G. Pautasso

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.

Plasma?surface interaction, scrape-off layer and divertor physics: implications for ITER

Recent research in scrape-off layer (SOL) and divertor physics is reviewed; new and existing data from a variety of experiments have been used to make cross-experiment comparisons with implications for further research and ITER. Studies of the region near the separatrix have addressed the relationship of profiles to turbulence as well as the scaling of the parallel power flow. Enhanced low-field side radial transport is implicated as driving parallel flows to the inboard side. The medium-n nature of edge localized modes (ELMs) has been elucidated and new measurements have determined that they carry ~10?20% of the ELM energy to the far SOL with implications for ITER limiters and the upper divertor. The predicted divertor power loads for ITER disruptions are reduced while those to main chamber plasma facing components (PFCs) increase. Disruption mitigation through massive gas puffing is successful at reducing PFC heat loads. New estimates of ITER tritium retention have shown tile sides to play a significant role; tritium cleanup may be necessary every few days to weeks. ITERs use of mixed materials gives rise to a reduction of surface melting temperatures and chemical sputtering. Advances in modelling of the ITER divertor and flows have enhanced the capability to match experimental data and predict ITER performance.

Energy flux to the ASDEX-Upgrade diverter plates determined by thermography and calorimetry

A new thermography system with high time resolution was put into operation at ASDEX-Upgrade and is routinely used to determine the energy flux onto the lower diverter plates. The measurements allow the power deposition to be characterized during dynamic events such as ELMs and disruptions, as well as the asymmetry of the inboard/outboard power load. A power balance is set up even during single discharges and the losses are found to be fairly equal to the power input.

Transient heat loads in current fusion experiments, extrapolation to ITER and consequences for its operation

New experimental results on transient loads during ELMs and disruptions in present divertor tokamaks are described and used to carry out a extrapolation to ITER reference conditions and to draw consequences for its operation. In particular, the achievement of low energy/convective type I edge localized modes (ELMs) in ITER-like plasma conditions seems the only way to obtain transient loads which may be compatible with an acceptable erosion lifetime of plasma facing components (PFCs) in ITER. Power loads during disruptions, on the contrary, seem to lead in most cases to an acceptable divertor lifetime because of the relatively small plasma thermal energy remaining at the thermal quench and the large broadening of the power flux footprint during this phase. These conclusions are reinforced by calculations of the expected erosion lifetime, under these load conditions, which take into account a realistic temporal dependence of the power fluxes on PFCs during ELMs and disruptions.

Disruption studies in ASDEX Upgrade in view of ITER

Experiments on ASDEX Upgrade and other tokamaks have shown that the magnitude of mechanical forces and thermal loads during disruptions can be significantly reduced by raising the plasma density with massive injection of noble gases. This method should be applicable to ITER too. Nevertheless, the suppression of the runaway electron (RE) avalanche requires a much larger (two order of magnitude) density rise. This paper reports on recent experiments aimed at increasing the plasma density towards the critical value, needed for the collisional suppression of REs. An effective electron density equal to 24% of the critical density has been reached after injection of 3.3?bar?l of neon. However, the resultant large plasma density is very poloidally and toroidally asymmetric; this implies that several valves distributed around the plasma periphery become necessary at this level of massive gas injection to ensure a homogeneous density distribution.

Plasma shut-down with fast impurity puff on ASDEX Upgrade

The massive injection of impurity gas into a plasma has been proved to reduce forces and localized thermal loads caused by disruptions in tokamaks. This mitigation system is routinely used on ASDEX Upgrade to shut down plasmas with a locked mode. The plasma response to impurity injection and the mechanism of reduction of the mechanical forces is discussed in the paper.

Vertical displacement events and halo currents

This review examines results from all non-circular tokamaks with a distinct emphasis on investigations in ASDEX-Upgrade. There a major fraction of the experimental time has been dedicated studying vertical displacement events of single null plasmas over a large range of q-values in an attempt to obtain the scaling of both the displacement dynamics and the splitting of forces between those associated with poloidal and toroidal plasma currents as a function of q and Bt. These studies on different tokamaks are accompanied by simulations with-among other codes-the tokamak simulation code TSC, in a version where halo currents flowing in the plasma scrape-off layer (SOL) evolve self-consistently. The technical consequences of VDEs for the machine design, measures taken and first predictions are discussed. Safety setups that have been developed and possible avoidance strategies are briefly described.

Disruption control on FTU and ASDEX upgrade with ECRH

The use of ECRH has been investigated as a promising technique to avoid or postpone disruptions in dedicated experiments in FTU and ASDEX Upgrade. Disruptions have been produced by injecting Mo through laser blow-off (FTU) or by puffing deuterium gas above the Greenwald limit (FTU and ASDEX Upgrade). The toroidal magnetic field is kept fixed and the ECRH launching mirrors have been steered before every discharge in order to change the deposition radius. The loop voltage signal is used as disruption precursor to trigger the ECRH power before the plasma current quench. In the FTU experiments (Ip = 0.35–0.5 MA, Bt = 5. 3T ,PECRH = 0.4–1.2 MW) it is found that the application of ECRH modifies the current quench starting time depending on the power deposition location. A scan in deposition location has shown that the direct heating of one of the magnetic islands produced by magnetohydrodynamic (MHD) resistive instabilities (either m/n = 3/2, 2/1 or 3/1) prevents its further growth and also produces the stabilization of the other coupled modes and the delay of the current quench or its full avoidance. Disruption avoidance and complete discharge recovery are obtained when the ECRH power is applied on rational surfaces. The modes involved in the disruption are found to be tearing modes stabilized by a strong local ECRH heating. The Rutherford equation has been used to reproduce the evolution of the MHD modes. In the ASDEX Upgrade experiments L-mode plasmas (Ip = 0.6 MA, Bt = 2. 5T ,PECRH = 0. 6M W∼ POHM) the injection of ECRH close to q = 2 significantly delays the 2/1 onset and prolongs the duration of the discharge: during this phase the density continues to increase. No delay in the onset of the 2/1 mode is observed when the injected power is reduced to 0.35 MW.

An adaptive real-time disruption predictor for ASDEX Upgrade

In this paper, a neural predictor has been built using plasma discharges selected from two years of ASDEX Upgrade experiments, from July 2002 to July 2004. In order to test the real-time prediction capability of the system, its performance has been evaluated using discharges coming from different experimental campaigns, from June 2005 to July 2007. All disruptions that occurred in the chosen experimental campaigns were included with the exception of those occurring in the ramp-up phase, in the ramp-down phase (if the disruption does not happen in the first 100 ms), those caused by massive gas injection and disruptions following vertical displacement events. The large majority of selected disruptions are of the cooling edge type and typically preceded by the growth of tearing modes, degradation of the thermal confinement and enhanced plasma radiation. A very small percentage of them happen at large beta after a short precursor phase. For each discharge, seven plasma diagnostic signals have been selected from numerous signals available in real-time. During the training procedure, a self-organizing map has been used to reduce the database size in order to improve the training of the neural network. Moreover, an optimization procedure has been performed to discriminate between safe and pre-disruptive phases. The prediction success rate has been further improved, performing an adaptive training of the network whenever a missed alarm is triggered by the predictor.

MHD stability and disruption physics in ASDEX Upgrade

The MHD activity giving rise to the beta - and the density limit in ASDEX Upgrade is analyzed. A detailed description of the MHD phenomena occuring prior to and during disruptions is given. The MHD characteristics of the different ELM types occurring in ASDEX Upgrade are described.