Design and manufacture of the poloidal field conductor insert coil

The Poloidal Field (PF) coils of ITER (International Thermonuclear Experimental Reactor) will supply the necessary magnetic field to initiate, shape, control and shutdown burning plasmas. The PF coils use NbTi cable-in-conduit superconductors, w

Design and manufacture of a full size joint sample (FSJS) for the qualification of the poloidal field (PF) Insert coil

Within the framework of the European Fusion Programme a FSJS has been designed and manufactured by European Industry using PF coil NbTi superconductor manufactured and supplied by the Russian Federation as part of the R&D for the PF Conductor Insert (PFCI) coil. In addition to the superconductor, this sample contains a number of unique features. In contrast to other samples previously manufactured in Europe, the FSJS has used the Central Solenoid Model Coil (CSMC) joint design with NbTi conductor and a thick square jacket. One leg of the FSJS has had the conductor and sub-petal steel wraps removed before jacketing to assess the difference in the conductor performances between the two configurations. This paper will report on the development and manufacture of the FSJS, in particular the use of the Central Solenoid jacketing and swaging tools for compaction of the conductor and swaging of the joints, the preparation and qualification of the manufacturing route for the joint, the jacketing of a special length without conductor wraps and the high level of instrumentation required for the testing of the joint. The sample has been instrumented with more sensors than any other previous European sample, including temperature sensors, a large number of voltage taps for quench detection and Tcs measurements, quadrupoles to detect uneven voltage distribution, hall arrays for current distribution measurements and saddle coils

European development of the ITER divertor target

The main European contribution to the ITER divertor project was the development of the divertor target with severe operating requirements such as peak heat loads of up to 20 MW/m(2). This development involving EU laboratories and industry included R&D on armour materials, thermo-hydraulics testing, component manufacture, high heat flux testing, design and manufacture of prototypes for later testing. The 4-year EU R&D effort achieved the demonstration of the feasibility of a robust divertor target design based on carbon and tungsten armour. This EU solution has eventually been adopted for the ITER reference design and could be valid also for other ITER high heat flux components such as limiters or baffles. (C) 1999 Elsevier Science S.A. All rights reserved

JT-60SA superconducting magnet system

International audienceThe JT-60SA experiment is one of the three projects to be undertaken in Japan as part of the Broader Approach Agreement, conducted jointly by Europe and Japan, and complementing the construction of ITER in Europe. The superconducting magnet system for JT-60SA consists of 18 Toroidal Field (TF) coils, a Central Solenoid (CS) and six Equilibrium Field (EF) coils. The TF magnet generates the field to confine charged particles in the plasma, the CS provides the inductive flux to ramp up plasma current and contribute to plasma shaping and the EF coils provide the position equilibrium of plasma current and the plasma vertical stability. The six EF coils are attached to the TF coil cases through supports with flexible plates allowing radial displacements. The CS assembly is supported from the bottom of the TF coils through its pre-load structure. The design status of the JT-60SA superconducting magnetic system is reviewed