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Routine reaction to approaching disruptions in tokamaks is currently largely limited to machine protection by mitigating an ongoing disruption, which remains a basic requirement for ITER and DEMO [1]. Nevertheless, a mitigated disruption still generates stress to the device. Additionally, in future fusion devices, high-performance discharge time itself will be very valuable. Instead of reacting only on generic features, occurring shortly before the disruption, the ultimate goal is to actively avoid approaching disruptions at an early stage, sustain the discharges whenever possible and restrict mitigated disruptions to major failures. Knowledge of the most relevant root causes and the corresponding chain of events leading to disruption, the disruption path, is a prerequisite. For each disruption path, physics-based sensors and adequate actuators must be defined and their limitations considered. Early reaction facilitates the efficiency of the actuators and enhances the probability of a full recovery. Thus, sensors that detect potential disruptions in time are to be identified. Once the entrance into a disruption path is detected, we propose a hierarchy of actions consisting of (I) recovery of the discharge to full performance or at least continuation with a less disruption-prone backup scenario, (II) complete avoidance of disruption to sustain the discharge or at least delay it for a controlled termination and, (III), only as last resort, a disruption mitigation. Based on the understanding of disruption paths, a hierarchical and path-specific handling strategy must be developed. Such schemes, testable in present devices, could serve as guidelines for ITER and DEMO operation. For some disruption paths, experiments have been performed at ASDEX Upgrade and TCV. Disruptions were provoked in TCV by impurity injection into ELMy H-mode discharges and in ASDEX Upgrade by forcing a density limit in H-mode discharges. The new approach proposed in this paper is discussed for these cases. For the H-mode density limit sensors used so far react too late. Thus a plasma-state boundary is proposed, that can serve as an adequate early sensor for avoiding density limit disruptions in H-modes and for recovery to full performance. © 2017 Max-Planck-Institut for Plasmaphysics.

Integrated core–pedestal–SOL modelling is performed for ITER using the MMM95 model for energy transport, modified with additional magnetic shear stabilization to produce a pedestal. The additional stabilization term is calibrated against experiment. The operating space of ITER predicted by the model with neoclassical accumulation of intrinsic carbon impurity is determined. The relationship to an SOL-based density limit is investigated. A reasonable window for operation, with Q larger than 10 is found.

High-density discharges in JT-60U have been analysed with the aim of establishing a mechanism for the H-mode density limit (DL). The data support the view that the density behaviour in the H-mode is mainly affected by confinement, with the hard limit on the H-mode density being imposed by the H-L transition. Analysis of highly radiating NBI-heated plasmas with MARFE and detachment has enabled one to discriminate between detachment and the H-L transition as mechanisms for the H-mode DL. Although strong detachment is observed near the H-L transition in JT-60U, by itself it was found to impose no limit on the density rise.

A model for an ultimate density limit in tokamaks is presented. The model is based on considerations of radiative power balance and ideal ballooning stability limits at the plasma edge. Predictions for MAST are made and a comparison with experimental data on MAST is presented and discussed.

Recent results on the understanding and scaling of several of the operational boundaries found in ELMy H-mode plasmas are presented. Divertor geometry is found to be important in determining the H-mode transition power and temperature. Recent experiments from different tokamaks find apparently conflicting evidence for the role of the edge neutral density in controlling the H-mode transition and further intermachine comparisons are clearly warranted. Significant progress has been made both on the understanding and on the avoiding of the H-mode density limit. Models based on divertor detachment and on MARFE physics reproduce the empirical density limit scalings. The density limit in ELMy H-modes has been increased by changing the profile of the density in the divertor and in the main plasma. Highly radiating ELMy H-modes are currently being studied and display some of the favourable characteristics of the TEXTOR radiative improved confinement mode.

At ASDEX Upgrade the influence of triangularity on the H-mode performance has been studied intensively. It has been found that confinement increases with δ for a fixed line-averaged density. Though confinement decreases with increasing density for all analysed values of δ, H-factors (ITERH-98P) at the Greenwald density could be raised to 1 for the highest δ values achieved so far. The H-mode density limit could be increased by ≈20%. There is a scatter of about 30% on the confinement data, which is anti-correlated to the average density in the scrape-off layer or the neutral fluxes outside the plasma. For nearly all discharges analysed so far, the temperature profiles are self-similar. This indication of profile stiffness could be verified by changing the heat-flux profile by changing the beam-voltage of the neutral-beam injection (NBI) at high density. At low density, first results indicate a deviation from this stiff behaviour.

This paper gives an overview on the specific contributions from the tokamak TEXTOR/TEXTOR-94 to plasma-wall-interaction research, putting emphasis on the erosion and deposition properties (in particular graphite), hydrogen recycling mechanisms (molecules and atoms) and the characteristics of impurity radiation. The improved knowledge about these processes helps one to understand the mechanisms underlying these techniques, which have been pioneered on TEXTOR-94 for actively influencing plasma-wall interaction: impurity control by thin coatings, power exhaust on large areas with radiation from seeded impurities and the extension of the density limit by the control of radiative instabilities (MARFE). Also, the novel concept of the Dynamic Ergodic Divertor is presented as an innovative tool for controlling plasma-wall interaction.

The access to overdense plasmas (n c > n c.cut off ) for both the electron Bernstein (EB) wave heating and the EB wave diagnostic via the O-X-B mode conversion window was successfully demonstrated at W7-AS. In the EB wave heating experiments clear evidence for both mode conversion steps were detected and resonant absorption with a narrow profile was demonstrated. Also, the inverse process, the EB wave radiation of an overdense plasma was measured with an oblique viewing angle. The spectrum consists of a local cyclotron emission part and a non-local broadband high-frequency part. Edge temperature perturbations excited by carbon injection were used to demonstrate the relation between frequency and position of emission. Since for EB waves no density limit exists, the operation window of the electron cyclotron heating and diagnostic could be extended to densities above the cut off density.

