Test results on systems developed for the SST-1 tokamak

摘要

The steady state superconducting tokamak (SST-1) is a large aspect ratio tokamak, configured to run double null diverted plasmas with significant elongation (κ) and triangularity (δ). Superconducting (SC) magnets are deployed for both the toroidal and poloidal field coils in SST-1. A NbTi based cable-in-conduit conductor (CICC) has been fabricated by M/S Hitachi Cables Ltd., Japan under specification and supervision of the Institute for Plasma Research (IPR). The suitability of this CICC for the SST-1 magnets has been validated through test carried out on a model coil wound from this CICC. Toroidal and poloidal SC magnets have been fabricated and factory acceptance tests have been performed. SC magnets require liquid helium (LHe) cooled current leads, electrical isolators at LHe temperature, superconducting bus bars and LHe transfer lines. Full scale prototypes of these have been developed and tested successfully. SC magnets will be cooled to 4.5 K by forced flow of supercritical helium through the CICC. A 1 kW grade liquefier/refrigerator has been installed and is in final stages of commissioning at IPR. SST-1 deploys a fully welded ultra high vacuum vessel, made up of 16 vessel sectors (VSs) having ports and 16 rings with D-shaped cross-section. To establish the fabrication methodology for this, a fullscale prototype of the vessel with two VSs and three rings has been fabricated and tested successfully. Based on this the fabrication of the VSs and rings is in final stage of fabrication. Liquid nitrogen cooled radiation shield are deployed between the vacuum vessel and SC magnets as well as SC magnets and cryostat, to minimize the radiation losses at the SC magnets. SST-1 will have three different high power radio frequency systems to additionally heat and non-inductively drive plasma current to sustain the plasma in steady state for a duration of up to 1000 s. Ion cyclotron resonance frequency (ICRF) and electron cyclotron resonance frequency (ECRF) systems will primarily be used for heating the plasma while lower hybrid waves will be used for non inductive lower hybrid current drive (LHCD). A neutral beam injection with peak power of 0.8 MW with variable beam energy in range of 10-80 keV will be used as additional auxiliary heating system. A number of prototypes for various critical components have confirmed the fabrication methodology. The fabrication of most of the subsystems is nearing completion and many components have already been accepted on site. Erection and installation of the base of the mechanical structure has already been initiated in the SST hall. This paper reports on the results of the tests on various prototypes and actual components to be used on SST-1 for various subsystems.