R. Granetz

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.

First results from Alcator-C-MOD

Early operation of the Alcator‐C‐MOD tokamak [I.H. Hutchinson, Proceedings of IEEE 13th Symposium on Fusion Engineering, Knoxville, TN, edited by M. Lubell, M. Nestor, and S. Vaughan (Institute of Electrical and Electronic Engineers, New York, 1990), Vol. 1, p. 13] is surveyed. Reliable operation, with plasma current up to 1 MA, has been obtained, despite the massive conducting superstructure and the associated error fields. However, vertical disruptions are not slowed by the long vessel time constant. With pellet fueling, peak densities up to 9×1020 m−3 have been attained and ‘‘snakes’’ are often seen. Initial characterization of divertor and scrape‐off layer is presented and indicates approximately Bohm diffusion. The edge plasma shows a wealth of marfe‐like phenomena, including a transition to detachment from the divertor plates with accompanying radiative divertor regions. Energy confinement generally appears to exceed the expectations of neo‐Alcator scaling. A transition to Ohmic H mode has been observed. Ion cyclotron heating experiments have demonstrated good power coupling, in agreement with theory.

Disruptions in JET

In JET, both high density and low-q operation are limited by disruptions. The density limit disruptions are caused initially by impurity radiation. This causes a contraction of the plasma temperature profile and leads to an MHD unstable configuration. There is evidence of magnetic island formation resulting in minor disruptions. After several minor disruptions, a major disruption with a rapid energy quench occurs. This event takes place in two stages. In the first stage there is a loss of energy from the central region. In the second stage there is a more rapid drop to a very low temperature, apparently due to a dramatic increase in impurity radiation. The final current decay takes place in the resulting cold plasma. During the growth of the MHD instability the initially rotating mode is brought to rest. This mode locking is believed to be due to an electromagnetic interaction with the vacuum vessel and external magnetic field asymmetries. The low-q disruptions are remarkable for the precision with which they occur at qψ = 2. These disruptions do not have extended precursors or minor disruptions. The instability grows and locks rapidly. The energy quench and current decay are generally similar to those of the density limit.

X-ray tomography on JET

Measurements of soft X-ray emission from the JET plasma have been analysed with tomographic reconstruction methods. Because there are two detector arrays, two-dimensional images of X-ray emissivity are obtained without having to resort to rotation models. Several algorithms are employed in order to get as much detail as possible in the images while keeping any guiding assumptions to a minimum. The data analysed so far have been used principally to study MHD instabilities, and illustrative examples of the sawtooth crash and of disruptions are described.

Experimental investigation of transport phenomena in the scrape-off layer and divertor

Abstract Transport physics in the divertor and scrape-off layer of Alcator C-Mod is investigated for a wide range of plasma conditions. Parallel (∥) transport topics include: low recycling, high-recycling, and detached regimes, thermoelectric currents, asymmetric heat fluxes driven by thermoelectric currents, and reversed divertor flows. Perpendicular (⊥) transport topics include: expected and measured scalings of ⊥ gradients with local conditions, estimated χ⊥ profiles and scalings, divertor neutral retention effects, and L-mode/H-mode effects. Key results are: (i) classical ∥ transport is obeyed with ion-neutral momentum coupling effects, (ii) ⊥ heat transport is proportional to local gradients, (iii) χ⊥ αTe−0.6 n−0.6 L−0.7 in L-mode, insensitive to toroidal field, (iv) χ⊥ is dependent on divertor neutral retention, (v) H-mode transport barrier effects partially extend inside the SOL, (vi) inside/outside divertor asymmetries may be caused by a thermoelectric instability, and (vii) reversed ∥ flows depend on divertor asymmetries and their implicit ionization source imbalances.

Edge radial electric field structure and its connections to H-mode confinement in Alcator C-Mod Plasmas

High-resolution charge-exchange recombination spectroscopic measurements of B5+ ions have enabled the first spatially resolved calculations of the radial electric field (Er) in the Alcator C-Mod pedestal region [E. S. Marmar, Fusion Sci. Technol. 51, 261 (2006)]. These observations offer new challenges for theory and simulation and provide for important comparisons with other devices. Qualitatively, the field structure observed on C-Mod is similar to that on other tokamaks. However, the narrow high-confinement mode (H-mode) Er well widths (5 mm) observed on C-Mod suggest a scaling with machine size, while the observed depths (up to 300 kV/m) are unprecedented. Due to the strong ion-electron thermal coupling in the C-Mod pedestal, it is possible to infer information about the main ion population in this region. The results indicate that in H-mode the main ion pressure gradient is the dominant contributor to the Er well and that the main ions have significant edge flow. C-Mod H-mode data show a clear correl...

