Design of the PF system for EAST(HT-7U) tokamak

Author

Weiyue, Wu ; Baozeng, Li ; Zhuning

Author_Institution

Inst. of Plasma Phys., Chinese Acad. of Sci., Hefei, China

fYear

2003

fDate

14-17 Oct. 2003

Firstpage

436

Lastpage

439

Abstract

EAST(HT-7U) superconducting tokamak is an advanced steady-state plasma physics experimental device. The poloidal field (PF) coil system is used with superconducting conductor. The NbTi cable-in-conduit conduct (CICC) cooled by supercritical helium at 4.5 K is chosen as superconductor for all of the PF magnets. It was consisted of fourteen superconducting coils, located symmetrically about the equator plane of the machine. The PF system was supported by the case of TF coils. The final engineering design of PF system will be report in this paper. The maximum capacity of the volt seconds for EAST(experimental advanced superconducting tokamak) PF system is about 10 Web and the peak magnetic field in PF coils is about 4.5 T.

Keywords

Tokamak devices; niobium alloys; plasma toroidal confinement; superconducting coils; superconducting magnets; titanium alloys; 4.5 K; EAST(HT-7U) superconducting tokamak; NbTi cable-in-conduit conduct; PF magnets; PF system; advanced steady-state plasma physics experimental device; engineering design; equator plane; experimental advanced superconducting tokamak; magnetic field; poloidal field coil system; superconducting coils; superconducting conductor; supercritical helium; Conductors; Niobium compounds; Physics; Plasma devices; Steady-state; Superconducting cables; Superconducting coils; Superconducting magnets; Titanium compounds; Tokamaks;

fLanguage

English

Publisher

ieee

Conference_Titel

Fusion Engineering, 2003. 20th IEEE/NPSS Symposium on

Print_ISBN

0-7803-7908-X

Type

conf

DOI

10.1109/FUSION.2003.1426677

Filename

1426677