The superconducting coils of the ITER magnet system have hundreds of electrical lap joints interconnecting superconducting cables. The joints operate in a magnetic field of up to 4 T, field derivatives of 0.5 T/s, and currents up to 70 kA. The acceptance tests for the toroidal field (TF), poloidal field (PF), and correction coil (CC) coils will be performed at 77 K, before they are assembled in the pit. Hence there will be no possibility to measure the resistance of the joints in the superconducting state before the whole magnet system is enclosed in the Tokamak cryostat. In addition, no reliable nondestructive method has been found to spot the joints with a failure at room temperature. Therefore, the production of the joints relies on the strict adhesion to established robust manufacturing procedures during the qualification phase. As additional quality monitoring, a periodic test of the joint samples manufactured in parallel with a coil fabrication is foreseen to control the reproducibility of the joint electrical performance. In order to qualify the manufacturing procedures, to establish a series production tools and worker teams, a comprehensive qualification program has been set up for manufacturers of the coils in Russia (Poloidal Coil 1, PF1), China (PF6, feeders, CC), Japan (TF), Europe (TF and PF), and USA (Central Solenoid, CS). This program includes a set of mockups manufactured according to the process to be used for the coils and submitted to different tests. They include mechanical testing of materials, electrical tests of full size joint samples, destructive microscopic examination of the joint mockups, and mechanical testing of the full size joint mockups. All tests are carried out in specialized laboratories qualified for this type of work. This paper describes the main items of the qualification program, the tests performed, and the acceptance criteria. The test results are reported and compared to the criteria.

Qualification Program of Lap Joints for ITER Coils

The ITER central solenoid (CS) components are currently being manufactured. This Nb3Sn superconducting magnet will provide the magnetic flux swing required to induce up to 15 MA as plasma current. It includes six identical coils, called modules, stacked on top of each other to form a solenoid, enclosed inside a structure split into nine subsets, to provide vertical precompression and mechanical support. High mechanical stresses in materials and high voltages call for the use of structural materials with high strength and toughness and high dielectric strength insulating materials, respectively. The pulsed operation imposes materials with high fatigue strength at cryogenic temperatures. Unlike for the structure, where large existing manufacturing tools were usable, the modules required the construction of a dedicated manufacturing line. A comprehensive qualification programme is performed at the manufacturers before applying procedures for the production of the CS components. The main characteristics of the CS components, their manufacturing routes and the different elements of the qualification programme are described. The overall plan for the manufacture is reported. The status of the first series production components manufactured is presented as well as the planned delivery schedule to the ITER site.

Manufacture of the ITER Central Solenoid Components

Development of welding and processing technologies of helium inlet for ITER Correction Coil

The uniqueness of the ITER fusion project drives purpose and scope here to fulfil (functional) Procurement Arrangement (PA) requirements for the Central Solenoid (CS) and Correction Coil (CC) magnets' electrical insulation [1–3]; as used and undertaken for covering high voltage testing operations of the magnet coil winding pack (WP) vacuum pressure insulation (VPI) beam qualification, followed in 2016 by start of magnets series production in the USA and China [1, 2]. Useful to be considered common state-of-the-art electrical power components testing practices [3, 4] are presented here as performed to verify that the integrity and insulation of the various components are within acceptable limits during different phases of the fabrication and to certify acceptance on delivery of the complete coils. Initial measurement plans were complemented into electrical tests with typical high voltage engineering implemented measures as a pre-requisite to successfully validate produced results. The above permitted the first quantitative validation of the obtained final production, including life time behavior.

Electrical insulation testing for ITER fusion tokamak

The correction coils (CCs) are a system of 18 superconducting coils composed of top CCs, side CCs, and bottom CCs, and are required to correct asymmetries and reduce magnetic error fields detrimental to the plasma performance of the ITER machine. The CC terminals will be connected with those of the superconducting busbars with twin-box joints. Qualification of the manufacturing procedure of the coil terminals is achieved by performing electrical tests of prototype joints in relevant conditions of current, temperature, and background field (4.5 K, 10 kA, and 2.5 T). In order to control the dc resistance and ac loss in series production, special tooling and manufacturing processes were developed. The main technological steps are conductor dejacketing, nickel removal, and silvering of the cable in the terminal, followed by cable compaction and tungsten inert gas welding of the termination box. To this end, two joint samples were manufactured in 2015 and 2016. The performance of the joint samples was verified before and after 1000 electromagnetic cycles. The qualification tests were carried out in the SULTAN facility and ASIPP facility. The test results show the dc resistance below 5- $\text{n}\Omega $ criterion and ac losses below 7 J/cycle.

Qualification of ITER Correction Coil Superconducting Joint

The system of correction coils (CC) is a component of the ITER magnet system, required to correct toroidal asymmetries and reduce error magnetic fields detrimental for physical processes in the plasma. It includes 18 coils, inserted in between toroidal field coils and poloidal field coils and split into 3 sets of 6 coils each: bottom correction coils (BCC), side correction coils (SCC), and top correction coils (TCC). BCC and TCC are planar coils, whereas SCC are wound on a cylinder. All CC coils are wound using a 10 kA NbTi cable-in-conduit conductor and are manufactured by ASIPP laboratory (Institute of Plasma Physics, Chinese Academy of Sciences), under the responsibility of ITER China. A manufacturing line was installed in 2013 at ASIPP in a dedicated workshop for the construction of the CC. In order to qualify the manufacturing procedures, a comprehensive qualification program has been set up. This program includes a set of mock-ups, manufactured according to the process to be used for the coils and submitted to different tests. These qualification items are winding, insulation and vacuum pressure impregnation, helium inlet/outlet, terminal joints, case material, filler material between winding-pack and case, case assembly, and terminal service box. Qualification of conductor winding, He inlet/outlet manufacture, winding-pack turn and ground insulation installation and impregnation, case material, winding-pack-case filler material is achieved. This included mechanical testing of materials at room and cryogenic temperature in specialized testing laboratories and high-voltage tests performed at the CC workshop. Joint qualification, relying on electrical tests of joints in a dedicated test facility, is nearly complete. Remaining qualification items are case assembly, winding-pack insertion into case, and case closure welding. Manufacture of the first coil started in 2015 and its winding-pack is near completion.

E. Niu

Papers

Qualification of the Manufacturing Procedures of the ITER Correction Coils

Electrical insulation testing for ITER fusion tokamak