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Publisher Copyright: IEEE In this article, we suggest and study a novel omnidirectional wireless power transfer (WPT) system based on unusual double-toroidal helix coils. The suggested antenna is a double-helix coil wounded around a torus. It creates two mutually orthogonal magnetic fields. One is toroidal and confined inside the torus, and another one is poloidal and is distributed outside it. The intensities of these two fields can be independently tuned via the geometry of the turns. The coupling coefficients are tailored to ensure high efficiency regardless of the receiver position and orientation, without active circuitry for tuning or control. This topology provides self-tuning functionality which is achieved when the coupling between the transmitter and receiver is minimized while the coupling in the pairs transmitter-repeater and repeater-receiver is strong. The system consists of three transmitter-repeater pairs orthogonal to each other, ensuring omnidirectional functionality. Experimental results confirm self-tunability and high efficiency for any position and orientation of receivers. The proposed double-helix antenna is applicable to different WPT applications where the coupling/decoupling between coils can be freely controlled.

T he Comprehensive Research Facility for Fusion Technology (CRAFT), national big science facility in China, is focused on exploring key technologies and building prototype systems for the Chinese Fusion Engineering Test Reactor (CFETR). The toroidal field (TF) coil is one of the most important parts of CRAFT, targeting at manufacturing a prototype TF coil for CFETR. The TF coil is a gigantic “D” shape, which consists of a winding pack (WP) enclosed in the coil case. The winding pack is composed by embedding three sub-WPs, called low, medium, and high field WPs. The conductor of the high and medium WPs is fabricated from Nb3Sn, follow the “winding and reaction” approach. As one of the most critical steps, turn insulation has to be wrapped after the heat treatment of the conductor. Since the superconducting property of Nb3Sn is deeply sensitive to the strain, the strain must be strictly controlled during the wrapping process. In this article, the concept design of the wrapping system is introduced. Analysis of strain and construction of the wrapping mathematic model are performed and verified by the wrapping trial. Based on the abovementioned results, the detail design of the wrapping system has been completed and the machine is currently being manufactured.

Nine toroidal field (TF) coils have been developed in Japan for the international thermonuclear experimental reactor (ITER). The joint resistance of TF coil should satisfy the requirement of smaller than 3 nano-ohm at 2 T of external magnetic field and 68 kA of transport current. Full-size joint sample (FSJS) tests were performed for joint development and for TF coil manufacture, as part of the process control. 11 FSJS tests are conducted in total. FSJS tests were conducted with assistance from a test faculty in the National Institute for Fusion Science as reported in a previous paper. All FSJS tests successfully satisfied the requirement of resistance less than 3 nΩ at 2 T. Additionally, the TF coil joints are subjected to cyclic electromagnetic force and warm-up/cool-down during the ITER operation. The authors investigated the joint performance for the abovementioned influence. The results showed no degradation in the joint resistance. Thus, the TF joint developed in Japan was qualified successfully.

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.

National Institutes for Quantum and Radiological Science and Technology (QST), which has been Japan Atomic Energy Agency up to March 2016, as Japan Domestic Agency, has responsibility to procure 19 coil cases for ITER toroidal field coils (TFCs) as in-kind components. The TFC case (TFCC) is a large structure having a D-shape with a height of 16.5 m, a width of 9 m, and a weight of about 200 tons. Because TFCC is a large welded structure, severe dimensional tolerance is required, such as a +/−2.0 mm profile. In order to minimize the technical risk during manufacture of the TFCCs, QST has performed trials to establish manufacturing procedures before commencement of the actual manufacturing. After the establishment of manufacturing procedures, QST began the manufacture of the first TFCC in August 2012. In parallel with actual manufacturing processes, some trials have been conducted to rationalize some manufacturing processes, while establishing others. Finally, the first unit of TFCC is scheduled to arrive at coil assembly at the end of 2017.

