The assembly of the feeder machine joint has started under challenging conditions, making it difficult to meet the stringent requirements. The challenges of making machine joints and their surrounding support structure in an industrial environment must be successfully overcome. The quality of the worker skills and process control is managed through a rigorous training and qualification program as well as tool acceptance testing. Full-scale mock-ups provide candidate workers with valuable experience in an environment similar to the construction site. The first-of-a-kind assembly on the segmented superconducting busbars and surrounding structural parts is progressing using an approach in which risks are identified, assessed, and prioritized. Lessons learned during the construction of the ITER machine joints are implemented in subsequent assemblies. Status of the construction on the first-ever feeder joint assembly is reported.

The ITER Magnet Feeders are composed of three main sections which are connected together on the ITER site. During this process, the superconducting busbars which carry the electrical current to the ITER magnets are connected together via superconducting joints. As the magnet system operates under high voltage, it is necessary to apply high voltage insulation over the joints and seamlessly connect it to the existing high voltage insulation on the superconducting busbars. This paper describes the materials and detailed techniques used to apply the high voltage insulation on the superconducting joints in the ITER assembly environment. Prior to performing the work on-site, the entire process was qualified on mock-ups, and a process put in place to qualify the technicians who perform the work. These provisions were put in place so as to ensure the quality of the on-site work. [GRAPHICS]

China has contributed to the manufacturing of the Error Field Correction Coils (CC) and the Magnet Feeders for ITER (International Thermonuclear Experimental Reactor). The manufacturing projects have been carried by ASIPP (Institute of Plasma Physics Chinese Academy of Sciences). In this paper, the lessons learned from these two manufacturing projects will be described with special focus on some key manufacturing processes. These experiences gained from the work carried so far in CC and magnet feeder manufacturing and testing are very valuable not only for the remaining manufacturing tasks of these two projects, but also for similar systems of other Tokamak fusion device.

For the quench detection, the high voltage (HV) instrumentation wires should be led out from the electrical insulation layers of ITER magnet feeders. Based on the latest design, at least four HV instrumentation wires need to penetrate from the busbar joint, which break the integrity of the intact insulation layer. A perforated crack may generate along the instrumentation wire in the insulation layer. At some specific ambient pressure and operational voltage, the crack will cause Paschen discharge inevitably, which is the most serious risk threatening the safe and steady operation of the feeders. There are over 200 busbar joints with the instrumentation wires penetration need to be manufactured on the ITER site, and a reliable and durable insulation structure and technology need to be developed for the extracting HV instrumentation wires of ITER magnet feeder. In this paper, the HV instrumentation wire extraction and the specimen design are presented. The electrical properties of the specimen, including the DC hi-pot test, Paschen test and Partial Discharge (PD) test, are reported and discussed.

Feeder-type joint assembly is identified as a non-conventional process that require robust qualification not only on the process, but also on the skill of the operators so as to grant an access to construction site. The joint assembly qualification sample underwent a test program in regards to different operation conditions in various magnetic fields, currents and helium mass flows, etc. The results support the proven process of the joint assembly performed by knowledgeable personnel who service s the operator training and qualification. It is also aimed to report that the test performance of the qualification sample reaffirms the previous test result. Further verification on the joint characteristics was performed. The first-of-a-kind assembly on the segmented superconducting busbars is being prepared using an approach where risks are identified, assessed and prioritized. On-site assembly readiness on the busbar joint assembly process through an operator qualification program is addressed.

International Thermonuclear Experimental Reactor (ITER) is the largest global cooperation project, which will be built as the world's largest tokamak to explore the commercial production of fusion-based electricity. In order to connect the busbars electrically and hydraulically, the “shaking hands” twin-box type joint was designed and implemented for ITER project. Different from the conventional method, the indium bonding technology was developed instead of the Tin soldering method for the joint connection. In order to qualify for the ITER correction coil (CC) feeder joint structure and the manufacturing technology, Institute of Plasma Physics, Chinese Academy of Science (ASIPP) has launched the joint qualification on the manufacture process and the cryogenic performance. By means of the indium bonding, the joint resistance can be lower than 1 nΩ before and after the mechanical fatigue. This paper reports the structure and manufacturing process of CC feeder joint, describes the establishing of the testing setup, and finally discusses the cryogenic testing results. Besides the joint resistance, the pressure drop and the critical temperature are reported.

•Feeders are the life lines of the ITER magnet system.•The PF4 CFT is the first ITER magnet component to be installed in the tokamak pit.•The Manufacture Inspection Plan (MIP) guarantees a good proficiency of the process of the Receiving, Inspection and Tests (RIT) phase.•Preliminary checks or trials are fundamentals avoiding potential future problems and saving time in the pit in the tokamak building.

