H. Zohm

Chapter 3: MHD stability, operational limits and disruptions

Progress in the area of MHD stability and disruptions, since the publication of the 1999 ITER Physics Basis document (1999 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 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.

Edge localized modes (ELMs)

The phenomenology of edge localized modes (ELMs), an MHD instability occurring in the edge of H-mode plasmas in toroidal magnetic fusion experiments, is described. ELMs are important to obtain experimental control of the particle inventory of fusion plasmas. From an analysis of the ELM behaviour of different magnetic fusion experiments, three distinct types are identified, namely dithering cycles, type III and type I ELMs. A physical picture of these phenomena is established on the grounds of theoretical models put forward to describe the different ELM phenomena.

Beta limits in long-pulse tokamak discharges

The maximum normalized beta achieved in long-pulse tokamak discharges at low collisionality falls significantly below both that observed in short pulse discharges and that predicted by the ideal MHD theory. Recent long-pulse experiments, in particular those simulating the International Thermonuclear Experimental Reactor (ITER) [M. Rosenbluth et al., Plasma Physics and Controlled Nuclear Fusion (International Atomic Energy Agency, Vienna, 1995), Vol. 2, p. 517] scenarios with low collisionality nu(e)*, are often limited by low-m/n nonideal magnetohydrodynamic (MHD) modes. The effect of saturated MHD modes is a reduction of the confinement time by 10%-20%, depending on the island size and location, and can lead to a disruption. Recent theories on neoclassical destabilization of tearing modes, including the effects of a perturbed helical bootstrap current, are successful in explaining the qualitative behavior of the resistive modes and recent results are consistent with the size of the saturated islands. Also, a strong correlation is observed between the onset of these low-m/n modes with sawteeth, edge localized modes (ELM), or fishbone events. consistent with the seed island required by the theory. We will focus on a quantitative comparison between both the conventional resistive and neoclassical theories, and the experimental results of several machines, which have all observed these low-min nonideal modes. This enables us to single out the key issues in projecting the long-pulse beta limits of ITER-size tokamaks and also to discuss possible plasma control methods that can increase the soft beta limit, decrease the seed perturbations, and/or diminish the effects on confinement

Experiments on neoclassical tearing mode stabilization by ECCD in ASDEX Upgrade

The reduction of neoclassical tearing modes by ECCD is demonstrated experimentally. It is shown that with an averaged ECCD power of only 4-8% of the total heating power injected into the discharge, the island width can be reduced by 40%, provided that the centre of deposition is very close to the resonant surface. The reduction in mode amplitude results in a partial recovery of the loss of stored energy induced by the mode. This experimental result is well reproduced by modelling calculations.

Overview of the ITER EC upper launcher

The ITER electron cyclotron (EC) upper port antenna (or launcher) is nearing completion of the detailed design stage and the final build-to-print design stage will soon start. The main objective of this launcher is to drive current locally to stabilize the neoclassical tearing modes (NTMs) (depositing ECCD inside of the island that forms on either the q = 3/2 or 2 rational magnetic flux surfaces) and control the sawtooth instability (deposit ECCD near the q = 1 surface). The launcher should be capable of steering the focused beam deposition location to the resonant flux surface over the range in which the q = 1, 3/2 and 2 surfaces are expected to be found for various plasma equilibria susceptible to the onset of NTMs and sawteeth. The aim of this paper is to provide the design status of the principal components that make up the launcher: port plug, mm-wave system and shield block components. The port plug represents the chamber that provides a rigid support structure that houses the mm-wave and shield blocks. The mm-wave system comprises the components used to guide the RF beams through the port plug structure and refocus the beams far into the plasma. The shield block components are used to attenuate the nuclear radiation from the burning plasma, protecting the fragile in-port components and reducing the neutron streaming through the port assembly. The design of these three subsystems is described; in addition, the relevant thermo-mechanical and electro-magnetic analyses are reviewed for critical design issues.

Stabilization of neoclassical tearing modes by electron cyclotron current drive

The generalized Rutherford equation for the neoclassical tearing mode is studied. New analytical expressions for the nonlinear stability criterion, the seed island width, and the saturated island width are derived. These are especially useful when the saturated island width is small. A formalism for calculating the current needed to stabilize the mode is established by adding an externally driven current. Inserting the reference parameters of the International Thermonuclear Experimental Reactor (ITER) [ITER-JCT and Home Teams, Plasma Phys. Controlled Fusion 37, A19 (1995)], a value of 160 kA to be driven by Electron Cyclotron Current Drive (ECCD) in order to completely stabilize an m=2 mode is found, well within the capabilities of the ITER ECCD system. If higher currents can be driven, the local βp at the resonant surface can be increased significantly.The generalized Rutherford equation for the neoclassical tearing mode is studied. New analytical expressions for the nonlinear stability criterion, the seed island width, and the saturated island width are derived. These are especially useful when the saturated island width is small. A formalism for calculating the current needed to stabilize the mode is established by adding an externally driven current. Inserting the reference parameters of the International Thermonuclear Experimental Reactor (ITER) [ITER-JCT and Home Teams, Plasma Phys. Controlled Fusion 37, A19 (1995)], a value of 160 kA to be driven by Electron Cyclotron Current Drive (ECCD) in order to completely stabilize an m=2 mode is found, well within the capabilities of the ITER ECCD system. If higher currents can be driven, the local βp at the resonant surface can be increased significantly.