The Divertor Tokamak Test facility (DTT) is an experimental tokamak machine to be built in Frascati, Italy, at the ENEA research center. During its development, the DTT has gone through several design iterations. Developing a rigorous finite-element methodology to evaluate the performance of all its components has thus been a critical part of the verification phase of each new update. This work summarizes the assessment procedure that currently supports the design of the Toroidal Field magnet system, including the superconducting winding pack ( $\mathrm{\mathbf {Nb_3Sn}}$ ), the casing structure and all of the inter-coil structures. Given the high complexity of the 3D structure to reproduce, we implemented some modeling simplifications to solve the problems. Here we describe the finally adopted methodology next to all the motivations we examined to make sure the results were sound and reliable from an engineering point of view.

139 conductors were manufactured for the magnetic system of the ITER toroidal field coils in total by 6 different teams based on 5 different suppliers of Nb3Sn wires. Despite the fact that all conductors and superconducting wires were manufactured according to the same criteria, there were some differences in the manufacturing technologies of the components themselves, and in the result of the acceptance tests. Samples of some of the conductors passed acceptance tests, where the dependence of Tcs on the number of cycles of inputting the operating current into the samples when they were in a magnetic field was determined. The results of these tests showed that there are not only quantitative, but also qualitative differences in the Tcs (N) dependence. These differences are also observed within the same supplier. It would be useful to understand which parameters most strongly contribute to Tcs (N) in order to possibly improve this characteristic. For this purpose, the article provides a statistical analysis of the test results of 49 samples of conductors. The results are being discussed.

JT-60SA will be the world’s largest superconducting tokamak when it is assembled in 2020 in Naka, Japan (R = 3 m, a = 1.2 m). It is being constructed jointly by institutions in the EU and Japan under the Broader Approach agreement. The assembly of its 400-tonne toroidal field (TF) magnet, designed for an on-axis field of 2.25 T, was completed in July 2018. Consideration of its assembly throughout the design process, with associated consultation and testing, allowed high positional accuracy and hence respect for magnetic field tolerances to be achieved.

Sector Lifting Tool (SLT) is purpose-built tool for the lifting and transferring ITER components. The purpose of SLT is to lift and transfer Vacuum Vessel (VV) and Toroidal Field Coil (TFC) from Upending Tool to Sector Sub-assembly Tool (SSAT). After the sub-assembly at SSAT in ITER assembly hall, 40 °Sector which is composed of VV, VVTS and two TFCs is transferred from SSAT to ITER tokamak pit. Sector Lifting Tool and lifted components are connected with the lifting attachments. The lifting attachments have a same function as adaptor for interface between SLT and lifted component. The weight of assembled 40 °Sector is about 1250 tonnes. SLT is designed to withstand even two times component’s weight considering the safety factor of lifting accessary. This paper describes the design of Sector Lifting Tool and shows that the structural integrity has been verified through analysis. The structural analysis was performed in accordance with the ITER load specification and tooling procedure. In analysis result, the stresses and deflections turned out under the allowable ranges.

China Fusion Engineering Test Reactor (CFETR), the next fusion device in China, is proposed to bridge the gaps between ITER and the demonstration reactor (DEMO). The Water Cooled Ceramic Breeder (WCCB) blanket, one candidate option of breeding blanket conception for CFETR, impacts the toroidal field (TF) ripple in CFETR, as the reduced activation ferritic/martensitic (RAFM) steel was used as its structure material. The purpose of this paper is to investigate and evaluate the ripple effect of WCCB blanket on TF ripple in CFETR. An analysis is performed using three-dimensional finite element model with 16 TF coils and WCCB blanket modules. ANSYS code is employed for this study. The results indicated that the TF ripple was less than 0.5% at all points in target separatrix when it was estimated with TF coils only. We also calculated TF ripple in case of WCCB blanket modules and various port configuration schemes. It is found that TF ripple cannot meet the design requirements by the presence of WCCB blanket modules, even in difference port configuration. Discussions are made on the implication of the WCCB blanket modules and ports on TF ripple. Potential methods for reducing the TF ripple are studied in this paper.