In this paper we shall highlight some topics of specific concern for both concepts for magnetic confinement: 2D (tokamaks) and 3D (helical) systems. Section 2 will address the primary goals of fusion: density, confinement and temperature. We shall summarize the density limit, the global confinement time scaling with engineering parameters and some physical elements of improved confinement regimes with edge (H-mode) and core transport barriers (reversed shear cases); progress in relation to temperature originates from the evidence for α-particle heating from TFTR and the successes in drift optimization in helical systems with some confirmation from W7-AS. In the section 3 on the secondary goals of fusion, β will be discussed in terms of equilibrium (including the necessary improvements in helical systems with examples from W7-AS) and in terms of stability (including fast particle-driven Alfven waves with toroidally induced eigenmodes in cases with high shear and global eigenmodes in cases with low shear). Finally, we shall summarize the status of exhaust physics, with reference to the poloidal field divertor in tokamaks and the island divertor in helical systems. The intention of this report is to emphasize the complementarity of the two concepts of toroidal confinement.

H-mode operation presents challenges for ITER and for reactor application in the areas of performance, accessibility, sustainability and divertor compatibility. Some issues are more severe for ITER, while others are more problematic for an economic fusion reactor producing 3 MW of neutron flux to its walls. The problems that are more severe for ITER include the power requirements for the H-mode, and confinement uncertainties. Problems which are more severe for a reactor include capability for high operation, the role of the Greenwald density limit, and divertor compatibility. Within the present ITER concept, it would be very helpful to have a well defined upgrade path for the auxiliary heating systems, including options for strong profile control. The world fusion research programme should maintain a major focus on improving the reactor compatibility of the H-mode, with an eye to the special advantages of H-mode plasmas with strong shaping.

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.

Disruption mitigation is essential for the next generation of tokamaks. The prediction of plasma disruption is the key to disruption mitigation. A neural network combining eight input signals has been developed to predict the density limit disruptions on the J-TEXT tokamak. An optimized training method has been proposed which has improved the prediction performance. The network obtained has been tested on 64 disruption shots and 205 non-disruption shots. A successful alarm rate of 82.8% with a false alarm rate of 12.3% can be achieved at 4.8 ms prior to the current spike of the disruption. It indicates that more physical parameters than the current physical scaling should be considered for predicting the density limit. It was also found that the critical density for disruption can be predicted several tens of milliseconds in advance in most cases. Furthermore, if the network is used for real-time density feedback control, more than 95% of the density limit disruptions can be avoided by setting a proper threshold.

The effect of resonant magnetic perturbations (RMPs) on high density limit and low-q limit discharges is studied in J-TEXT tokamak. It is found that a moderate amplitude of applied RMPs increases the density limit and delays the disruption, and the disruption precursor is suppressed with a slight reduction in toroidal plasma rotation. A too large amplitude of RMPs, however, leads to a lower density limit and an earlier disruption. For the optimal RMPs amplitude, the disruption is delayed by about 150 ms. For low-q discharges, the applied static RMP can either lower the limit of the edge safety factor q a from 2.15 to nearly 2.0 or sustain stable plasma discharge with q a = 2.1, and the disruptive precursor is also suppressed.

Techniques for controlling the instabilities that lead to major disruptions are reviewed and their applicability to different types of disruptions (density limit beta -limit, low-q, and disruptions driven by error fields) in investigated. Particular emphasis is given to schemes for controlling the m=2, n=1 resistive tearing mode, whether arising from error fields or from natural plasma instability.

Marfes are toroidally symmetric bands of radiating plasma that form at the smaller major radius side of tokamaks. A model of marfes and density limit is studied. Though very simple, the model provides alternative explanation of the marfe onset beyond the density limit, explains the marfe poloidal location, states that a marfe can onset only when the criterion ( in /q)2 chi /sub // approximately /< (To/TL)2 chi perpendicular to is satisfied, where To is the central temperature and TL approximately 10eV is the temperature of intensive impurity radiation. Density fluctuations observed in marfe experiments are explained as the manifestation of sound waves which are intermediate between the isothermal and adiabatic.

In the regime of Improved Ohmic Confinement (IOC) in ASDEX the energy confinement time tau E increases linearly with increasing line-averaged density ne up to the density limit. The establishment of the IOC is accompanied by a substantial reduction of the external gas feed, concomitant with large decreases of all plasma edge fluxes. However, the data do not supply conclusive evidence that the IOC is primarily connected with the recycling conditions. More recent observations with very clean machine conditions seem to indicate that the impurity radiation plays a significant role.

This paper reviews the experimental results on lower hybrid current drive in tokamak plasmas. Pulse lengths of 3.5 seconds and currents above 400 kA have been generated at plasma densities such that the wave frequency is greater than about twice the lower hybrid frequency. Current drive ceases above a critical density, nc. However, nc increases with wave frequency. So that for f = 4.6 GHz current drive has been seen at ec ≈1014 cm-3 and a density limit has yet to be established. Evidence for a collisional scaling law for current-drive efficiency is summarized. Detailed measurements of bremsstrahlung X-rays show a distribution which is qualitatively similar to that predicted by quasilinear theory. Microwave emission at frequencies less than the plasma frequency may shed light on the current-drive mechanisn. Applications of current drive including plasma and current start-up and transformer recharging are discussed.