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

Volume

16

Issue

2

fYear

2006

fDate

6/1/2006 12:00:00 AM

Firstpage

914

Lastpage

917

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*I_p/a*B_t) 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;

fLanguage

English

Journal_Title

Applied Superconductivity, IEEE Transactions on

Publisher

ieee

ISSN

1051-8223

Type

jour

DOI

10.1109/TASC.2006.873273

Filename

1642996