma conditions will require real-time identification of the unstable mode and its radial location. If the plasma rotation needed to stabilize the resistive wall mode cannot be maintained, feedback stabilization with external coils will be required. Feedback experiments have begun in DIII–D and HBT-EP, and feedback control should be explored for the RFP and other configurations. Physics understanding of these active control techniques will be directly applicable between configurations. Disruption Mitigation The techniques discussed above for improving MHD stability are the principal means of avoiding disruptions. However, in the event that these techniques do not prevent an instability, the effects of a disruption can be mitigated by various techniques. Experiments in JT–60U have demonstrated reduction of electromagnetic stresses through operation at a neutral point for vertical stability. Pre-emptive removal of the plasma energy by injection of a large gas puff or an impurity pellet has been demonstrated in tokamak experiments, and ongoing experiments in C–Mod, JT–60U, ASDEX–U, and DIII–D will improve the understanding and predictive capability. Cryogenic liquid jets of helium are another proposed technique, which may be required for larger devices. Mitigation techniques developed for tokamaks will be directly applicable to other configurations. See also List of plasma physics articles List of hydrodynamic instabilities named after