TY - GEN
T1 - The snowflake divertor
T2 - 38th EPS Conference on Plasma Physics 2011, EPS 2011
AU - Soukhanovskii, V. A.
AU - Ahn, J. W.
AU - Battaglia, D.
AU - Bell, R. E.
AU - Diallo, A.
AU - Gerhardt, S.
AU - Kaita, R.
AU - Kaye, S.
AU - Kolemen, E.
AU - LeBlanc, B. R.
AU - Maingi, R.
AU - McLean, A.
AU - Menard, J. E.
AU - Mueller, D.
AU - Paul, S. F.
AU - Podesta, M.
AU - Raman, R.
AU - Rognlien, T. D.
AU - Roquemore, A. L.
AU - Ryutov, D. D.
AU - Scotti, F.
AU - Umansky, M. V.
PY - 2011
Y1 - 2011
N2 - Introduction The spherical tokamak (ST) is viewed as a candidate concept for future fusion and nuclear science applications [1,2]. Divertor experiments in NSTX, a high-power density medium size ST (R = 0.85 m; a = 0.65 m) with graphite plasma-facing components (PFCs), have demonstrated the features of the inherently compact ST divertor. ITER-scale steady-state peak divertor heat fluxes qpk < 15 MW/m2 and q\\ < 200 MW/m2 have been measured in Ip = l.o -1.2 MA discharges heated by 6 MW NBI [3]. As a result of this and other ST- or NSTX-specific geometry features, e.g., a small in/out SOL power ratio, a small divertor PFC area, an open divertor geometry and reduced divertor volumetric (radiated power and momentum) losses, a reduced operating space of the conventional heat flux mitigation techniques, such as the divertor geometry and radiative divertor, has been found [4,5,6]. Viewing this as the challenge and the opportunity for plasma-material interface (PMI) development, NSTX research is now focusing on developing the PMI for future devices, in particular for NSTX-Upgrade [7], where steady-state qpk < 25 - 40 MW/m2 are predicted based on the present scalings [3].
AB - Introduction The spherical tokamak (ST) is viewed as a candidate concept for future fusion and nuclear science applications [1,2]. Divertor experiments in NSTX, a high-power density medium size ST (R = 0.85 m; a = 0.65 m) with graphite plasma-facing components (PFCs), have demonstrated the features of the inherently compact ST divertor. ITER-scale steady-state peak divertor heat fluxes qpk < 15 MW/m2 and q\\ < 200 MW/m2 have been measured in Ip = l.o -1.2 MA discharges heated by 6 MW NBI [3]. As a result of this and other ST- or NSTX-specific geometry features, e.g., a small in/out SOL power ratio, a small divertor PFC area, an open divertor geometry and reduced divertor volumetric (radiated power and momentum) losses, a reduced operating space of the conventional heat flux mitigation techniques, such as the divertor geometry and radiative divertor, has been found [4,5,6]. Viewing this as the challenge and the opportunity for plasma-material interface (PMI) development, NSTX research is now focusing on developing the PMI for future devices, in particular for NSTX-Upgrade [7], where steady-state qpk < 25 - 40 MW/m2 are predicted based on the present scalings [3].
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M3 - Conference contribution
AN - SCOPUS:84867659389
SN - 9781618395931
T3 - 38th EPS Conference on Plasma Physics 2011, EPS 2011 - Europhysics Conference Abstracts
SP - 53
EP - 56
BT - 38th EPS Conference on Plasma Physics 2011, EPS 2011 - Europhysics Conference Abstracts
Y2 - 27 June 2011 through 1 July 2011
ER -