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

被引：72

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

来源：

NUCLEAR FUSION | 2011年 / 51卷 / 01期

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

页数：4

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