A. Foussat

Overview of the ITER Correction Coils Design

The Correction Coils (CC) of the ITER Tokamak are developed to reduce the range of magnetic error fields created by imperfections in the location and geometry of the other coils used to confine, heat, and shape the plasma. The proposed system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and using a NbTi cable-in-conduit superconducting conductor (CICC). Within each set, the coils are connected in pairs to produce a toroidal field to reduce the most troublesome, lower order, poloidal mode number fields (m = 1,2,3) in order to operate below the locked mode threshold. The conductor is designed to operate up to 6 T. The winding uses pancakes of one-in-hand conductor (quadpancakes for SCC, octopancakes for TCC and BCC), thus avoiding internal joints. The winding-pack is enclosed inside a 20 mm thick stainless steel casing. The coils are supported by rigid connections to the Toroidal Field (TF) coils. The structural design of the CC is mainly driven by the allowable fatigue stress levels in the conductor jacket, in the case material and in the glass-polyimide electrical insulation system. The boundary conditions on the CC are imposed by the TF coils deformation and the electromagnetic interactions with the PF coils system. The thermo-hydraulic and electrical performance of the CICC is also addressed.

Design of the HTS Current Leads for ITER

Following the design, fabrication and test of a series of trial leads, designs of the three types of current leads required for ITER have been developed, and targeted trials of specific features are in progress on the way to fabrication and testing of prototype units. These leads are of the hybrid type with a cold section based on the use of high temperature superconductor (HTS) and a resistive section cooled by forced flow of helium gas, optimized for operation at 68 kA, 55 kA and 10 kA. The leads incorporate relevant features of the large series of current leads developed and constructed for the CERN-LHC, relevant features of the trial leads built for ITER, and additional features required to fully satisfy the exigent constraints of ITER with regard to cooling, insulation, and interfaces to feeder and powering systems. In this report a description of the design of the leads is presented, together with plans for the preparation of prototype manufacture and testing at ASIPP.

Research on Manufacture and Enclosure Welding of ITER Correction Coils Cases

Extensive research and analysis has illustrated that there are field errors existed on ITER magnet system due to the misalignment of the coils and winding deviations from the nominal shape. To compensate the errors, correction coils (CCs) are developed and employed on ITER. The CCs consist of 6 top CCs (TCC), 6 bottom CCs (BCC) and 6 side CCs (SCC), arranged toroidally around the machine inside the PF coils and mounted on the TF coils. The CC case provides structural reinforcement to the winding packs. It is made of 316LN stainless steel and has a thickness of 20 mm. The main characteristics of the case are small section (~240 mm × 147 mm), large dimensions (~7 m) over wide angle (~60°) and large bending radius (~8 m). During the manufacture, the SCC case is divided into two L-shaped parts and the B/TCC case is a U-shaped part and a flat cover plate. Both the L-shaped part and the U-shaped part are obtained by respectively welding L-shaped sub-parts and U-shaped sub-parts which are manufactured by extrusion. During the final enclosure welding of the case, the fiber laser multi-pass welding technique with filler wire is proposed as a manufacture route due to its little deformation and narrow heat affected zone. The configuration of the enclosure welding machine, the tests for welding process parameters and the welding procedures are discussed in the paper.

From Design to Development Phase of the ITER Correction Coils

The Correction Coils system (CC) within ITER, is intended to reduce the range of magnetic error fields created by assembly or geometrical imperfections of the other coils used to confine, heat, and shape the plasma. The proposed magnet system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and uses a NbTi cable-in-conduit superconducting conductor (CICC) operating at 4.2 K. The ITER Organization (IO) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly preparing the definition of the technical specifications and the upcoming qualification program for the Correction Coils. The proposed design consists of a one in hand conductor winding without internal joint inserted in a structural casing which reacts the electromagnetic loads. The development of major items such as terminal joints, casing manufacture, and vacuum impregnation system, is an essential phase before the series production which will take place at the premises of the supplier. This paper discusses the key technologies on CC coils and future plans for short sample prototypes fabrication.

Validation of Helium Inlet Design for ITER Toroidal Field Coil

The ITER organization has performed design and its validation tests on a helium inlet structure for the ITER Toroidal Field (TF) coil under collaboration with CERN, KIT, and CEA-Cadarache. Detailed structural analysis was performed in order to optimize the weld shape. A fatigue resistant design on the fillet weld between the shell covers and the jacket is an important point on the helium inlet structure. A weld filler material was selected based on tensile test at liquid helium temperature after Nb3Sn reaction heat treatment. To validate the design of the weld joint, fatigue tests at 7 K were performed using heat-treated butt weld samples. A pressure drop measurement of a helium inlet mock-up was performed by using nitrogen gas at room temperature in order to confirm uniform flow distribution and pressure drop characteristic. These tests have validated the helium inlet design. Based on the validation, Japanese and European Union domestic agencies, which have responsibilities of the TF coil procurement, are preparing the helium inlet mock-up for a qualification test.

