Research on the stability of spherical torus plasmas at and above the no-wall beta limit is being addressed on the National Spherical Torus Experiment [M. Ono et al., Nucl. Fusion 40, 557 (2000)], that has produced low aspect ratio plasmas, R/a∼1.27 at plasma current exceeding 1.4 MA with high energy confinement (TauE/TauE_ITER89P>2). Toroidal and normalized beta have exceeded 25% and 4.3, respectively, in q∼7 plasmas. The beta limit is observed to increase and then saturate with increasing li. The stability factor βN/li has reached 6, limited by sudden beta collapses. Increased pressure peaking leads to a decrease in βN. Ideal stability analysis of equilibria reconstructed with EFIT [L. L. Lao et al., Nucl. Fusion 25, 1611 (1985)] shows that the plasmas are at the no-wall beta limit for the n=1 kink/ballooning mode. Low aspect ratio and high edge q theoretically alter the plasma stability and mode structure compared to standard tokamak configurations. Below the no-wall limit, stability calculations show the perturbed radial field is maximized near the center column and mode stability is not highly effected by a nearby conducting wall due to the short poloidal wavelength in this region. In contrast, as beta reaches and exceeds the no-wall limit, the mode becomes strongly ballooning with long poloidal wavelength at large major radius and is highly wall stabilized. In this way, wall stabilization is more effective at higher beta in low aspect ratio geometry. The resistive wall mode has been observed in plasmas exceeding the ideal no-wall beta limit and leads to rapid toroidal rotation damping across the plasma core.
Experimental results from NSTX indicate that the snowflake divertor (D. Ryutov, Phys. Plasmas 14, 064502 (2007)) may be a viable solution for outstanding tokamak plasma-material interface issues. Steady-state handling of divertor heat flux and divertor plate erosion remains to be critical issues for ITER and future concept devices based on conventional and spherical tokamak geometry with high power density divertors. Experiments conducted in 4–6 MW NBI-heated H-mode plasmas in NSTX demonstrated that the snowflake divertor is compatible with high-confinement core plasma operation, while being very effective in steady-state divertor heat flux mitigation and impurity reduction. A steady-state snowflake divertor was obtained in recent NSTX experiments for up to 600 ms using three divertor magnetic coils. The high magnetic flux expansion region of the scrape-off layer (SOL) spanning up to 50% of the SOL width λq was partially detached in the snowflake divertor. In the detached zone, the heat flux profile flattened and decreased to 0.5–1 MW/m2 (from 4–7 MW/m2 in the standard divertor) indicative of radiative heating. An up to 50% increase in divertor, Prad in the snowflake divertor was accompanied by broadening of the intrinsic C III and C IV radiation zones, and a nearly order of magnitude increase in divertor high-n Balmer line emission indicative of volumetric recombination onset. Magnetic reconstructions showed that the x-point connection length, divertor plasma-wetted area and divertor volume, all critical parameters for geometric reduction of deposited heat flux, and increased volumetric divertor losses were significantly increased in the snowflake divertor, as expected from theory.
In this paper, a linear model for plasma current, position and shape control based on the plasma rigid motion assumption is presented and implemented in an EAST tokamak simulator. The simulator models the plasma, poloidal field (PF) coils, and power supplies, and is used to verify the control algorithm and optimize control parameters and PF coil current trajectories. Plasma position and shape control has been achieved during the last several EAST operation campaigns due to successful decoupling of plasma current, plasma position and shape. The control logic used and experimental results are described in detail. Diverted plasma shapes, including double null, upper and lower single null, and with elongation up to 2.0, triangularity in the range 0.4–0.6 and X point control accuracy of 1 cm, were successfully controlled. Smooth shape transition in the current ramp-up ensures that volt–seconds are saved and that plasma disruptions are avoided. Such control capability provides the basis for future high performance plasma operation.
Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in the solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.
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