Abstract
The Correction Coils (CC) of the ITER Tokamak are developed to reduce the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. The proposed system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and using a NbTi cable-in-conduit superconducting conductor (CICC). Within each set, the coils are connected in pairs to produce a toroidal field to reduce the most troublesome, lower order, poloidal mode number fields (m = 1,2,3) in order to operate below the locked mode threshold. The conductor is designed to operate up to 6 T. The winding uses pancakes of one-in-hand conductor (quadpancakes for SCC, octopancakes for TCC and BCC), thus avoiding internal joints. The winding-pack is enclosed inside a 20 mm thick stainless steel casing. The coils are supported by rigid connections to the Toroidal Field (TF) coils. The structural design of the CC is mainly driven by the allowable fatigue stress levels in the conductor jacket, in the case material and in the glass-polyimide electrical insulation system. The boundary conditions on the CC are imposed by the TF coils deformation and the electromagnetic interactions with the PF coils system. The thermo-hydraulic and electrical performance of the CICC is also addressed.
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