The main components of the ITER tokamak are assembled from the nine sub-assemblies of the 40° sectors. During the sub-assembly, the Toroidal Field coils (TFCs) are installed outside of the Vacuum Vessel Thermal Shield (VVTS) by rotation. The clearance between TFCs and VVTS is very narrow, in relation to the cooling tube end points, which are connected to the thermal shield manifolds. The dimensional analysis shows that in-situ bending of the VVTS cooling tube is required instead of pre-bending. This paper describes the development of the in-situ bending tool and its validations. The specialized tool is designed and the in-situ bending procedure is verified by structural analysis and mock-up test.

Within the Broader Approach Program between Europe and Japan for the early realization of fusion with the construction of the JT-60SA tokamak, ENEA, the Italian Agency for New Technologies, Energy and Sustainable Economic Development, was in charge, among others, to supply ten superconducting Toroidal Field (TF) coils for the JT-60SA magnet system. The related procurement started in 2011 and was concluded in 2018 with the delivery of the tenth coil to the Cold Test Facility in France. The construction of each TF coil module consisted of two main phases: manufacturing of the superconducting winding pack and integration in the corresponding casing structure. With all the foreseen modules already completed, this paper provides a critical review of the contract for the manufacturing of the coils highlighting the main difficulties occurred in the different manufacturing steps and evaluating the performances obtained from the measurements taken at the acceptance tests.

QST, as Japan Domestic Agency in ITER, has responsibility to procure nine ITER toroidal field (TF) coils and all 19 TF coil cases (CC). The nine TF coils are procured by two suppliers to accelerate their production. The series production of double pancakes (DPs) and winding pack (WP) are in progress in both manufacturing lines. In total, 45 DP windings, 30 DP impregnation and curing, and the first WP cold test and the second WP impregnation have been completed. In addition, four CCs have been completed. Assembly of the WP and CC of the first TF coil will start from the beginning of 2019. To qualify the assembly of the WP and CC, trials including those to estimate closure welding deformation and qualification of tooling have been performed. The assembly procedure has been developed based on these results.

In the framework of the EU-JA Broader Approach agreement, the 20 Toroidal Field (TF) coils (18 coils for the tokamak + 2 spares) produced for the JT-60SA tokamak were qualified in cold conditions and at full current (25.7 kA) in self-field conditions and heated up until quench occurs, along standardized testing procedures, at the Cold Test Facility (CTF) installed at CEA-IRFU (Saclay, France). The analysis of the tests results will be presented using simple dynamic models for the coolant exchange with cable in the steady-state regime. In addition to this approach, the coils operating limits in CTF conditions will be explored, integrating the global statistics of their strands' critical performances. The latter are built by processing TF strand performances data obtained during strand production phase. Those statistical data will be used to calculate the TF coils performances in the CTF configuration. Results will be discussed and compared to the TF coils experimental statistics, followed by a tentative quantified interpretation: The results will be transposed into variation of effective macroscopic parameters such as local heat load on pancakes or temperature margin potential increase. Once this macroscopic study is assessed, it stands as the first step toward more refined future analyses. Finally, using the above method, a first predictive application will be conducted on the JT-60SA tokamak operation configuration.

China was responsible for supplying 7.5% of ITER toroidal field (TF) Nb 3 Sn conductors. It has delivered all the required conductors unit lengths and fulfilled the TF conductor package. The conductor package delivery consisted of four phases, in terms of phase I (call for tender), II (qualification process), III (pre-production process), and IV (production process). This paper focuses on work from phases II to IV. The Superconducting Strand Test Laboratory of the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) was the reference laboratory of the Chinese Domestic Agency and undertook the task of Nb 3 Sn strands verification based on the requirements of ITER procurement arrangements. The verification results showed that the strands properties fulfill ITER specifications. In addition, in order to know well the strand properties and supply essential data for conductors' performance prediction and analysis, the mechanical properties and critical current Ic as a function of axial strain, temperature, and magnetic field for the TF strands have been tested and summarized.

