A. G. Kellman

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.

An optimization of beta in the DIII-D tokamak

Accurate equilibrium reconstruction and detailed stability analysis of a strongly shaped, double‐null, βT=11% discharge shows that the plasma core is in the second stable regime to ideal ballooning modes. The equilibrium reconstruction using all the available data (coil currents, poloidal magnetic loops, motional Stark effect data, the kinetic pressure profile, the magnetic axis location, and the location of the two q=1 surfaces) shows a region of negative magnetic shear near the magnetic axis, an outer positive shear region, and a low shear region connecting the two. The inner negative shear region allows a large positive shear region near the boundary, even at low q (q95=2.6), permitting a large outer region pressure gradient to be first regime stable. The inner region is in the second stable regime, consistent with the observed axial beta [βT(0)=44%]. In the low shear region p’ vanishes, consistent with Mercier stability. This is one way to extend the ballooning limit in shaped plasmas while maintainin...

Higher beta at higher elongation in the DIII-D tokamak

A theoretical and experimental evaluation of axisymmetric stability and axisymmetric control has led to a modification of the vertical position control in the DIII‐D tokamak, which now allows operation to within a few percent of the ideal magnetohydrodynamic (MHD) n=0 limit. It is found that the onset the departure from rigid shift behavior in D‐shaped plasmas limits plasma elongation to 2.5 in DIII‐D. The possibility of avoiding the vertical instability in future tokamaks with highly elongated plasmas is discussed. Recent experiments have focused on utilizing this capability for axisymmetric control to construct plasma shapes optimized to increase the achievable beta. Operation near the axisymmetric stability limit allows an increase in the achieved normalized current Ip/aBT, where Ip is the total plasma current, a is the minor radius, and BT is the toroidal field. Based on stability calculations, an equilibrium was developed to achieve marginal stability simultaneously to axisymmetric, kink, and balloon...

Stationary high-performance discharges in the DIII-D tokamak

Discharges which can satisfy the high gain goals of burning plasma experiments have been demonstrated in the DIII-D tokamak under stationary conditions at relatively low plasma current (q95>4). A figure of merit for fusion gain (?NH89/q952) has been maintained at values corresponding to Q = 10 operation in a burning plasma for >6?s or 36?E and 2?R. The key element is the relaxation of the current profile to a stationary state with qmin>1. In the absence of sawteeth and fishbones, stable operation has been achieved up to the estimated no-wall ? limit. Feedback control of the energy content and particle inventory allow reproducible, stationary operation. The particle inventory is controlled by gas fuelling and active pumping; the wall plays only a small role in the particle balance. The reduced current lessens significantly the potential for structural damage in the event of a major disruption. In addition, the pulse length capability is greatly increased, which is essential for a technology testing phase of a burning plasma experiment where fluence (duty cycle) is important.

Energy confinement properties of H-mode discharges in the DIII-D tokamak

Neutral beam heated DIII-D expanded boundary divertor discharges have exhibited ASDEX-like H-mode behaviour over a wide parameter range. The deuterium H-mode energy confinement of 120 ms remained near the Ohmic value for up to 6 MW of neutral beam heating, where it was 2-2.5 times higher than the L-mode value at a plasma current of 1 MA. The hydrogen and helium H-mode energy confinement times were similar and substantially below the deuterium H-mode confinement time. The H-mode confinement times decreased with increasing neutral beam power and were only 30% better than the L-mode confinement times at 5 MW. In an H-mode with a mixture of hydrogen and deuterium ([H]/[H+D] 40%), the confinement time was in between the values obtained in the pure hydrogen and deuterium cases, increased linearly with plasma current for q95 > 3.2, and decreased with increasing neutral beam power. The confinement quality in these plasmas was 85 ms per MA at a heating power of 5.6 MW. The lower energy confinement in the non-deuterium H-modes and the degradation of energy confinement with neutral beam power were both accompanied by an increase in the edge localized mode (ELM) amplitude and frequency. The changing ELM characteristics make a determination of the intrinsic isotopic and neutral beam effect on confinement difficult. For values of BT < 0.9 T and q95 < 3, the confinement quality in the deuterium and hydrogen/deuterium H-modes deteriorated to values near the L-mode level. This deterioration in energy confinement was not related to operation at high beta but instead appears to be due to a combined action between sawteeth and ELMs that becomes more pronounced at low q and low BT. L-mode energy confinement was independent of ion species and in good agreement with Kaye-Goldston scaling. Odajima-Shimomura scaling disagreed with the present L-mode τE data in terms of isotopic mass dependence; their prediction for the hydrogen L-mode exceeds the present measurements by a factor of two.

