A. Isayama

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.

Achievement of high fusion triple product, steady-state sustainment and real-time NTM stabilization in high-βp ELMy H-mode discharges in JT-60U

This paper reports results on the progress in steady-state high-βp ELMy H-mode discharges in JT-60U. A fusion triple product, nD(0)τETi(0), of 3.1 × 1020 m−3 s keV under full non-inductive current drive has been achieved at Ip = 1.8 MA, which extends the record value of the fusion triple product under full non-inductive current drive by 50%. A high-beta plasma with βN ~ 2.7 has been sustained for 7.4 s (~60τE), with the duration determined only by the facility limits, such as the capability of the poloidal field coils and the upper limit on the duration of injection of neutral beams. Destabilization of neoclassical tearing modes (NTMs) has been avoided with good reproducibility by tailoring the current and pressure profiles. On the other hand, a real-time NTM stabilization system has been developed where detection of the centre of the magnetic island and optimization of the injection angle of the electron cyclotron wave are done in real time. By applying this system, a 3/2 NTM has been completely stabilized in a high-beta region (βp ~ 1.2, βN ~ 1.5), and the beta value and confinement enhancement factor have been improved by the stabilization.

Complete stabilization of a tearing mode in steady state high-βp H-mode discharges by the first harmonic electron cyclotron heating/current drive on JT-60U

A tearing mode with m = 3 and n = 2, destabilized in the steady state high-βp H-mode discharges with edge localized mode (ELM), was completely stabilized by local heating and current drive using the 110 GHz first harmonic O-mode electron cyclotron (EC) wave. Here, m and n are poloidal and toroidal mode numbers, respectively. The optimum EC wave injection angle was determined by identifying the mode location from an electron temperature perturbation profile and a safety factor profile. The optimum injection angle was also determined by scanning a steerable mirror during a discharge. In a typical discharge where the tearing mode is completely stabilized, the ratio of the electron cyclotron heating power to the total heating power is 0.17, and the ratio of the EC driven current to the total plasma current is 0.02. Stored energy and neutron emission rate were higher for the case with EC wave injection than that without EC wave injection, which suggests that the reduction of the stored energy and the neutron emission rate was recovered by the tearing mode stabilization.

Neoclassical tearing mode control using electron cyclotron current drive and magnetic island evolution in JT-60U

The results of stabilizing neoclassical tearing modes (NTMs) with electron cyclotron current drive (ECCD) in JT-60U are described with emphasis on the effectiveness of the stabilization. The range of the minimum EC wave power needed for complete stabilization of an m/n = 2/1 NTM was experimentally identified for two regimes using unmodulated ECCD to clarify the NTM behaviours with different plasma parameters: 0.2 < jEC/jBS < 0.4 for Wsat/dEC ~ 3 and Wsat/Wmarg ~ 2, and 0.35 < jEC/jBS < 0.46 for Wsat/dEC ~ 1.5 and Wsat/Wmarg ~ 2. Here, m and n are the poloidal and toroidal mode numbers; jEC and jBS the EC-driven current density and bootstrap current density at the mode rational surface; Wsat, Wmarg and dEC the full island width at saturation, marginal island width and full-width at half maximum of the ECCD deposition profile, respectively. Stabilization of a 2/1 NTM using modulated ECCD synchronized with a mode rotation of about 5 kHz was performed, in which it was found that the stabilization effect degrades when the phase of the modulation deviates from that of the ECCD at the island O-point. The decay time of the magnetic perturbation amplitude due to the ECCD increases by 50% with a phase shift of ±50° from the O-point ECCD, thus revealing the importance of the phasing of modulated ECCD. For near X-point ECCD, the NTM amplitude increases, revealing a destabilization effect. It was also found that modulated ECCD at the island O-point has a stronger stabilization effect than unmodulated ECCD by a factor of more than 2.

Energy loss for grassy ELMs and effects of plasma rotation on the ELM characteristics in JT-60U