In the framework of the Broader Approach program, ENEA is in charge of supplying the Toroidal Field (TF) coil casings for JT-60SA tokamak. ENEA commissioned the manufacture of the full set of eighteen casings for the integration of the TF coils plus two additional spare casings to the company Walter Tosto (WT, Chieti, Italy). The main casing components are one outboard straight leg, an outboard curved leg and three inboard covers. The preliminary design of the casing components has been coordinated by Fusion for Energy (F4E) and the following detail design has been finalized involving the industrial companies in charge of the subsequent integration of the coils. The principal milestones of the five year contract are here summarized. The contract started in 2012 with the realization of two sets of mock-ups representative of the principal casings cross sections of the two legs. In 2013 detail design of the casing components was completed; qualifications of special manufacturing processes were performed and manufacturing procedures were defined. Casing manufacturing activities started in 2014. Seven casings were manufactured during 2015, other seven casings in 2016 and the last four casings were delivered in the first quarter of 2017. The full procurement was then completed with the fabrication of two additional casings in July 2017. This paper provides an overview of the casing procurement describing the production process and manufacturing improvements, reporting the satisfactory acceptance results confirmed by the subsequent proper integration of the superconducting TF coils.

In this paper we analyze fast ion motion in the divertor tokamak test (DTT) device (Albanese et al 2017 Nucl. Fusion 57 016010). It is planned that DTT will be heated through a mix of 45 MW heating power, including 15 MW negative-ion-based neutral beam heating (NNBI) which is currently being developed by Consorzio RFX in Padova, Italy (Agostinetti et al 2019 Fusion Eng. Design 146 441-446). An issue for DTT is that a toroidal field (TF) ripple with a maximum value of about similar to 0.42% (with respect to the on-axis magnetic field B (0)) is expected on the low-field side, and this ripple interacts with fast ions through the rather well-known phenomena of ripple-precession resonances, in addition to prompt losses of ions which do not complete a full orbit in the poloidal plane. We will show that, with the planned geometry of NNBI, prompt losses are negligible, and ripple-precession losses amount to a maximum of 0.15%. The calculations are performed with the guiding center code Orbit using two different equilibria, and a beam with an energy of 400 keV and the injection angle alpha (inj) = 40 degrees (measured w.r.t. the first wall), which corresponds to a pitch of injected particles lambda = v (parallel to)/v approximate to sin alpha (inj) = 0.65. The main resonances are of the form omega (b) - nN omega (d) = 0, omega (b) and omega (d) being the bounce and precession frequency, respectively, N = 18 the ripple periodicity and 3

The paper deals with the modelling and simulation of a Fast Discharge Unit (FDU) for quench protection of the Toroidal Field (TF) magnets of the Divertor Tokamak Test, an experimental facility under design and construction in Frascati (Italy). The FDU is a safety key component that protects the superconducting magnets when a quench is detected through the fast extraction of the energy stored in superconducting magnets by adding in the TF magnets a dump (or discharge) resistor. In the paper, two different configurations of dump resistors (fixed and variable respectively) have been analysed and discussed. As a first result, it is possible to underline that the configuration with variable dump resistor is more efficient than the one with a fixed dump resistor.

The experimentally measured single helical states in the low aspect ratio (R/a=2) RELAX device have been compared with the predictions of a cylindrical relaxation theory and a substantial deviation in terms of some macroscopic characteristics has been found, in particular for cases corresponding to a large reversal of the edge toroidal field. In this paper we consider the effect of the toroidal geometry by employing the VMEC and the RelaxFit equilibrium codes. We show that the inclusion of toroidicity produces significant corrections to the cylindrical states, especially for the deep reversed cases.

