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soukhanovskii_icc_2011_v1.pdf2011-09-07 08:54:05Vsevolod Soukhanovskii
soukhanovskii_icc2011.pdf2011-09-07 08:52:40Vsevolod Soukhanovskii

The snowflake divertor configuration: a game-changer for magnetic fusion devices?

Author: Vsevolod A. Soukhanovskii
Requested Type: Consider for Invited
Submitted: 2011-06-10 09:06:06

Co-authors: V. A. Soukhanovskii, R. E. Bell, S. P. Gerhardt, E. Kolemen, B. P. LeBlanc, J. E. Menard, A. McLean, S. F. Paul, T. D. Rognlien, D. D. Ryutov, F. Scotti, J.-W. Ahn, D. Battaglia, A. Diallo, R. Kaita, S. Kaye, H. W. Kugel, R. Maingi, D. Mueller, M. Podesta

Contact Info:
Lawrence LIvermore National Laboratory
7000 East Ave.
Livermore, CA   94551

Abstract Text:
Recent results from NSTX support the snowflake divertor (SFD) configuration [1] as a promising plasma-material interface (PMI) concept for future magnetic fusion energy devices, through the demonstration of the SFD with significant divertor peak heat flux reduction and impurity control simultaneously with good H-mode confinement. In ITER and future tokamaks, the divertor PMI must be able to exhaust steady-state heat fluxes up to 10 MW/m2 with minimal material erosion. In spherical tokamaks (STs), these requirements are aggravated by the inherently compact divertor geometry. The NSTX, a medium-size low-aspect ratio (A≤1.4) tokamak with high divertor heat flux (qpk ≤ 15 MW/m2, q|| ≤ 200 MW/m2), is well suited for novel divertor configuration studies. The SFD concept uses a second-order null-point created by bringing close two first-order null-points of the standard divertor configuration [1]. In recent NSTX experiments, a steady-state SFD has been maintained for up to 600 ms using three divertor magnetic coils running with pre-programmed currents in 4 MW NBI-heated H-mode discharges of 1.0-1.2 s duration. A dedicated effort to develop a real-time feedback control of the primary and secondary null positions by the NSTX plasma control system is underway. When compared to the standard divertor geometry, the SFD in NSTX showed an increase in plasma-wetted area by 100-200 % and an increased divertor volume (with an X-point connection length increased by 50-100%) [2]. The SFD formation in NSTX was always accompanied by a partial detachment of the outer strike point with an up to 50 % increase in divertor radiation from intrinsic carbon, the peak divertor heat flux reduction from 3-6 MW/m2 to 0.5-1 MW/m2, and a significant increase in divertor volume recombination. High core confinement was maintained with the SFD, evidenced by the τE, WMHD and the H98(y,2) factors similar to those of the standard divertor discharges. Core carbon concentration and radiated power were reduced by 30-70 %, apparently as a result of reduced divertor physical and chemical sputtering in the SFD and ELMs. In the SFD discharges, the MHD stability of the H-mode pedestal region was altered leading to the re-appearance of medium size (ΔW/W=5-10 %), Type I, ELMs otherwise suppressed due to lithium conditioning [3]. Fast measurements showed that impulsive particle and heat fluxes due to the ELMs were significantly dissipated in the high magnetic flux expansion region of the SFD. The benefits of the SFD concept are being combined in NSTX with radiative divertor techniques and lithium-coated plasma facing components in PMI experiments aimed at the next step ST-based toroidal facilities. Supported by the U.S. DOE under Contract DE-AC52-07NA27344, DE-AC02-09CH11466, DE-AC05-00OR22725, DE-FG02-08ER54989.
[1] D. D. Ryutov, Phys. Plasmas 14, 064502 (2007)
[2] V.A. Soukhanovskii et. al, Nucl. Fusion 51 (2011) 012001
[3] R. Maingi et. al, Phys. Rev. Lett. 103 (2009) 0750

Characterization: A6

Consider for Invited

University of Washington

Workshop on Innovation in Fusion Science (ICC2011) and
US-Japan Workshop on Compact Torus Plasma
August 16-19, 2011
Seattle, Washington

ICC 2011