The ITER high temperature superconducting current lead is a critical component for the magnet system, which has the benefit of reduction in the heat load of the cryogenic system compare with the conventional lead. The current lead is located in the coil terminal box and dry box at the end of the ITER feeder system. As a warm to cold transition section, the current leads feed the huge current from the power supply system into the coils. At the outer surface of the current lead, one layer of a composite insulation is applied to isolate the high voltage potential of the current lead to the ground potential of the environment. The main body of the current lead is that of a long cylinder, but at the cold termination and the cooling inlet, the local geometry is much more irregular. So the insulation wrapping and curing technology of the current lead had to be developed to acquire its uniform mechanical and electrical performance. Now, the multi-stage autoclave curing technology has been qualified in ASIPP and the series ITER current leads are being manufactured based on the qualified procedure. In this paper, the latest insulation progresses for the ITER current lead are introduced, the high voltage testing results as part of the formal qualification are presented and discussed.

High temperature superconducting current leads (HTS-CL) are designed to supply the current to the large superconducting ITER magnets for the operation with reduced heat load to the cryogenic system. The Toroidal field (TF) current leads are the largest with a current capacity of 68 kA each. The Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) is responsible for the supply of the current leads based on a design jointly developed with the ITER organization. Before the supply of the TF HTS-CL series, a pair of prototypes was manufactured in 2014 by ASIPP and associated manufacturers according to previously qualified manufacturing procedures. Rigorous quality control measures were developed and applied in preparation for series manufacturing. To verify compliance of the prototypes with the ITER specification, thorough testing was conducted in 2015. The test items under particular scrutiny were: the pressure drop in the counter-flow heat exchanger, the loss of flow accident test after steady state operation at 68 kA current, the so-called overheating time of the HTS module following an induced quench, the electrical resistances of the soldered joints inside the lead assembly (i.e., the low temperature superconducting (LTS) to busbar and the HTS to LTS joints), and the conduction heat load per lead to the 4.5 K end. In this paper, the main manufacturing steps are discussed, and test results are presented and discussed.

The ITER Feeder is an important subsystem, which transmits the electrical power to the ITER Tokamak magnet system. In the Feeder, all the HV potential components, including the current leads, the busbars, and the joints, need to be insulated with solid composite materials to electrically isolate the HV potential from ground. Based on over three years of preliminary qualification and experimental comparison, the prepreg tape and relevant curing techniques were finally chosen as the formal feeder insulation material and method. The formal insulation qualification was launched, including the material qualification and the component qualification. Now the static tensile/shear strength, the fatigue tensile strength, the compression-shear strength, the push-out strength and the void content test for the prepreg material qualification has been completed. This paper describes the whole research improvement of ITER feeder electrical insulation qualification activities, introduces the selection of procured materials and the manufacturing trials, and summarizes the formal qualification items and their test results.

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 nondestructivemethod 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.

This paper describes the result of the ITER feeder main busbar joint sample qualification test as confirmation of the requirement of busbar joint resistance, 2 nΩ at 70 kA at zero background field, as well as that of joint performance in various magnetic fields to investigate stability and current distribution characteristics in feeder-type joint box. The results support the quality of the joint manufacturing process for ITER main busbar joint. The qualification sample design was prepared to be tested in SULTAN facility. The SULTAN joint sample consists of joints to be qualified at the level of the peak field and upper terminations. In bottom joints, twin-box feeder-type praying hands configuration is applied. In upper terminations, one of them is made with solder-filled cable for optimum current distribution. The other takes the same length of the copper sole and contact with the busbar cable as those positioned in bottom of the sample. The sample undergoes a test program, which includes joint resistance measurement, ac losses, and stability margin test. The outcomes of those test programs are reported.

ITER magnets are in the final phases of production and are preparing for the upcoming assembly challenges. The ITER magnet team has expressed the need for close-range support labs in order to perform qualification tests, procedure tuning, mockup testing, site acceptance tests, as well as training of assembly staff and logistics management. In the framework of its general support to the ITER project, CEA's Institute for Magnetic Fusion Research proposed a support structure in line with this need on CEA Cadarache premises. In July 2014, the MIFI agreement (Magnet Infrastructure Facilities for ITER) was signed between ITER organization and CEA. It led to the creation of four laboratories and a storage area dedicated to magnet instrumentation components. This paper gives a general description of the work organization in MIFI, and describes the main ongoing activities with specific focuses on critical qualification activities like high-voltage testing and diagnostic of Glass-Kapton-Glass resin impregnated insulation and the scale 1 mock-up and test of the intermediate outer intercoil structure assembly procedure.

•Two curing methods for the pre-preg on the superconducting busbar are researched.•Vaccum bag and silicone rubber is used for pre-preg curing as complement of VPI in fusion filed.•The results of mechanical properties and void content is described and discussed.