The magnetic field, necessary for the plasma confinement in the International Thermonuclear Experimental Reactor (ITER), Cadarache, France, is provided by the 18 Toroidal Field Coils (TFC). In each coil, this magnetic field is produced by circulating a current of 68 kA through 4.5 km of Nb3Sn cable-in-conduit superconductor, which is assembled into a mechanical structure capable of withstanding the huge Lorentz forces produced. The Fusion for Energy, the European Domestic Agency for the ITER, is responsible for the supply of 10 TFC. This article gives an overview of the manufacturing and test processes applied during the series production of all sub- and final assemblies, as well as the production status. Special emphasis will be put on some particular characteristics of the Nb3Sn superconductor and other problems faced during manufacturing and strategies applied to overcome them.

Western Superconducting Technologies produced approximately 35 tons of Nb 3 Sn strands for international thermonuclear experiment reactor (ITER) Toroidal field coils using the internal tin process. All performances of the strands met the ITER requirement well. The delivery of Nb 3 Sn strands was finished in 2015. At the present, several kinds of Nb 3 Sn strands with different diameters and structures have been developed. In this study, the influence of heat treatment on different strand diameters was investigated. The results showed that extending the heat treatment time was effective for increasing Jc in strands with larger diameters. A new layout of Nb 3 Sn strand with more subelements and gap between the filaments was also fabricated at Western Superconducting Technologies Corporation, Ltd. After a final heat treatment step for 250 h at 650 °C, the critical current density in the non-Cu areas of the improved 1.2 mm diameter strand was 1110 A/mm 2 at 12 T and 4.2 K, which was approximately 10% higher than the ITER strand. The hysteresis loss of the strand can be significantly reduced to 288 mJ/cm 3 at ±3 T and 4.2 K. Approximately 57 kg of this improved strand was contributed to the prototype conductor for the Toroidal field coils of the EURO fusion DEMO.

The U.S. Domestic Agency is one of the six suppliers of the toroidal field (TF) conductor for ITER. To qualify conductors per ITER requirements, we prepared or provided to the Swiss Plasma Center (SPC) eight test articles, sixteen conductors total, that were tested in the SULTAN facility at SPC in the ITER relevant conditions. The strands that were used in these SULTAN samples were fully characterized in the laboratories. In this paper, we report both test results and analysis of the conductors’ performance against expected strand performance. The U.S. TF conductors showed a slightly better than average current sharing temperature and a relatively low sensitivity to warm up-cooldown cycles in comparison with other suppliers’ conductors. However, the trend in current sharing temperature versus cycles and warm ups did not saturate, which means a continuing slow degradation of the conductor performance if the number of warm up and cooldown cycles will be significantly higher than expected. The ac losses in the U.S. TF conductors are in line with losses in the other TF conductor suppliers.

The general and detailed design of the JT-60SA toroidal field (TF) coil system was done by all the Voluntary Contributors in the project, CEA, ENEA, and F4E. The French part including the supply of 9 + 1 spare of the 18 TF winding packs and their integration in casings was entrusted mid-2011 to Alstom (Belfort, France), now General Electric. Manufacturing flow definition, drawings, and quality assurance documentation were achieved prior qualification of the critical processes. The procurement and commissioning of the needed tooling was led in parallel and the production was started in January 2014. The first winding pack was ready in the end of 2014. The integration inside the casing started in March 2015 for coil completion end 2015. Then, the coil was delivered to the coil test facility for testing at nominal operation condition ( T = 4.5 K, I = 25, 7 kA) and for qualifying the coil resilience to quench events. In parallel, the successive coils were engaged in the manufacture process. The paper reports the status of the manufacture and gives a feedback on the main issues and on the technical solutions implemented. A focus is made on the mastering of the manufacturing processes, which is highlighted by the learning curve of the TF coils manufacturing.

