M. A. Henderson

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.

Effects of localized electron heating and current drive on the sawtooth period

Localized electron heating and current drive, like those produced by electron cyclotron heating (ECH) systems, are powerful tools for controlling the sawtooth period. They allow the direct modification of the plasma parameters which determine the sawtooth stability. In this paper we report a set of new experimental results obtained in the Tokamak a Configuration Variable (TCV) and a set of related simulations obtained applying a sawtooth period model in a transport code. The TCV device, equipped with a very flexible and powerful ECH system, is specifically suited for this kind of study. In previous works, the experimental behaviour observed in TCV and JET was found consistent with a sawtooth period model first proposed to predict the sawtooth period in burning plasmas. In this paper, new experimental results have motivated a set of simulations which allow the identification of the effects of localized heating and current drive separately. In particular, two heating locations exist at opposite sides of the q = 1 surface which allow most efficiently sawtooth stabilization and destabilization. Moreover, the modelling shows that counter- and co-current drive alone, without the presence of heating, have opposite effects on the sawtooth period at symmetrical locations as compared with the position of the q = 1 surface. The main features of the experimental behaviour can be explained as due to the modification of the local plasma parameters involved in the linear resistive stability threshold of the internal kink, in particular the dynamics of the magnetic shear at the q = 1 surface. However it is shown that the most effective locations to modify the sawtooth period are not exactly at q = 1.

Critical interface issues associated with the ITER EC system

The ITER ECH system is an in-kind procurement consisting of four different types of gyrotrons (from EU, IN, JA and RF), transmission lines (from US) and two types of launchers (from EU and JA). Each subsystem must interface not only with the other but also with the auxiliary systems control and data acquisition computer and with the plasma (in the case of the launchers). The definition and management of interfaces is therefore essential for the system to guarantee performance, availability and reliability. The proper description of each interface boundary is essential for assembly and operation of the entire system as a single unit. In addition, progress has been made in the development of high power, long pulse systems and associated components that have not been integrated into the ITER EC design since the present ITER EC system was essentially specified prior to 2000. The ultimate physics performance and operational reliability in some situations is limited by this old design, which has not taken advantage of the knowledge and experience gained in operating the multi-megawatt ECH systems on present tokamaks and stellarators.The objective of this paper is to review the present ITER ECH system, which includes the power supplies, gyrotrons, transmission lines and launchers. Modifications are proposed which are performance driven and are engineered for reliability and maintainability, whilst reducing complexity and cost. Potential operating scenarios are discussed which require an intelligent and automatic decision making process, for example, directing the EC power to either of the two EC launchers, based on the immediate physics requirements. The interfaces between the subsystems are described and when possible improvements to each interface are proposed.

Full absorption of third harmonic ECH in TCV tokamak plasmas in the presence of second harmonic ECCD

An experimental study of the extraordinary mode (X mode) absorption at the third electron cyclotron harmonic frequency has been performed on the TCV tokamak in plasmas preheated by X mode at the second harmonic. Full single pass absorption of injected X3 power was measured with X2 preheating in co-current drive (CO-ECCD). The measured absorption exceeds that predicted by the linear ray tracing code TORAY-GA by more than a factor of 2 for the CO-ECCD case. Experimental evidence indicates that a large fraction of the X3 power is absorbed by electrons in an energetic tail created by the X2 ECCD preheating.

Behaviour of central plasma relaxation oscillations during localized electron cyclotron heating on the TCV tokamak

During initial studies of ECRH in the TCV tokamak, non-standard central MHD activities, such as humpbacks and saturated and inverted sawteeth, have been observed while changing the heating location, the ECRH power, the plasma shape and the safety factor. For edge safety factors q(alpha) > 4.5, safety factors on-axis q(0) < 1 and small plasmas, complete sawtooth stabilization was achieved with the present 1 MW gyrotron power, and it is likely that sawtooth stabilization can be achieved for all conditions at. higher ECRH power. The conditions under which the various relaxation activities are produced or suppressed are reported, and the origins for such tron-standard behaviour are discussed.

Recent results from the electron cyclotron heated plasmas in Tokamak à Configuration Variable (TCV)

In noninductively driven discharges, 0.9 MW second harmonic (X2) off-axis co-electron cyclotron current drive deposition is combined with 0.45 MW X2 central heating to create an electron internal transport barrier (eITB) in steady plasma conditions resulting in a 1.6-fold increase of the confinement time (τEe) over ITER-98L-mode scaling. The eITB is associated with a reversed shear current profile enhanced by a large bootstrap current fraction (up to 80%) and is sustained for up to 10 current redistribution times. A linear dependence of the confinement improvement on the product of the global shear reversal factor (q0/qmin) and the reversed shear volume (ρq-min2) is shown. In other discharges heated with X2 the sawteeth are destabilized (respectively stabilized) when heating just inside (respectively outside) the q=1 surface. Control of the sawteeth may allow the avoidance of neoclassical tearing modes that can be seeded by the sawtooth instability. Results on H-mode and highly elongated plasmas using the...