The ITER superconducting magnet system consists of 18 toroidal field (TF) coils, 6 poloidal field coils, 6 central solenoid modules, and 18 correction coils. Each coil is connected with a magnet feeder to support and control the high-current operation at cryogenic temperature. This paper presents the structural designs and finite-element (FE) analyses of the special construction of the TF in-cryostat feeders (ICFs). A global finite element model (FEM) was built to assess the distribution of the mechanical loads affected by gravitational, thermomechanical, and electromagnetic forces, as well as the displacement by the coils. Design challenges have been taken into account, and critical issues have been identified by global analysis. Valuable changes have been made based on global analysis results to solve the issues. Displacements obtained from global analysis were applied to local models as boundary conditions. Detailed FE analyses were performed to identify the local stress level in welded connections and the interaction between insulated bus bars and their clamps. The results of global and local analyses verify that the ICF ring box and bus bar jacket of the improved TF ICF design can withstand the operational loads with sufficient safety margin.

This paper presents the key manufacturing and testing processes of the prototype ITER feeder seismic bellows. The design of the double bellows was iterated with the results of analysis based on the Expansion Joint Manufacturers Association (EJMA) standard. Each inner and outer bellows was supported in dedicated molds and formed by a hydraulic pressure machine rated at 800 tons. The double bellows were constructed by welding individual collars to the end flanges. The seismic bellows was tested with cyclic pressurization of the interlayer space to 2 bars absolute pressure for 5 cycles. This was followed by 200 cycles of tensile fatigue test with 90 mm of stretching from the nominal design length. After the mechanical fatigue test, a full tensile test with 315 mm of stretching from its nominal length was conducted. Helium leak tests, with the sensitivity of the helium leak detector set to 1 x 10(-9) Pa m(3)/s of helium, were performed at different stages of pressure and mechanical tests. The prototype bellows was qualified for its leak tightness at all time during the qualification test. (C) 2016 Elsevier B.V. All rights reserved.

The first series components of large D-shaped toroidal field coils (TFC) on the ITER Tokamak project are being fabricated and assembled at European Fusion for Energy (F4E) and Japanese Domestic Agency (JADA) premises since 2013. The TF magnet system consists of 18 individual coils connected in series based on a Nb 3 Sn cable-in-conduit conductors supplied by a 68-kA rated current with an overall 41-GJ stored energy and a peak magnetic field of 11.8 T. One of the key challenges of the construction of the 18 TFCs and their assembly resides in the control of the integration of the large individually manufactured coil components and in the ultimate management of tolerances on the final assembly into the Tokamak pit. This paper presents the integration aspects related to main TFCs subcomponents under fabrication starting from the TF conductor production, the winding of individual double pancakes, and their heat treatment and impregnation. This includes the fabrication of key prototypes for qualification purpose such as helium supply inlets, the electrical joints, and the design of the winding pack insertion into the structural TFC case during the final welding enclosure. Each preassembled 40° sector of a TFCs pair is then integrated into the torus according to tight tolerance requirements to provide both the so-called TF magnetic center line data and to guarantee the final operating wedged design into the inner leg region. The assembly of the coil's terminal is then completed by connecting services through the power feeder busbars, the quench detection high voltage cables and the cryogenics interfaces pipe system.

The Magnet Feeder system in the International Thermonuclear Experimental Reactor (ITER) deploys electrical currents and supercritical helium to the superconducting magnets and the magnet diagnostic signals to the operators. In the current design, the feeders located in the upper L3 level of the Tokamak gallery penetrate the Tokamak coolant water system vault, the biological shield and the cryostat. As a secondary confinement to contain the activated coolant water in the vault in the case of water pipe burst accident, a water barrier is welded between the penetration in the water pipe chase outer wall and the mid-plane of the vacuum jacket of the Feeder Coil Terminal Box (CTB). A thin-wall stainless steel diaphragm with an omega shape profile is welded around the CTB as the water barrier to endure 2bar hydraulic pressure. In addition, the barrier is designed as a flexible compensator to withstand a maximum of 15mm of axial displacement of the CTB in case of helium leak accident without failure. This paper presents the detail configuration, the manufacturing and assembly processes of the water barrier. Test results of the prototype water barrier under simulated accident conditions are also reported. Successful qualification of the design and manufacturing process of the water barrier lays a good foundation for the series production of this subsystem.

The joints connecting the ITER magnet busbars and coils utilize the twin-box "shaking hands" concept: inside a helium leak tight box, a bare cable is pressed in an indium-tinned copper base. To form the joint, two boxes are tightly compressed against each other on the copper side with a layer of indium in between. This concept was chosen to address the different requirements of the joints: to provide low electrical resistance without degradation and need of maintenance during ITER lifecycle, to sustain cyclic electromagnetic and pressure loads, to be easily dismountable in case of failure, to be tolerant to manufacturing and assembly misalignments, to provide low coolant flow impedance, and to limit ac losses and at the same time to facilitate current redistribution in the busbars. This paper presents the design; the key results of the electromagnetic, thermal, and stress analysis; and the major manufacturing and qualification steps. The latter includes cryogenic fatigue test of the joint welds and joint resistance measurements.