Quantum and Radiological Science and Technology (QST), a Japan Domestic Agency in ITER, is responsible for procuring nine ITER Toroidal Field coils. QST developed an optimized manufacturing plan to solve technical issues in double-pancake (DP) fabrication and winding pack (WP) fabrication, such as optimization of inter-DP glass fabric thickness to avoid damage of DP and WP insulation and degradation of conductor near joint due to glass fabric compression during WP insulation and impregnation. To this purpose, DP thickness was successfully increased by 3 mm by putting glass fabric during DP insulation. Then, this compensates local distortion, which is originated by cover plate (CP) welding and difficult to correct. This DP expansion made it easier to achieve tight tolerance in the WP height. The series production of DP and WP then proceeded and 31 DP winding, 15 DP impregnation, and curing of first WP impregnation was completed in Japan.

Quantum and radiological science and technology performed a toroidal field (TF) insert coil (TFIC) test in 2016–2017, which was newly developed in 2014 with the ITER TF conductor, in the central solenoid model coil test facility. According to the test program, the TFIC underwent more than 1000 cycles of energization to 68 kA under a 10.8 T background field, which represents the lifetime electromagnetic loads of ITER TF coils. Then, five cycles of warm-up to 290 K and cool-down to a temperature lower than 5 K were also performed. Through the test campaign, the transition of the current sharing temperature ( T cs) was measured under the above energization condition. In parallel, two SULTAN test samples fabricated from the same TF conductor as the TFIC were tested by the similar test sequence to that of the TFIC, and the T cs measurement results of the TFIC and the SULTAN samples were compared. In this paper, the test results mentioned above are reported.

The International Thermonuclear Experimental Reactor (ITER) magnetic system includes 18 toroidal field (TF) coils constructed using Nb3Sn cable-in-conduit superconductor. Each TF coil comprises a winding pack (WP) composed of seven double pancake modules stacked together, impregnated and inserted into a stainless steel coil case. Ten TF coils are being produced in Europe, under the responsibility of Fusion for Energy (F4E, the European Domestic Agency), while the remaining nine TF coils are being produced in Japan. F4E has implemented a strategy dividing the procurement into three packages. One is related to the construction of 70 radial plates (RP), another to the fabrication of 10 WP, and a third to the cold test and coil-case insertion of 10 WP. After 7 years of R&D and qualification activities and of industrial production, the first ITER TF coil WP has been completed in Europe. Factory acceptance tests, including leak, dimensional, and electrical tests at room temperature, were completed in May 2017 and the series production of the remaining nine TF WPs in Europe is underway. The first package has been completed and all 70 RP have been delivered. Commissioning of major tooling for the third package is to be performed at the end of 2017. In this paper, we report on the test of the first TF WP and on the status of the remaining production.

The Sector Sub-assembly Tool is a special tool for assembly of ITER Tokamak and is used to sub-assemble the 40° Tokamak sector which consists of vacuum vessel sector, vacuum vessel thermal shield sector and two toroidal field coils. The sector assembled in the assembly building is a basic and fundamental unit for the construction of the ITER Tokamak. The functions of the Sector Sub-assembly Tool are to support and handle heavy components which weigh up to 1200 t. The Sector Sub-assembly Tool is one of the most important tools to perform the assembly of ITER. To assess the design and structural integrity of the Sector Sub-assembly tool, the structural analyses have been performed under the load combinations according to load specifications. The results of the structural analyses show that the stresses of the Sector Sub-assembly Tool are below the allowable stress. This paper provides briefly the result of structural analysis for the Sector Sub-assembly Tool.

The purpose of the upending tool (UT) is upending the vacuum vessel (VV) 40-degree sectors and the toroidal field coils (TFC) from the horizontal delivery orientations to the vertical assembly orientations. According to ITER assembly procedure, this upending operation is carried out by four hooks of the tokamak crane. VV or TFC upended by UT is transferred to sector sub-assembly tool (SSAT) for sector sub-assembly. For this reason, UT is classified as a lifting accessory in the ITER load specification. This paper describes UT design and upending sequence by two cranes with target components (VV and TFC) and presents the analysis results performed for the final design to verify the structural integrity of UT. The analysis cases for this verification are composed of the horizontal and vertical supporting on the floor and lifting. The simplified FE models for the lifted components (VV and TFC) that are provided from ITER Organization Central Team (IOCT) are used for this structural analysis.

