Taming the plasma-material interface with the 'snowflake' divertor in NSTX

被引:71
作者
Soukhanovskii, V. A. [1 ]
Ahn, J-W. [2 ]
Bell, R. E. [3 ]
Gates, D. A. [3 ]
Gerhardt, S. [3 ]
Kaita, R. [3 ]
Kolemen, E. [3 ]
LeBlanc, B. P. [3 ]
Maingi, R. [2 ]
Makowski, M. [1 ]
Maqueda, R. [4 ]
McLean, A. G. [2 ]
Menard, J. E. [3 ]
Mueller, D. [3 ]
Paul, S. F. [3 ]
Raman, R. [5 ]
Roquemore, A. L. [3 ]
Ryutov, D. D. [1 ]
Sabbagh, S. A. [6 ]
Scott, H. A. [1 ]
机构
[1] Lawrence Livermore Natl Lab, Livermore, CA 94551 USA
[2] Oak Ridge Inst Sci & Educ, Oak Ridge, TN 37831 USA
[3] Princeton Plasma Phys Lab, Princeton, NJ 08543 USA
[4] Nova Photon Inc, Princeton, NJ 08540 USA
[5] Univ Washington, Seattle, WA USA
[6] Columbia Univ, New York, NY 10027 USA
关键词
H-MODE DISCHARGES; DETACHMENT; POWER;
D O I
10.1088/0029-5515/51/1/012001
中图分类号
O35 [流体力学]; O53 [等离子体物理学];
学科分类号
070204 ; 080103 ; 080704 ;
摘要
Steady-state handling of divertor heat flux is a critical issue for ITER and future conventional and spherical tokamaks with compact high-power density divertors. A novel 'snowflake' divertor (SFD) configuration was theoretically predicted to have significant magnetic geometry benefits for divertor heat flux mitigation, such as an increased plasma-wetted area and a higher divertor volume available for volumetric power and momentum loss processes, as compared with the standard divertor. Both a significant divertor peak heat flux reduction and impurity screening have been achieved simultaneously with core H-mode confinement in discharges with the SFD using only a minimal set of poloidal field coils.
引用
收藏
页数:4
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