基于TSC的CFETR偏滤器靶板热流数值模拟
Simulation of Target Plate Heat Flux of CFETR Divertor Based on TSC
DOI: 10.12677/NST.2022.102011, PDF,    国家自然科学基金支持
作者: 宋 强, 杨锦宏, 陆 野:安徽大学物质科学与信息技术研究院,安徽 合肥;汪卫华:安徽大学物质科学与信息技术研究院,安徽 合肥;中国科学院等离子体物理研究所,安徽 合肥
关键词: 中国聚变工程实验堆偏滤器TSC热流China Fusion Engineering Test Reactor Divertor TSC Heat Flux
摘要: 位形放电,在程序中添加偏滤器靶板,最外封闭磁面以外的等离子体沿磁力线打到偏滤器靶板上。模拟得到了两种偏滤器位形放电下的靶板热流,重点分析了平顶阶段偏滤器靶板热流分布情况,对下单零偏滤器位形和准雪花偏滤器位形的热流进行了比较。结果表明热流在等离子体电流爬升阶段基本上小于1 MW/m2,在平顶阶段准雪花偏滤器热流峰值为23.34 MW/m2,下单零偏滤器热流峰值为39 MW/m2,准雪花偏滤器靶板热流小于下单零偏滤器。计算结果为下一步热结构多物理场耦合分析提供了热源分布,对CFETR的工程设计具有一定的参考价值。
Abstract: In order to obtain the heat flux distribution of CFETR divertor target plate, TSC (Tokamak Simula-tion Code) program is used to simulate the configuration discharge of Lower Single Null and Quasi- Snowflake divertor of CFETR. The divertor target plate is added to the program, and the plasma outside the outermost enclosed magnetic surface hits the divertor target plate along the magnetic force line. The heat flux of the target plate under the discharge of the two divertor configurations is simulated. The heat flux distribution of the divertor target plate in the flat-top stages is analyzed, and the heat flux of the Lower Single Null divertor configuration and the Quasi-Snowflake divertor configuration are compared. The results show that the heat flow is basically less than 1MW/m2 in the plasma current climbing stage. In the flat-top stage, the peak heat flux of the Quasi-Snowflake divertor configuration is 23.34 MW/m2, the peak heat flux of the Lower Single Null divertor config-uration is 39 MW/m2, and the target heat flux of the Quasi-Snowflake divertor is less than that of the Lower Single Null divertor. The calculation results provide the heat source distribution for the multi physical field coupling analysis of thermal structure in the next step, and have a certain ref-erence value for the engineering design of CFETR.
文章引用:宋强, 杨锦宏, 陆野, 汪卫华. 基于TSC的CFETR偏滤器靶板热流数值模拟[J]. 核科学与技术, 2022, 10(2): 103-115. https://doi.org/10.12677/NST.2022.102011

参考文献

[1] Zheng, J.X., Liu, X.F., Song, Y.T., et al. (2013) Concept Design of CFETR Superconducting Magnet System Based on Different Maintenance Ports. Fusion Engineering and Design, 88, 2960-2966. [Google Scholar] [CrossRef
[2] 高翔, 万宝年, 宋云涛, 等. CFETR物理与工程研究进展[J]. 中国科学: 物理学∙力学∙天文学, 2019, 49(4): 7-14.
[3] 崔学武, 潘宇东, 张锦华, 等. 用B2.5-EIRENE优化偏滤器靶板与第一壁热负载[J]. 核聚变与等离子体物理, 2012, 32(1): 44-50.
[4] Wan, Y.X., Li, J.G., Liu, Y., et al. (2017) Overview of the Present Progress and Activities on the CFETR. Nuclear Fusion, 57, Article No: 102009. [Google Scholar] [CrossRef
[5] Song, Y.T., Wu, S.T., Li, J.G., et al. (2013) Concept Design of CFETR Tokamak Machine. 2013 IEEE 25th Symposium on Fusion Engineering (SOFE), San Francisco, CA, USA, 10-14 June 2013, 1-6. [Google Scholar] [CrossRef
[6] Wan, B., Ding, S., Qian, J., et al. (2014) Physics Design of CFETR: Determination of the Device Engineering Parameters. IEEE Transactions on Plasma Science, 42, 495-502. [Google Scholar] [CrossRef
[7] Chan, V.S., Costley, A.E., Wan, B.N., et al. (2015) Evaluation of CFETR as a Fusion Nuclear Science Facility using multiple system codes. Nuclear Fusion, 55, Article No. 023017. [Google Scholar] [CrossRef
[8] 卯鑫, 宋云涛, 叶民友, 等. CFETR偏滤器概念设计[J]. 原子能科学技术, 2015, 49(S1): 481-486.
[9] 刘秀, 曹磊, 李磊, 等. CFETR偏滤器靶板的概念设计与分析[J]. 核聚变与等离子体物理, 2017, 37(1): 81-86.
[10] 黄文玉, 卢勇, 张龙, 等. CFETR氦冷偏滤器初步设计与分析[J]. 核聚变与等离子体物理, 2021, 41(1): 37-44.
[11] 周一夫, 毛世峰, 赵登, 等. CFETR不同位置充气下的辐射偏滤器模拟研究[J]. 核聚变与等离子体物理, 2020, 40(4): 336-343.
[12] Zhang, C.J., Chen, B., Xing, Z., et al. (2016) Estimation of Peak Heat Flux onto the Targets for CFETR with Extended Divertor Leg. Fusion Engineering and Design, 109-111, 1119-1122. [Google Scholar] [CrossRef
[13] Jardin, S.C., Bell, M.G. and Pomphrey, N. (1993) TSC simulation of Ohmic discharges in TFTR. Nuclear Fusion, 33, 371-380. [Google Scholar] [CrossRef
[14] Jardin, S.C., Pomphrey, N. and Hoffmann, F. (1986) Dynamic Modeling of Transport and Positional Control of Tokamaks. Journal of Computational Physics, 66, 481-507. [Google Scholar] [CrossRef
[15] Hoffman, F., Jardin, S.C., et al. (1992) Application of a New Algorithm to Plasma Shape Control in BPX. Nuclear Fusion, 32, 897-912. [Google Scholar] [CrossRef
[16] Eich, T., Leonard, A.W., Pitts, R.A., et al. (2013) Scaling of the Tokamak Near the Scrape-off Layer H-Mode Power Width and Implications for ITER. Nuclear Fusion, 53, Article No. 093031. [Google Scholar] [CrossRef
[17] Ryutov, D.D., et al. (2015) The Snowflake Divertor. Physics of Plasmas, 22, Article No. 110901. [Google Scholar] [CrossRef