The ITER superconducting magnet system consists of 18 Toroidal Field (TF) coils, six Central Solenoid (CS) modules, six Poloidal Field (PF) coils and 18 Correction Coils. The Japan Atomic Energy Agency (JAEA), serving as the Japanese Domestic Agency (JADA) in the ITER project, is responsible for the procurement of nine TF coil winding packs (WP), structures for 19 TF coils (including one spare), and assembly of the WP and the coil structures for nine TF coils. JAEA signed the procurement arrangements for the TF conductor in November 2007 and the ones for the nine TF coils and their TF coil structures in November 2008. The manufacture of the TF conductor has already started and its progress is reported in this special issue. In addition, sub- and full-scale trials were performed to achieve compliance for the manufacturing procedure of the TF coil and its structure. These results are also reported in detail in this special issue. Based on these successful results, JAEA is planning to start manufacturing the first TF coil from 2012 to meet the required schedule and complete the 18th TF coil at the end of 2017 in cooperation with the European Domestic Agency (F4E).
Superconducting conductors are applied in the toroidal field (TF) coils, poloidal field coils, and central solenoid (CS) in the ITER. The Japan Atomic Energy Agency plans to procure 25% of the TF conductors and 100% of the CS conductors. Mass-produced Nb 3Sn superconducting strands for TF conductors have been supplied by two manufacturers since 2008. The total length of the strands is approximately 23,000 km; thus, quality control is extremely important. A statistical process control has been adopted in order to reduce the dispersion of strand performance, and stable performance of the mass-produced strands was achieved. Both manufacturers improved the fabrication yield through mass production. Approximately 72% of the Japanese share in TF strands has been produced as of October 2011.
Cable-in-conduit conductors for ITER toroidal field (TF) coils will be operated at 68 kA and 11.8 T. The cable is composed of 1,422 strands with a diameter of 0.82 mm. There were two options for initial procurement. For option 2, the twist pitches at lower stages are longer than in option 1. Trials were performed to assess the feasibility of these options. In the trials for option 1, the nominal outer diameter of sub-cables and reduction schedule of final cables were evaluated and finalized. In the trials for option 2, problems were encountered at the third stage cabling. These problems were resolved through increasing the die size in that stage and improving the tension balance of the second-stage cables to reduce friction between the die and the cable, and also through avoiding loose twisting at both edges of the third cable. Option 2 was finally selected in 2009 based on superconducting performance enhancement of the cable. After the qualification of the fabrication procedure using fabrication of a 760-m dummy cable and a 415-m superconducting cable, mass production of the cables started in March 2010.
The Japan Atomic Energy Agency (JAEA) is responsible for procuring 25% of the ITER toroidal field (TF) coil conductors as the Japanese Domestic Agency (JADA) for the ITER project. The TF conductor is a circular shaped, cable-inconduit conductor, composed of a cable and a stainless-steel conduit (jacket). The outer diameter and maximum length of the TF conductor are 43.7 mm and 760 m, respectively. JAEA has constructed a new conductor manufacturing facility. Prior to starting the conductor manufacturing, JAEA manufactured a 760 m-long Cu dummy conductor as a conductor manufacturing process qualification, such as processes of welding, cable insertion, compaction and spooling. All manufacturing processes have been qualified and JAEA has started to fabricate superconducting conductors for the TF coils.
Cable-in-conduit (CIC) conductors using Nb 3Sn strands are used in ITER toroidal field (TF) coils. The wound TF conductor must be inserted in the groove of the radial plate (RP), which is part of the mechanical structure supporting the large electromagnetic force. Since the available gap between the conductor and RP groove is 0.5~3 mm, the tolerance of the circumferences of the winding and RP groove are +/- 0.023%. Since a tolerance of approximately +/-0.01% is needed for the RP machining, the conductor length of the winding must be controlled with an accuracy of +/- 0.01%. In this study, in order to resolve the above technical issues, the authors performed several trials on winding as the part of the activities in Phase II of TF coil manufacture. In these trials, the accuracy of the conductor length measurement system using the optical equipment was evaluated, and winding trials were performed on a 1/3-scale double-pancake (DP) winding to demonstrate use of the winding system. In addition, the conductor length of the 1/3-scale DP trial winding was evaluated, and the elongation of the conductor due to bending was clarified.
Cable-in-conduit (CIC) conductors using Nb 3 Sn strands are used in ITER toroidal field (TF) coils. Heat treatment generates thermal strain in CIC conductors because of the difference in thermal expansion between the Nb3Sn strands and the stainless-steel jacket. The elongation/shrinkage of the TF conductor may make it impossible to insert a wound TF conductor into the groove of a radial plate. In addition, it is expected that the deformation of the winding due to heat treatment-based release of the residual force in the jacket may also make it impossible to insert the winding in the groove, and that correcting the winding geometry to allow insertion of the winding may influence the superconducting performance of the TF conductor. The authors performed several trials using heat treatment as the part of activities in Phase II of TF coil procurement aiming to resolve the above-mentioned technical issues, and evaluated the elongations of 0.064, 0.074 and 0.072% for the straight and curved conductors and 1/3-scale double-pancake (DP) winding, respectively. It was confirmed that correction of the deformed winding did not influence the superconducting performance of the conductor.