Role of trapped electron mode turbulence in internal transport barrier control in the Alcator C-Mod Tokamak

Nonlinear gyrokinetic simulations of trapped electron mode (TEM) turbulence, within an internal particle transport barrier, are performed and compared with experimental data. The results provide a mechanism for transport barrier control with on-axis radio frequency heating, as demonstrated in Alcator C-Mod experiments [S. J. Wukitch et al., Phys. Plasmas 9, 2149 (2002)]. Off-axis heating produces an internal particle and energy transport barrier after the transition to enhanced Dα high confinement mode. The barrier foot reaches the half-radius, with a peak density 2.5 times the edge density. While the density profile peaks, the temperature profile remains relatively unaffected. The peaking and concomitant impurity accumulation are controlled by applying modest central heating power late in the discharge. Gyrokinetic turbulence simulations of the barrier formation phase, using the GS2 code [W. Dorland et al., Phys. Rev. Lett. 85, 5579 (2000)] show that toroidal ion temperature gradient driven modes are sup...

Characterization of enhanced Dα high-confinement modes in Alcator C-Mod

Regimes of high-confinement mode have been studied in the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)]. Plasmas with no edge localized modes (ELM-free) have been compared in detail to a new regime, enhanced Dα (EDA). EDA discharges have only slightly lower energy confinement than comparable ELM-free ones, but show markedly reduced impurity confinement. Thus EDA discharges do not accumulate impurities and typically have a lower fraction of radiated power. The edge gradients in EDA seem to be relaxed by a continuous process rather than an intermittent one as is the case for standard ELMy discharges and thus do not present the first wall with large periodic heat loads. This process is probably related to fluctuations seen in the plasma edge. EDA plasmas are more likely at low plasma current (q>3.7), for moderate plasma shaping, (triangularity ∼0.35–0.55), and for high neutral pressures. As observed in soft x-ray emission, the pedestal width is found to scale with the same parameters that determine the EDA/ELM-free boundary.

Nonaxisymmetric field effects on Alcator C-Mod

A set of external coils (A-coils) capable of producing nonaxisymmetric, predominantly n=1, fields with different toroidal phase and a range of poloidal mode m spectra has been used to determine the threshold amplitude for mode locking over a range of plasma parameters in Alcator C-Mod [I. H. Hutchinson, R. Boivin, F. Bombarda, P. Bonoli, S. Fairfax, C. Fiore, J. Goetz, S. Golovato, R. Granetz, M. Greenwald et al., Phys. Plasmas 1, 1511 (1994)]. The threshold perturbations and parametric scalings, expressed in terms of (B21∕BT), are similar to those observed on larger, lower field devices. The threshold is roughly linear in density, with typical magnitudes of order 10−4. This result implies that locked modes should not be significantly more problematic for the International Thermonuclear Experimental Reactor [I. P. B. Editors, Nucl. Fusion 39, 2286 (1999)] than for existing devices. Coordinated nondimensional identity experiments on the Joint European Torus [Fusion Technol. 11, 13 (1987)], DIII-D [Fusion T...

Transport-driven scrape-off layer flows and the x-point dependence of the L-H power threshold in Alcator C-Mod

Factor of ∼2 higher power thresholds for low- to high-confinement mode transitions (L-H) with unfavorable x-point topologies in Alcator C-Mod [Phys. Plasmas 1, 1511 (1994)] are linked to flow boundary conditions imposed by the scrape-off layer (SOL). Ballooning-like transport drives flow along magnetic field lines from low- to high-field regions with toroidal direction dependent on upper/lower x-point balance; the toroidal rotation of the confined plasma responds, exhibiting a strong counter-current rotation when B×∇B points away from the x point. Increased auxiliary heating power (rf, no momentum input) leads to an L-H transition at approximately twice the edge electron pressure gradient when B×∇B points away. As gradients rise prior to the transition, toroidal rotation ramps toward the co-current direction; the H mode is seen when the counter-current rotation imposed by the SOL flow becomes compensated. Remarkably, L-H thresholds in lower-limited discharges are identical to lower x-point discharges; SOL...