The energy loss for grassy edge localized modes (ELMs) has been studied to investigate the applicability of the grassy ELM regime to ITER. The grassy ELM regime is characterized by high frequency periodic collapses of 800–1500 Hz, which is ~15 times faster than that for type I ELMs. The divertor peak heat flux due to grassy ELMs is less than 10% of that for type I ELMs. This smaller heat flux is caused by a narrower radial extent of the collapse of the temperature pedestal. The different radial extent between type I ELMs and grassy ELMs agrees qualitatively with the different radial distribution of the eigenfunctions as determined from ideal MHD stability analysis. The dominant ELM energy loss for grassy ELMs appears to be caused by temperature reduction, and its ratio to the pedestal stored energy was 0.4–1%. This ratio is lower by a factor of about 10 than that for type I ELMs, which typically have between 2–10% fractional loss of the pedestal energy. A systematic study of the effects of counter (CTR) plasma rotation on the ELM characteristics has been performed using a combination of tangential and perpendicular neutral beam injections (NBIs) in JT-60U. In the high plasma triangularity (δ) regime, ELM characteristics (e.g. amplitude, frequency and type) can be changed from type I ELMs to high frequency grassy ELMs as the CTR plasma rotation is increased. On the other hand, in the low δ regime, complete ELM suppression (QH-mode) can be sustained for long periods up to 3.4 s (~18τE or energy confinement times), when the plasma position in terms of the clearance between the first wall and the plasma separatrix is optimized during the application of CTR-NBIs. In JT-60U, a transient QH phase was also observed during the CO-NBI phase with almost no net toroidal rotation at the plasma edge.

High performance experiments in JT-60U reversed shear discharges

The operation of JT-60U reversed shear discharges has been extended to a high plasma current, low q regime keeping a large radius of the internal transport barrier (ITB), and a record value of equivalent fusion multiplication factor in JT-60U, QDTeq = 1.25, has been achieved at 2.6 MA. Operational schemes to reach the low q regime with good reproducibility have been developed. The reduction of Zeff was obtained in the newly installed W shaped pumped divertor. The β limit in the low qmin regime, which limited the performance of L mode edge discharges, has been improved in H mode edge discharges with a broader pressure profile, which was obtained by power flow control with ITB degradation. Sustainment of the ITB and improved confinement for 5.5 s has been demonstrated in an ELMy H mode reversed shear discharge.

Extended steady-state and high-beta regimes of net-current free heliotron plasmas in the Large Helical Device

The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as 5 × 10 20 m -3 , which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high β and long pulse.

Disappearance of giant ELMs and appearance of minute grassy ELMs in JT-60U high-triangularity discharges

In JT-60U H-mode plasmas, giant (type I) ELMs disappear and minute grassy ELMs appear when triangularity δ, edge safety factor q95 and βp are high enough. Complete suppression of giant ELMs was observed at δ0.45, q956 and βp1.6. At higher δ (0.54), giant ELMs can disappear at a lower q95 (~4.0). In the grassy ELMy H-mode, edge temperature and pressure can be higher than those in giant ELMy H-mode and a favourable confinement can be sustained without an increase of the impurity concentration. An edge stability analysis suggests that the edge plasma is accessing the second stability regime of the high n ballooning mode in the grassy ELMy discharges.

Experimental studies of ITER demonstration discharges

Key parts of the ITER scenarios are determined by the capability of the proposed poloidal field (PF) coil set. They include the plasma breakdown at low loop voltage, the current rise phase, the performance during the flat top (FT) phase and a ramp down of the plasma. The ITER discharge evolution has been verified in dedicated experiments. New data are obtained from C-Mod, ASDEX Upgrade, DIII-D, JT-60U and JET. Results show that breakdown for Eaxis < 0.23–0.33 V m−1 is possible unassisted (ohmic) for large devices like JET and attainable in devices with a capability of using ECRH assist. For the current ramp up, good control of the plasma inductance is obtained using a full bore plasma shape with early X-point formation. This allows optimization of the flux usage from the PF set. Additional heating keeps li(3) < 0.85 during the ramp up to q95 = 3. A rise phase with an H-mode transition is capable of achieving li(3) < 0.7 at the start of the FT. Operation of the H-mode reference scenario at q95 ~ 3 and the hybrid scenario at q95 = 4–4.5 during the FT phase is documented, providing data for the li (3) evolution after the H-mode transition and the li (3) evolution after a back-transition to L-mode. During the ITER ramp down it is important to remain diverted and to reduce the elongation. The inductance could be kept ≤1.2 during the first half of the current decay, using a slow Ip ramp down, but still consuming flux from the transformer. Alternatively, the discharges can be kept in H-mode during most of the ramp down, requiring significant amounts of additional heating.

Comparison of transient electron heat transport in LHD helical and JT-60U tokamak plasmas

Transient transport experiments are performed in plasmas with and without internal transport barriers (ITB) on LHD and JT-60U. The dependence of χe on the electron temperature, Te, and on the electron temperature gradient, ∇Te, is analysed with an empirical non-linear heat transport model. In plasmas without an ITB, two different types of non-linearity of the electron heat transport are observed from cold/heat pulse propagation: the χe depends on Te and ∇Te in JT-60U, while the ∇Te dependence is weak in LHD. Inside the ITB region, there is none or weak ∇Te dependence both in LHD and JT-60U. Growth of the cold pulse driven by the negative Te dependence of χe is observed inside the ITB region (LHD) and near the boundary of the ITB region (JT-60U).