The manufacturing process for ITER toroidal field (TF) coils must be demonstrated as effective in order to meet the criteria for qualification. Since impregnation of the insulation system is one criterion for qualification in this study, we performed impregnation tests using a cyanate-ester/epoxy blended resin, chosen for its excellent resistance to radiation. To establish the insulation and impregnation procedure in the TF coil manufacturing process, three types of trials were performed: 1) impregnation tests using an acrylic model to fix the impregnation conditions; 2) an impregnation test using a metallic model to confirm that no void remains in the insulation layer after curing in the D-shaped configuration; and 3) insulation and impregnation trials using a 1/3-scale double-pancake (DP) model. The results of these trials were used to determine the insulation and impregnation procedures for manufacturing ITER TF coils.
In an ITER toroidal field (TF) coil, tight tolerances of 1 mm in flatness and a few millimeters in profile are required to manufacture a radial plate (RP), although the height and width of the RP are 13 m and 9 m, respectively. In addition, since cover plates (CPs) should be fitted to a groove in the RP with tolerance of 0.5 mm, tight tolerances are also required for the CPs.The authors therefore performed preliminary and full-scale trials to achieve tight tolerances that meet the required RP manufacturing schedule, such as one RP every three weeks. Before the full-scale trials, preliminary trials were performed to optimize machining procedures, welding conditions and assembly procedures for the RP, and the manufacturing processes for the straight and curved CP segments. Based on these preliminary trial results, full-scale RP and CPs were fabricated. The flatness achieved for the RP is 1 mm, except at the top and bottom where gravity support is insufficient. If the gravity support is suitable, it is expected that a flatness of 1 mm is achievable. The profile of the RP was measured to be within the targeted range, better than 2 mm. In addition, most of the CPs fit the corresponding groove of the RP. Although the issue of hot-cracking in the weld still remains, the test results indicate that this problem can be prevented by improving the geometry of the welding joint.Thus, we can conclude that the manufacturing procedures for RP and CP have been demonstrated.
In ITER Toroidal Field (TF) coils, cover plates (CP) are welded to the teeth of the radial plate (RP) to fix conductors in the grooves of the RP. Though the total length of the welds is approximately 1.5 km and the height and width of the RP are 14 and 9 m, respectively, welding deformation of smaller than 1 mm for local out-of-plane distortion and smaller than several millimeters for in-plane deformation is required. Therefore, laser welding is used for CP welding to reduce welding deformation as much as possible. However, the gap in welding joints is expected to be a maximum of 0.5 mm. Thus, a laser welding technique to enable welding of joints with a gap of 0.5 mm in width has been developed. Applying this technology, a CP welding trial using an RP mock-up was successfully performed. The achieved local flatness, that is, the flatness of the cross-section of the RP mock-up, is 0.6 mm. The analysis using inherent strains, which are derived from the welding test using flat plates, also indicates that better local flatness can be achieved if the initial distortion is zero. In addition, the welding deformation of a full-scale RP is evaluated via analysis using the inherent strain. The analytical results show that in-plane deformation is approximately 5 mm and large out-ofplane deformation, consisting of approximately 5 mm-long wave distortion and a twist of approximately 1.5 mm in the RP crosssection, is generated. It is expected that the required profile can be achieved by determining the original geometry of an RP by simulating deformation during welding. It is also expected that the required local flatness of a DP can be achieved, since out-of-plane deformation can be reduced by increasing the number of RPs turned over during CP welding. A more detailed study is required.
The Japan Atomic Energy Agency (JAEA) has conducted activities in Japan since March 2009 in order to establish the manufacturing procedure for the ITER toroidal field (TF) coil structures. A TF coil structure consists of a TF coil case and components. The activities include ensuring that the structural materials and welding procedure comply with the Japan Society of Mechanical Engineers (JSME) code for fusion devices, and demonstrating the manufacturing method and procedures using full-scale segments of the TF coil structure. From the results of these activities, the JAEA confirmed that the quality control method of actual series TF coil structures complies with the JSME code. Therefore, the quality of structural materials and weld joints using gas tungsten arc welding (GTAW) satisfy the ITER requirements. In addition, the JAEA obtained knowledge regarding the welding deformation of the actual TF coil structures. This paper describes the results of these compliance and development activities for ITER TF coil structures.
Based on the results of the sub- and full-scale trials, the toroidal field (TF) coil and TF coil structure manufacturing procedures were considered. The radial plate (RP) will be manufactured by assembling ten sets of segments using laser-beam welding. The cover plates (CPs) will be manufactured using three different methods, depending on their geometry. For the winding pack (WP), a winding system has been designed that enables measurement of the conductor length with an accuracy of 0.01% for serial production. The assembly procedure and groove types of the narrow-gap tungsten inert gas (TIG) welding for coil structures were determined. Hereafter, technical improvements will be considered, aiming to further optimize manufacturing.