Editing Plasma stability (section)

== Opportunities for Improving MHD Stability ==
{{unreferenced section|date=September 2018}}

=== Configuration ===
The configuration of the plasma and its confinement device represent an
opportunity to improve MHD stability in a robust way. The benefits of discharge shaping and low aspect ratio for ideal MHD stability have been clearly demonstrated in tokamaks and STs, and will continue to be investigated in experiments such as [[DIII-D]], [[Alcator C-Mod]], [[National Spherical Torus Experiment|NSTX]], and [[Mega Ampere Spherical Tokamak|MAST]]. New stellarator experiments such as [[National Compact Stellarator Experiment|NCSX]] (proposed) will test the prediction that addition of appropriately designed helical coils can stabilize ideal kink modes at high beta, and lower-beta tests
of ballooning stability are possible in HSX. The new ST experiments provide an opportunity to test predictions that a low aspect ratio yields improved stability to tearing modes, including neoclassical, through a large stabilizing “[[Glasser effect]]” term associated with a large Pfirsch-Schlüter current. Neoclassical tearing modes can be avoided by minimizing the bootstrap current in quasi-helical and quasi-omnigenous stellarator configurations. Neoclassical tearing modes are also stabilized with the appropriate relative signs of the bootstrap current and the magnetic shear; this prediction is supported by the absence of NTMs in central negative shear regions of tokamaks. Stellarator configurations such as the proposed NCSX, a quasi-axisymmetric stellarator design, can be created with negative magnetic shear and positive bootstrap current to achieve stability to the NTM. Kink mode stabilization by a resistive wall has been demonstrated in RFPs and tokamaks, and will be investigated in other configurations including STs (NSTX) and spheromaks (SSPX). A new proposal to stabilize resistive wall modes by a flowing liquid lithium wall needs further evaluation.

=== Internal Structure ===
Control of the internal structure of the plasma allows more active avoidance of MHD instabilities. Maintaining the proper [[current density]] profile, for example, can help to maintain stability to tearing modes. Open-loop optimization of the pressure and current density profiles with external heating and current drive sources is routinely used in many devices. Improved diagnostic measurements along with localized heating and current drive sources, now becoming available, will allow active feedback control of the internal profiles in the near future.
Such work is beginning or planned in most of the large tokamaks ([[Joint European Torus|JET]], [[JT-60|JT&ndash;60U]], [[DIII-D (tokamak)|DIII&ndash;D]], [[Alcator C-Mod|C&ndash;Mod]], and [[ASDEX Upgrade|ASDEX&ndash;U]]) using [[radio frequency|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 [[Alfvén wave|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&ndash;U and COMPASS-D with promising results, and are planned for next year {{clarify|reason=Which year?|date=October 2014}} in DIII&ndash;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&ndash;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&ndash;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&ndash;Mod, JT&ndash;60U, ASDEX&ndash;U, and DIII&ndash;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.