Snowflake divertor configuration studies in National Spherical Torus Experiment

Abstract

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.

Advanced divertor configurations with large flux expansion

“…The SF divertor phase had a profound effect on plasma impurity content: the total carbon inventory N c was reduced by 50-70 %. The observed reduction was attributed to the reduction of carbon physical sputtering fluxes in the SF divertor (due to very low divertor T e ), and to the particle expulsion effect from ELMs that appeared in the SF phase [20]. In the standard divertor Hmode discharge, lithium coatings on lower divertor PFCs reduced recycling and led to modified edge plasma pressure and current profiles and low-n peeling-ballooning mode stabilization [26,27], as the pedestal stability op- close to the kink-balooning stability boundary, and the SF-plus configuration was consistent with improved kink-ballooning stability [23,24,28].…”

“…In NSTX, a significant between-ELM reduction of divertor peak heat flux was measured [19,20]. The SF-minus formation was always accompanied by a stable partial detachment of the outer strike point otherwise inaccessible in the standard divertor at P SOL = 3 MW [29,30].…”

“…In NSTX, the SF configuration was compatible with high confinement plasma operation, with no degradation in H-mode core performance [19,20].…”

“…In both tokamaks, thanks to the open divertor geometry and a flexible plasma control system, ideal SF, SF-plus and SF-minus configurations were obtained [19,20,22,21]. In TCV, SF-plus configurations with σ 0.5 were used to study H-mode access, pedestal stability, ELMs and divertor properties.…”

“…The NSTX divertor plate geometry H-mode discharges. The ion ∇B drift direction was toward the lower Xpoint, and the toroidal field was B t = 0.45 T. Three existing lower divertor coils with currents in the 0.5-5 kA range were used to obtain steady-state SF configurations [18,19,20].…”