On the physics guidelines for a tokamak DEMO

The physics base for the ITER Physics Design Guidelines is reviewed in view of application to DEMO and areas are pointed out in which improvement is needed to arrive at a consistent set of DEMO Physics Design Guidelines. Amongst the proposed improvements, the area of power exhaust plays a crucial role since predictive capability of present-day models is low and this area is expected to play a major role in limiting DEMO designs due to the much larger value of Ptot/R in DEMO than in present-day devices and even ITER.

Impurity behaviour in the ASDEX Upgrade divertor tokamak with large area tungsten walls

At the central column of ASDEX Upgrade, an area of 5.5 m2 of graphite tiles was replaced by tungsten-coated tiles representing about two-thirds of the total area of the central column. No negative influence on the plasma performance was found, except for internal transport barrier limiter discharges. The tungsten influx ΓW stayed below the detection limit only during direct plasma wall contact or for reduced clearance in divertor discharges spectroscopic evidence for ΓW could be found. From these observations a penetration factor of the order of 1% and effective sputtering yields of about 10-3 could be derived, pointing to a strong contribution by light intrinsic impurities to the total \mbox{W-sputtering}. The tungsten concentrations ranged from below 10-6 up to a few times 10-5. Generally, in discharges with increased density peaking, a tendency for increased central tungsten concentrations or even accumulation was observed. Central heating (mostly) by ECRH led to a strong reduction of the central impurity content, accompanied by a very benign reduction of the energy confinement. The observations suggest that the W-source strength plays only an inferior role for the central W-content compared to the transport, since in the discharges with increased W-concentration neither an increase in the W-influx nor a change in the edge parameters was observed. In contrast, there is strong experimental evidence, that the central impurity concentration can be controlled externally by central heating.

H mode discharges with feedback controlled radiative boundary in the ASDEX Upgrade tokamak

Puffing of impurities (neon, argon) and deuterium gas in the main chamber is used to feedback control the total radiated power fraction and the divertor neutral particle density simultaneously in the ASDEX Upgrade tokamak. The variation of Psep=Pheat-Prad(core) by impurity radiation during H mode shows a similar effect on the ELM behaviour as that obtained by a change of the heating power. For radiated power fractions above 90%, the ELM amplitude becomes very small and detachment from the divertor plates occurs, whilst no degradation of the global energy confinement is observed (completely detached high confinement mode). Additional deuterium gas puffing is found to increase the radiated power per impurity ion in the plasma core owing to the combined effect of a higher particle recycling rate and a lower core penetration probability. The outer divertor chamber, which is closed for deuterium neutrals, builds up a high neutral pressure, the magnitude of which is determined by the balance of particle sources and pumping. For this particular situation, the effective pumping time of neon and argon is considerably reduced, to less than 0.3 s, mainly owing to an improved divertor retention capability. The radiation characteristics of discharges with a neon driven radiative mantle are modelled using a 1-D radial impurity transport code that has been coupled to a simple divertor model describing particle recycling and pumping. The results of simulations are in good agreement with experiment

On the Requirements to Control Neoclassical Tearing Modes in Burning Plasmas

Neoclassical tearing modes (NTMs) are magnetic islands which increase locally the radial transport and therefore degrade the plasma performance. They are self-sustained by the bootstrap current perturbed by the enhanced radial transport. The confinement degradation is proportional to the island width and to the position of the resonant surface. The q = 2 NTMs are much more detrimental to the confinement than the 3/2 modes due to their larger radii. NTMs are metastable in typical scenarios with βN ≥ 1 and in the region where the safety factor is increasing with radius. This is due to the fact that the local perturbed pressure gradient is sufficient to self-sustain an existing magnetic island. The main questions for burning plasmas are whether there is a trigger mechanism which will destabilize NTMs, and what is the best strategy to control/avoid the modes. The latter has to take into account the main aim which is to maximize the Q factor, but also the controllability of the scenario. Standardized and simplified equations are proposed to enable easier prediction of NTM control in burning plasmas from present experimental results. The present expected requirements for NTM control with localized electron cyclotron current drive (ECCD) in ITER are discussed in detail. Other aspects of the above questions are also discussed, in particular the role of partial stabilization of NTMs, the possibility to control NTMs at small size with little ECH power and the differences between controlling NTMs at the resonant surface or controlling the main trigger source, for the standard scenario namely the sawteeth. It is shown that there is no unique best strategy, but several tools are needed to most efficiently reduce the impact of NTMs on burning plasmas.