共 31 条
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.
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