Qualification Phase of Key Technologies for ITER Correction Coils

The ITER Magnet system [1] consists of four main coils sub-systems: 18 Toroidal Field Coils (TF-coil), a Central Solenoid (CS), 6 Poloidal Field Coils (PF-coil) and 18 Correction Coils (EFCC). The main contract of the EFCCs supply is awarded to the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) by the Chinese Domestic Agency (CNDA). According to the pre-qualification program, ASIPP is implementing the procurement phase to qualify and validate key technologies and manufacturing methods. The Correction coils qualification activities are conducted within the framework of the procurement arrangement set up between the ITER Organization and CNDA. The paper describes the CC development including first results of the coils winding qualification trials and, qualification of a S-Glass fiber-polyimide based insulation system The CC casing assembly process and the first results of the welding trials are reported. The weld qualification, according to ASTM for 316LN austenitic steel is reported in terms of fracture toughness, fatigue crack growth, and tensile property at 4 K. The Vacuum Pressure Impregnation of CC short mock-up, with low viscosity bisphenol-F (DGEBF) epoxy resin, aims to optimization of the curing and insulation mechanical properties.

Development of Insulation Technology With Vacuum-Pressure-Impregnation (VPI) for ITER Correction Coil

ITER Magnet System includes 18 Correction Coils (CC), made with NbTi cable-in-conduit 10 kA conductor, wound into multiple pancakes. The turn and ground insulation were electrically insulated with a glass tape/polyimide interleaved multilayer composite, which should be impregnated with epoxy resin. Some technical issues remaining have to be addressed in the manufacturing process of these coils. One of the issues is the impregnation and curing of the insulation system, which should achieve good mechanical properties and sufficient high voltage insulation (5 kV). The authors performed manufacturing trials of the insulation system using the vacuum pressure impregnation (VPI) method, which is an effective process to remove defects such as dry spots and over rich resins in the insulation system. In order to simulate and optimize the feasible VPI process, two VPI test samples with the length of 1000 mm have been fabricated. In this paper, we introduce the detailed VPI process including the insulation materials and insulation structures for CC and present the results of the related mechanical and electrical tests performed. The mechanical properties, including tensile strength, inter-laminar shear as well as compressive shear (45 ) were evaluated at room temperature and liquid nitrogen temperature (77 K). In addition, the dielectric strength of the turn insulation and the ground insulation on the test samples were also examined. The results of the tests were discussed versus the design requirements.

Applicability of Non-Destructive Examination to Iter TF Joints

The Iter TF joints are of a twin-box design and the critical parameters of the overall resistance are 1) the contact between cable and termination, and 2) the resistance between two terminations. This paper describes applicability of non-destructive examination (NDE) to these joints. The TFEU joint was adapted to make the joint demountable and the contact area was artificially degraded. The TFEU Joint was measured in the range 30-70 kA, 0-6 T. With no artificial degradation, the resistance of the TFEU Joint was measured to be better than the inter-pancake criterion of 3 nΩ at 2 T, 68 kA. At high fields (6 T) the voltage-current (V I) characteristic of the joint is nonlinear and the resistance is higher than expected. The nonlinearity is worse when the joint is artificially degraded. An FEA model was used to demonstrate that the magneto-resistant coppers contribution to the overall joint resistance is low (<; ~1 nΩ) and does not explain the high field behavior. The nonlinear V I behavior is due to poor current redistribution within the joint, which is related to the resistance of the strand-bundle to copper interface. CRPP is developing a room temperature NDE technique based on resistance profiles to investigate this interface. Resistance measurements at low current and field, or high current and low field, do not guarantee performance at high field; joint tests under the operating conditions are required. Tests on the upper terminations of the TFEU Joint showed that large defects in the contact area between two terminations could be tolerated, when the joint has a good strand-bundle to copper contact resistance and effective current redistribution.

He-Inlet of the Toroidal Field Coil: Qualification and Manufacturing Status

In this paper, we will report on the manufacturing of 6 helium inlet mock-ups for the EU ITER TF coils, and on the results of the mock-up tests and other qualification activities carried out in the European industry on this subject.

Overview of the ITER Toroidal Field Magnet System Integration

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 Nb3Sn 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 coils terminal is then completed by connecting services through the power feeder busbars, the quench detection high voltage cables and the cryogenics interfaces pipe system.