Investigation of lithium as plasma facing materials on HT-7

被引：37

作者：

Hu, J. S. ^{[1
]}

Zuo, G. Z. ^{[1
]}

Li, J. G. ^{[1
]}

Luo, N. C. ^{[1
]}

Zakharov, L. E. ^{[2
]}

Zhang, L. ^{[1
]}

Zhang, W. ^{[1
]}

Xu, P. ^{[1
]}

机构：

[1] Chinese Acad Sci, Inst Plasma Phys, Hefei 230031, Peoples R China

[2] Princeton Plasma Phys Lab, Princeton, NJ 08543 USA

来源：

FUSION ENGINEERING AND DESIGN | 2010年 / 85卷 / 06期

关键词：

Lithium; PFC; Coating; HT-7; SPHERICAL TORUS; CDX-U; TOKAMAK; LIMITER; FTU; PERFORMANCE; REMOVAL; DEVICES; SYSTEM;

D O I：

10.1016/j.fusengdes.2010.08.034

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

First experiment of liquid lithium limiter was successfully carried out on HT-7 tokamak and a few positive results were obtained. The results showed that by using lithium limiter, specially liquid lithium limiter. Ha intensity reduced 20-30%, the emission of CIII and OV decreased about 10-20%, loop voltage had a slight decline, the core electron temperature slightly increased, the particle confinement time increased by a factor of 2, and the energy confinement time increased 20%. After lithium coating, the hydrogen recycling decreased, and core electron temperature increased significantly by a factor of 2. At the same time, after lithium coating, electron density of edge plasmas obviously decreased while electron temperature slightly increased. These encouraging results are very useful for further research of long tray lithium limiter on HT-7 and liquid divertor on EAST. (C) 2010 Elsevier B.V. All rights reserved.

引用

页码：930 / 934

页数：5

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