Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene
The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak （EAST） L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer （SOL） width λq and heat spreading...
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| Published in | Plasma science & technology Vol. 19; no. 4; pp. 28 - 33 |
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| Main Author | |
| Format | Journal Article |
| Language | English |
| Published |
IOP Publishing
01.04.2017
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| Subjects | |
| Online Access | Get full text |
| ISSN | 1009-0630 1009-0630 |
| DOI | 10.1088/2058-6272/aa5802 |
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| Abstract | The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak （EAST） L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer （SOL） width λq and heat spreading S,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current Ip.Strong inverse scaling of the SOL width with Ip has been achieved for both L-mode and H-mode plasmas in the forms of λq,L-mode =4.98 × Ip-0.68 and λq,H-mode =1.86 × Ip-1.08.Similar trends have also been demonstrated in the study of heat spreading with SL-mode =1.95 × Ip-0.542 and SH-mode =0.756 × Ip-0.872.In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor （CFETR）. |
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| AbstractList | The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak (EAST) L-mode and ELM-free H-mode plasmas. The divertor power footprint widths, which consist of the scrape-off layer (SOL) width λq and heat spreading S, are important physical parameters for edge plasmas. In this work, a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current Ip. Strong inverse scaling of the SOL width with Ip has been achieved for both L-mode and H-mode plasmas in the forms of λ q , L - mode = 4.98 × I p − 0.68 and λ q , H - mode = 1.86 × I p − 1.08 . Similar trends have also been demonstrated in the study of heat spreading with S L - mode = 1.95 × I p − 0.542 and S H - mode = 0.756 × I p − 0.872 . In addition, studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current. The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor (CFETR). The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak （EAST） L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer （SOL） width λq and heat spreading S,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current Ip.Strong inverse scaling of the SOL width with Ip has been achieved for both L-mode and H-mode plasmas in the forms of λq,L-mode =4.98 × Ip-0.68 and λq,H-mode =1.86 × Ip-1.08.Similar trends have also been demonstrated in the study of heat spreading with SL-mode =1.95 × Ip-0.542 and SH-mode =0.756 × Ip-0.872.In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor （CFETR）. |
| Author | 邓国忠 刘晓菊 王亮 刘少承 许吉禅 冯威 刘建斌 刘欢 高翔 |
| AuthorAffiliation | Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, People＇s Republic of China University of Science and Technology of China, Hefei 230026, People＇s Republic of China |
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| Cites_doi | 10.1088/1009-0630/13/4/22 10.1016/j.fusengdes.2016.05.016 10.1002/ctpp.200610001 10.1088/1009-0630/13/3/07 10.1016/j.jnucmat.2014.09.065 10.1088/0029-5515/53/12/122001 10.1088/0029-5515/51/8/083028 10.1016/j.jnucmat.2013.01.074 10.1016/j.jnucmat.2013.01.028 10.1088/1009-0630/15/12/04 10.1016/j.jnucmat.2013.01.011 10.1016/j.fusengdes.2014.03.010 10.1016/j.jnucmat.2014.11.076 10.1016/j.fusengdes.2015.06.098 10.1016/0022-3115(87)90355-2 10.1016/j.jnucmat.2013.01.027 10.1063/1.4875721 10.1016/S0022-3115(98)00590-X 10.1088/0741-3335/48/8/003 10.1063/1.3566059 10.1088/1009-0630/15/6/01 10.1088/0029-5515/43/9/308 10.1088/0029-5515/53/9/093031 10.13182/FST47-172 10.1016/j.jnucmat.2011.01.103 10.1016/j.jnucmat.2007.01.063 10.1088/0029-5515/52/1/013009 10.1002/ctpp.2150360233 10.1007/s10894-015-9925-4 10.1103/PhysRevLett.107.215001 10.1016/j.jnucmat.2013.01.029 10.1088/0741-3335/48/6/010 10.1002/1521-3986(200006)40:3/4<328::AID-CTPP328>3.0.CO;2-Q 10.1016/j.jnucmat.2010.10.030 10.1016/S0022-3115(98)00522-4 10.1088/0029-5515/53/11/113024 10.1016/j.jnucmat.2013.01.086 10.1016/S0022-3115(98)00880-0 10.1016/j.jnucmat.2009.01.038 |
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| Copyright | 2017 Hefei Institutes of Physical Science, Chinese Academy of Sciences and IOP Publishing |
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| Notes | Guozhong DENG1,2, Xiaoju LIU1, Liang WANG1, Shaocheng LIU1, Jichan XU1, Wei FENG1, Jianbin LIU1, Huan LIU1 and Xiang GAO1,2（1 Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, People＇s Republic of China 2University of Science and Technology of China, Hefei 230026, People＇s Republic of China） The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak （EAST） L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer （SOL） width λq and heat spreading S,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current Ip.Strong inverse scaling of the SOL width with Ip has been achieved for both L-mode and H-mode plasmas in the forms of λq,L-mode =4.98 × Ip-0.68 and λq,H-mode =1.86 × Ip-1.08.Similar trends have also been demonstrated in the study of heat spreading with SL-mode =1.95 × Ip-0.542 and SH-mode =0.756 × Ip-0.872.In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor （CFETR）. divertor power footprint widths, plasma current, SOLPS 34-1187/TL Institute of Plasma Physics PST-2016-0285.R3 |
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| References | 22 Zwingmann W (35) 2003; 43 23 Liang W (8) 2011; 13 26 Chen Y (24) 2011; 13 Ahn J W (4) 2006; 48 Goldston R J (16) 2012; 52 Petrie T W (41) 2013; 53 Wan B (17) 2009; 49 Chen Y (29) 2011; 51 Wang L (15) 2014; 54 30 31 10 Xuewu C (28) 2013; 15 32 11 33 Fundamenski W (42) 2011; 51 12 34 13 36 37 38 Halpern F D (9) 2013; 53 39 18 19 Chankin A V (25) 2006; 48 Eich T (14) 2013; 53 1 2 Xuewu C (27) 2013; 15 3 5 6 7 40 20 21 |
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| Title | Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene |
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