National Institutes for Quantum and Radiological Science and Technology, (QST), as Japan Domestic Agency, has responsibility to procure 9 ITER Toroidal Field (TF) coils. QST completed proto double-pancake (DP) trials aiming at qualification and optimization of manufacturing procedure of TF coil in 2015. Series production of DPs then started and winding of 17 DPs, heat treatment of 16 DPs, fabrication of 15 radial plates (RP), transfer of 11 DPs and cover plate (CP) welding of 9 DPs were completed by end of Sep. 2016. Challenging tight tolerances in conductor length, ±0.01%, was achieved to enable transfer of heat-treated conductor into RP groove. 1 mm flatness was achieved in RP, whose height and width are 13 m and 9 m. In addition, about 2.5 mm flatness was achieved after CP welding by optimizing welding sequence. The DP insulation and impregnation were successfully completed on 5 DPs. The flatness of 2 mm could be achieved for all DPs completed so far and this makes DP stacking easier. These results justify that series production of TF coil DP in Japan is going well.

In the framework of the Broader Approach program, ENEA is in charge of supplying 20 Toroidal Field (TF) coil casings for the superconducting tokamak JT-60SA being assembled in Naka site, Japan. ENEA commissioned the company Walter Tosto (Chieti, Italy) to fabricate two sets of 10 casings each to be delivered to ASG Superconductors (Genoa, Italy) and General Electric (Belfort, France), in charge of the following integration of the winding packs into the casings. The composition of the casing components and the detail design of the interfaces have been finalized under the coordination of Fusion for Energy (F4E) and the agreement of the other parties. Two different sets of mock-ups representative of the components have been realized to validate fabrication methods and special welding processes. The manufacturing activities have been divided in different production steps: composition of the components by cutting, forming and welding and then machining to the final shape. The first seven casings have been completed during 2015 and are being used for the completion of the TF coils. On the base of manufacturing experience of the first casing components and the completion of the first coils, the production process has been improved and the schedule optimized. This paper reports the status of casing manufacturing.

In order to check the performance of the JT-60SA Toroidal Field (TF) coils and hence mitigate their possible fabrication risks, a series of tests have been carried out in the Cold Test Facility (CTF) at CEA Saclay in nominal conditions at 5 K and 25.7 kA. One major test performed is the so called “temperature margin test” during which the inlet helium temperature of the Winding Pack (WP) is controlled to increase progressively to the calculated current sharing temperature of 7.3 K in the high field region to launch the quench. The measurements of voltage, pressure and temperature signals allow us to study the thermo-hydraulic behavior of the JT-60SA coils during a quench. This paper will present the 6 quench tests on 5 TF coils accomplished between February and September 2016, of which the first manufactured coil TFC10 has carried out two quench tests and two pure Fast Discharges without any quenches. It will analyze in particular the temperature and pressure rise in helium, the corresponding physical phenomena after each voltage variation, the four quench dynamic phases, the quench propagation velocity and the dissipated power in the coils.

Tokamak Energy’s ST40 spherical tokamak, operating in a double null diverted configuration, has a design basis of toroidal field Bt = 3 T on-axis (R = 40 cm), and plasma current Ip = 2MA. The procurement and installation of the top and bottom divertors for installation into ST40 was anticipated to be completed on a quick timescale of approximately ten weeks. The difficulties inherent in a tight deadline were exacerbated by the necessary social distancing measures introduced to combat the global crisis of COVID-19. This paper describes the procurement and installation of the divertors, the design is described in the sister paper [1], authored by Daniel Iglesias.