The ITER Poloidal Field (PF) magnet system consists of six coils. Niobium-Titanium (NbTi) is used as superconducting material and cable-in-conduit conductor(CICC) type are used as a conductor. All coils are fabricated by stacking 6 to 9 double-pancakes wound by two-in-hand winding scheme. The six PF coils (PF1 to PF6) are attached to the Toroidal Field (TF) coil cases through flexible plates or sliding supports to allow small radial and vertical displacements. The outer diameters of the coils vary between 8 m and 24 m. Since the PF coil system provides magnetic field for plasma shaping and position control together with the Central Solenoid (CS) coil, it needs to operate in fast pulse mode, leading to induced voltages of up to 14 kV on the coil terminals during operation.PF coils will be procured by the European and Russian domestic agencies under separate procurement arrangements (PA). Manufacturing design, equipment installation including design and commissioning, and component qualification are the required first steps before entering series production. This paper presents the updated design for manufacturing of components such as tail, joint, clamp and protection cover. Their fabrication and assembly methods are also described. The paper concludes with a summary states report on component qualification program.
The poloidal field (PF) magnet system of the International Thermonuclear Experimental Reactor (ITER) consists of six pulsed coils. Each coil comprises independent modules connected to each other through "shaking hands" joints. In the paper the results of the analysis of the electro-magnetic and thermal performance of the joints during the ITER 15 MA plasma scenario are presented. Of special concern is the radial magnetic field component that is particularly high close to the upper and lower edges of the coils. Moreover, the orientation of the joints in the PF coils is such as to give scope to large current loops between the two cables, which could potentially reduce the temperature and current margins to critical levels. The study has been carried out with the code JackPot-AC, which has been recently upgraded to allow a strand-level detailed analysis of lap-type joints.
The ITER toroidal field (TF) system features eighteen coils that will provide the magnetic field necessary to confine the plasma. Each winding pack is composed of seven double pancakes (DP) connected through praying hands joints. Shaking hands joints are used to interface the terminals of the conductor with the feeder and inter-coil U-shaped bus bars.The feasibility of operating plasma scenarios depends on the ability of the magnets to retain sufficient temperature and current margins. In this respect, the joints represent a possible critical region due to the combination of steady state Joule heating in the resistance of the joint and coupling losses and currents in ramped operation.The temperature and current margins of both DP and terminal joints are analysed during the 15 and 17 MA plasma scenarios. The effect on the joint performance of feasible optimization solutions, such as rotation of the terminal joints and sole RRR increase, is explored. The characterization of the TF coil joints is completed by the estimation of the coupling loss time constant for different inter-strand and strand-to-joint resistance values. The study is carried out with the code JackPot-ACDC, allowing the analysis of lap-type joints with a strand-level detail.
The International Thermonuclear Experimental Reactor (ITER) feeder procurement is now well underway. The feeder design has been improved by the feeder teams at the ITER Organization (IO) and the Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP) in the last 2 years along with analyses and qualification activities. The feeder design is being progressively finalized. In addition, the preparation of qualification and manufacturing are well scheduled at ASIPP. This paper mainly presents the design, the overview of manufacturing and the status of integration on the ITER magnet feeders.
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