Preliminary Design of Toroidal Field Coils and Conductors for Superconducting Tokamak HT-7U

摘要

The Tokamak HT-7U under design at Institute of Plasma Physics, Chinese Academy of Sciences (CASIPP) is a nuclear fusion experimental device which is a fully superconducting, consisting of superconducting toroidal field (TF) coils and superconducting poloidal field (PF) coils. This paper describes the TF coil system. The TF coils will use NbTi/Cu cable-in-conduit conductor (CICC) cooled with supercritical helium, and with operating temperature of 4.5 K. The major parameters of TF system are: the toroidal field (Bt) 3.5 T on plasma axis, the maximum field on conductor (B_(max)) 5.8 T at rated current 9.5 kA. The configuration of CICC, conductor design criteria, coils structure and mechanical characters are given.