An overview of results from the TCV tokamak

The Tokamak Configuration Variable (TCV) tokamak (R = 0.88 m, a < 0.25 m, B < 1.54 T) programme is based on flexible plasma shaping and heating for studies of confinement, transport, control and power exhaust. Recent advances in fully sustained off-axis electron cyclotron current drive (ECCD) scenarios have allowed the creation of plasmas with high bootstrap fraction, steady-state reversed central shear and an electron internal transport barrier. High elongation plasmas, kappa = 2.5, are produced at low normalized current using far off-axis electron cyclotron heating and ECCD to broaden the current profile. Third harmonic heating is used to heat the plasma centre where the second harmonic is in cut-off. Both second and third harmonic heating are used to heat H-mode plasmas, at the edge and centre, respectively. The ELM frequency is decreased by the additional power. In separate experiments, the ELM frequency can be affected by locking to an external perturbation current in the internal coils of TCV. Spatially resolved current profiles are measured at the inner and outer divertor targets by Langmuir probe arrays during ELMs. The strong, reasonably balanced currents are thought to be thermoelectric in origin.

Effect of triangular and elongated plasma shape on the sawtooth stability

In the TCV tokamak the sawtooth period and the sawtooth amplitude are observed to depend strongly on the shape of the poloidal plasma cross section. Systematic scans of plasma elongation and triangularity show small sawteeth with short periods at high elongation or low and negative triangularity, and large sawteeth with long periods at low elongation or high triangularity. Additional central electron cyclotron heating power further amplifies the shape dependence of the sawtooth properties. The sawtooth period can increase or decrease with additional heating power depending on the plasma shape. This shape dependence is determined by the role of ideal or resistive MHD in triggering the sawtooth crash. For plasma shapes where additional heating and, consequently, a higher pressure shortens the sawtooth period, the low experimental limit of the pressure gradient within the q = 1 surface is consistent with ideal MHD predictions. The observed decrease of this limit with elongation is also in qualitative agreement with ideal MHD theory.

Modulated ECH power absorption measurements using a diamagnetic loop in the TCV tokamak

The additional power absorbed by the plasma can be determined from the time derivative of the total plasma energy, which can be estimated from the diamagnetic flux of the plasma using a diamagnetic loop. The main difficulty in using diamagnetic measurements to estimate the kinetic energy is the compensation of the flux measurement sensitivity to poloidal magnetic fields, which is not always easy to adjust. A method based on the temporal variations of the diamagnetic flux of the plasma during modulated electron cyclotron heating (MECH) has been developed. Using MECH has the advantage that these poloidal fields are not significantly modulated and a good compensation of these fields is not necessary. However, a good compensation of the vessel poloidal image current is crucial to ensure a sufficiently large bandwidth. The application of this diagnostic to studies of the extraordinary mode (X-mode) absorption at the third electron cyclotron harmonic frequency (X3) has been performed on the TCV tokamak in plasmas pre-heated by the X-mode at the second harmonic (X2). A MECH frequency scan has allowed the determination of an optimum modulation frequency, situated at about 200-250 Hz. Based on this diagnostic, full single-pass absorption of the injected X3 power was measured with the X2 pre-heating in co-current drive. This high absorption is more than a factor of 2 higher than that predicted by the linear ray tracing code TORAY. Experimental evidence indicates that a large fraction of the X3 power is absorbed by electrons in an energetic tail created by the X2 pre-heating.

Electron heat transport in shaped TCV L-mode plasmas

Electron heat transport experiments are performed in L-mode discharges at various plasma triangularities, using radially localized electron cyclotron heating to vary independently both the electron temperature T-e and the normalized electron temperature gradient R/L-Te over a large range. Local gyrofluid (GLF23) and global collisionless gyro-kinetic (LORB5) linear simulations show that, in the present experiments, trapped electron mode (TEM) is the most unstable mode. Experimentally, the electron heat diffusivity chi(e) is shown to decrease with increasing collisionality, and no dependence of chi(e) on R/L-Te is observed at high R/L-Te values. These two observations are consistent with the predictions of TEM simulations, which supports the fact that TEM plays a crucial role in electron heat transport. In addition, over the broad range of positive and negative triangularities investigated, the electron heat diffusivity is observed to decrease with decreasing plasma triangularity, leading to a strong increase of plasma confinement at negative triangularity.