An advanced plasma control system for the DIII-D tokamak

An advanced plasma control system is being implemented for the DIII-D tokamak utilizing digital technology. This system will regulate the position and shape of tokamak discharges that range from elongated limiter to single-null divertor and double-null divertor with elongation as high as 2.6. Through use of a control algorithm which can be varied in real time this system is expected to provide more precise control of the DIII-D discharge shape than is possible in the analog control system presently in use. In addition to shape control the system will provide a platform for research on real-time optimization of discharge performance. A relatively simple design has been achieved through the use of a small number of very-high-speed digital processors coupled with high-speed data acquisition hardware. The frequency response is expected to be adequate to control the unstable vertical motion of highly elongated discharge shapes.<<ETX>>

Observation of the H-mode in ohmically heated divertor discharges on DIII-D

On DIII-D, periods of improved particle and energy confinement have been observed in low q, low BT divertor discharges with Ohmic heating alone. The Ohmic H-mode has characteristics similar to those of the H-mode produced by auxiliary heating. In the Ohmic H-mode the energy confinement time can have values near those predicted by Neo-ALCATOR scaling and a factor of two above those of non-H-mode Ohmic discharges at similar densities. These observations indicate that the H-mode is not limited to an improvement in confinement over that of the L-mode in auxiliary heating discharges.

High-beta discharges in the DIII-D tokamak

Low-{ital q} ({ital q}{sub 95}{lt}3) double-null divertor discharges with values of the volume-average toroidal beta as high as 9.3% have been operated in the DIII-D tokamak (Fusion Technol. {bold 8}, 441 (1985)). In discharges with {ital q}{sub 95}{approx}5, values of {beta}{sub {ital T}}/({ital I}/{ital aB}) as high as 5 have been obtained. These discharges are shown to be at or below the stability limit to the value of beta for infinite-{ital n}, ideal ballooning modes. The discharges are significantly below the beta limit for ideal,low toroidal mode number kink modes. The kink mode beta limit is shown to be strongly dependent on the radial profiles of plasma pressure and current. The theoretical beta limit in DIII-D is shown to be in the range {beta}{sub {ital T}}/({ital I}/{ital aB})=4 --5 depending on the value of {ital I}/{ital aB}, and this is consistent with the experiment. High-beta discharges have been operated with ion temperature up to 17 keV. Steady-state, high-beta, low-{ital q} operation is demonstrated by a discharge with {ital I}/{ital aB}=2.6, {ital q}{sub 95}=2.7, in which {beta}{sub {ital T}}{gt}7% is maintained for 1.5 sec.

Doublet III operating regimes with improved energy confinement

Over the past three years, extensive energy confinement experiments have been performed on the Doublet III tokamak with up to 8 MW of both hydrogen and deuterium neutral beam heating power. These experiments covered a wide range of plasma parameters and almost a continuum of magnetic configurations from limited to fully diverted. It has been found that the global energy confinement time is linearly proportional to the plasma current and decreases with the application of neutral beam power for all configurations and beam species studied. The majority of our confinement experiments were conducted with discharges limited toward the outside of the vacuum vessel and heated with hydrogen neutral beams. With respect to this standard operating mode, several operating regimes were found with improved values of energy confinement time. Generally, deuterium neutral beam heating is superior to hydrogen beam heating and divertor operation yields superior confinement as compared to limiter operation. Our experiments also indicate that discharges limited toward the inside of the vacuum vessel exhibit somewhat better confinement relative to discharges limited toward the outside. Accordingly, the highest energy confinement values are obtained in high current, well diverted, deuterium beam heated discharges. Deuterium beam heated discharges limited toward the inside yield normalized energy confinement values similar to those from hydrogen beam heated divertor discharges.

Tritium in the DIII-D carbon tiles

The amount of tritium in the carbon tiles used as a first wall in the DIII-D tokamak was measured recently when the tiles were removed and cleaned. The measurements were made as part of the task of developing the appropriate safety procedures for processing of the tiles. The surface tritium concentration on the carbon tiles was surveyed and the total tritium released from tiles samples was measured in test bakes. The total tritium in all the carbon tiles at the time the tiles were removed for cleaning is estimated to be 15 mCi and the fraction of tritium retained in the tiles from DIII-D operations has a lower bound of 10%. The tritium was found to be concentrated in a narrow surface layer on the plasma facing side of the tile, was fully released when baked to 1000°C, and was released in the form of tritiated gas (DT) as opposed to tritiated water (DTO) when baked.