NSTX Report on FES Joint Facilities Research Milestone 2010

Technical Report · Thu Mar 24 00:00:00 EDT 2011

DOI:https://doi.org/10.2172/1010971· OSTI ID:1010971

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

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 ν*, beta β, 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.

• Research Organization: Princeton Plasma Physics Lab. (PPPL), Princeton, NJ (United States)
• Sponsoring Organization: USDOE Office of Science (SC)
• DOE Contract Number: DE-ACO2-09CH11466
• OSTI ID: 1010971
• Report Number(s): PPPL-4610; TRN: US1102430
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
DIVERTORS
GEOMETRY
HEAT FLUX
PLASMA
SIMULATION
SOLS
TRANSPORT
Scrape-off Layer (SOL)
Tokamaks
NSTX
Plasma-wall Interaction