Title :
Poloidal Field Coil Configuration and Plasma Shaping Capability in NCT
Author :
Matsukawa, M. ; Tamai, H. ; Fujita, T. ; Kizu, K. ; Sakurai, S. ; Tsuchiya, K. ; Kurita, G. ; Morioka, A. ; Ando, T. ; Miura, Y.
Author_Institution :
Japan Atomic Energy Agency, Ibaraki
fDate :
6/1/2006 12:00:00 AM
Abstract :
This paper describes the latest design of poloidal field coil configuration in the national centralized tokamak (NCT), that is a new design based on the former superconducting coil tokamak JT-60SC. The most notable design change was made for the outer equilibrium field (EF) coils to increase a plasma shaping parameter S(=q95*Ip /a*Bt) and to make off-axis heating using N-NBI possible. As a result, the maximum plasma shaping parameter of ~ 7 was obtained by double null divertor configuration with a lower aspect ratio of A ~ 2.6. It was confirmed that two additional EF coils are useful for the plasma squareness control, but only one is finally adopted for ITER plasma simulation and to keep compatibility of off-axis N-NBI heating
Keywords :
Tokamak devices; fusion reactor design; fusion reactor divertors; plasma confinement; plasma heating; superconducting coils; ITER plasma simulation; JT-60SC; NCT; double null divertor configuration; former superconducting coil tokamak; national centralized tokamak; off-axis N-NBI heating; off-axis heating; outer equilibrium field coils; plasma shaping capability; plasma squareness control; poloidal field coil configuration; Costs; Fusion reactor design; Fusion reactors; Heating; Inductors; Plasma simulation; Plasma sources; Shape; Superconducting coils; Tokamaks; JT-60SC; NCT; plasma shaping parameter; poloidal field coil configuration; squareness;
Journal_Title :
Applied Superconductivity, IEEE Transactions on
DOI :
10.1109/TASC.2006.873273
