NSTX Report on FES Joint Facilities Research Milestone 2010

Maingi, R. ; Ahn, J- W. ; Gray, T. K. ; McLean, A. G. ; Soukhanovskii, V. A.

Princeton Plasma Physics Laboratory

关键词: DIVERTORS;    GEOMETRY;    HEAT FLUX;    PLASMA;    SIMULATION;    SOLS;    TRANSPORT Scrape-off Layer (SOL);    Tokamaks;    NSTX;    Plasma-wall Interaction;   

DOI  :  10.2172/1010971
RP-ID  :  PPPL-4610
PID  :  OSTI ID: 1010971
Others  :  TRN: US1102430

学科分类：原子、分子光学和等离子物理

美国|英语

来源: SciTech Connect

【 摘 要 】

Annual Target: Conduct experiments on major fusion facilities to improve understanding of the heat transport in the tokamak scrape-off layer (SOL) plasma, strengthening the basis for projecting divertor conditions in ITER. The divertor heat flux profiles and plasma characteristics in the tokamak scrape-off layer will be measured in multiple devices to investigate the underlying thermal transport processes. The unique characteristics of C-Mod, DIII-D, and NSTX will enable collection of data over a broad range of SOL and divertor parameters (e.g., collisionality ??*, beta ??, parallel heat flux q||, and divertor geometry). Coordinated experiments using common analysis methods will generate a data set that will be compared with theory and simulation.