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The plasma confinement of the International Tokamak Experimental Rector (ITER) is provided by the magnetic field generated by 18 toroidal field (TF) coils while 6 poloidal field (PF) and 6 central solenoid coils have the function to drive, shape and pre-heat the plasma. Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the supply of 10 TF coils and 5 PF coils to the ITER project. The ITER Organization (IO) team is instead responsible for the design of such coils as well for the coordination of the activities of the different Domestic Agencies (DAs) producing the different components, and their assembly into the Tokamak. The PF coils utilize NbTi Cable-in-Conduit-Conductor and have different diameters between 8 and 24 meters and weights of up to 400 tons. Regarding the PF coils produced by F4E, so far one has been completed by the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) under a collaboration agreement with F4E. The other 4 PF coils are being produced at the ITER site in Saint Paul lez Durance, France, under F4E supervision. The first of these (PF5) will be completed by July 2020 while the last coil (PF3) will be ready be the end of 2023. The TF coils utilize Nb3Sn conductor and are manufactured with the “Wind, React & Transfer” method. The first TF coil is close to completion and will be delivered to the ITER site in early 2020. Subsequent TF coils will follow at a rate of about one every 3-4 months. In this article we will report on the production status of both PF and TF coils and, in particular, the different manufacturing strategies employed. The main challenges faced so far and the results obtained are also described.

International Fusion Energy Organization (ITER) is one of the most ambitious energy projects in the world today. The poloidal field (PF) coil is cooled down to the required operating temperature (4.2 K) by a forced flow of supercritical helium, which is introduced through the helium inlet. The mechanical reliability requirement on the irregular structure of helium inlet due to the introduction of orifice make the helium inlet one of the critical components of the PF coil cooling system. This paper focuses on the fatigue tests on ITER PF6 coil full-size helium inlet samples to validate whether the fatigue performance of the conduit with the welded inlet can withstand the lifetime of PF6 coil close to the working condition. According to the ITER requirement, the conductor jacket shall withstand a cyclic longitudinal strain (0.95 to 9.5) × 10 -4 during 600 000 cycles at 77K or (1.9-19) × 10 -4 during 30 000 cycles at 77 K, and the equivalent nominal tensile load range we calculated is about (33.5-335) kN. Two samples were designed and prepared to assess the structural design and weld quality. The fatigue test of the sample 1 was carried out with a load range of (33.5-335) kN at 4 Hz and 77 K, and it failed after 593 000 cycles. The scanning electron microscope images and finite element analysis showed that crack initiation area of the fractured sample 1 was at the bottom of the hole. The main reasons were that there were some stress concentrations inside the hole when the tensile load was provided, and the residual burrs assisted the crack initiation. Finally, the modified sample 2 with the completely removed burrs succeed after 600 000 cycles with a nominal load range of (33.5-335) kN. Therefore, it is concluded that the present optimized ITER PF6 helium inlet sample has good fatigue performance and meets the requirements of ITER, which can be utilized for practical applications.