The superconductor test facility, named SUperConducting Conductor Experiment (SUCCEX), is being designed since 2014 and the Korean government has lunched the official project in 2021. The SUCCEX magnet is being developed to evaluate superconducting Cable-in-Conduit Conductor (CICC) samples for the future fusion magnets. This split-pair solenoid magnet system is designed to generate magnetic fields over 15 T in a 600 mm diameter bore, with capability to test SULTAN-like conductor samples. The magnet consists of two types of coils: a high-field inner coil (IC) using high-performance Nb3Sn strands (Jc similar to 1000 A/mm(2) at 4.2 K, 16 T) and a low-field outer coil (OC) using ITER-grade Nb3Sn strands (Jc >1000 A/mm(2) at 4.2 K, 12 T). Operating at 24.8 kA, the magnet achieves a background field of 15.4 T, enabling sample testing about 16 T when combined with sample's self-field. This paper presents recent conceptual design updates of the SUCCEX magnet, specifically focusing on the modification of the inner coil (IC) jacket thickness from 4 mm to 5 mm. The modification aims to enhance the structural integrity of the high-field conductor while maintaining the magnet's overall performance characteristics.

The test facility for high-field superconducting conductor is currently in development in Korea with the main purpose of generating an external magnetic field of up to 16 T (Tesla). This test facility will be mainly used to test superconductors intended for K-DEMO, the next-generation fusion reactor in Korea. Achieving a magnetic field intensity of up to 16 Tesla with low-temperature superconducting materials, such as Nb3Sn, presents a significant engineering challenge. Additionally, a substantial increase in the overall costs for material and fabrication of coils has been shown. Consequently, extensive efforts have been made to address the various engineering challenges and the financial constraints. At present, the project is in the conceptual design phase to navigate through the various restrictions. Various design studies are actively progressing to generate the target magnetic field of 16 Tesla. This paper will discuss the current progress and provide the most recent technical updates concerning the construction of the future superconductor test facility.

A 16 Tesla, large bore superconducting magnet for testing the superconducting Cable-in-Conduit Conductor (CICC) is being designed in parallel with the design of the steady state Korea fusion demonstration reactor (K-DEMO) magnet system. The test facility magnet, named SUperConducting Conductor Experiment (SUCCEX), was designed in 2014 and some design modifications of the magnet have been conducted since 2020. The peak magnetic field of the K-DEMO toroidal field coil is about 16 T, and thus the required background field of the conductor test magnet is expected to be over 15 T in a large bore of 600 mm diameter. The background field will be in addition to the K-DEMO conductor sample's self-field. To reach the target magnetic field, high current density Nb3Sn strands (Jc > 1100 A/mm2 at 4.2 K, 16 T) were employed in the design of the high field CICC. The modified SUCCEX magnet consists of concentric solenoids, with a high field inner coil (IC) and low field outer coil (OC). They are connected in series with each other, and the operating current is about 24 kA. In this study, the overall design concept is presented in relation to the results of electro-magnetic, structural analyses.

For the next generation fusion reactor called K-DEMO, a project of constructing 16 T (Tesla) superconducting conductor test facility, named SUCCEX (SUper Conducting Conductor EXperiment), is currently underway. It would generate the maximum magnetic field of 16 T to develop and test the superconducting magnet for K-DEMO. The SUCCEX coil system consists of a split-pair solenoid magnet, and each magnet is divided into two concentric coils, called IC (Inner Coil) and OC (Outer Coil). The IC and the OC are connected by lap joint to supply current from a power supply to the superconducting magnet. A major challenge is finding a simple and low-resistive joint design. In this study, the design and DC resistance calculation of the joint were performed including a cabling simulation. Finite element analysis method was used for the calculation of contact area of the Nb 3 Sn strands and Cu sole, and the relationship between the calculated contact resistance and pressing depth of piston was derived. The optimal design of the joints was proposed, and the calculated DC resistance was 0.82 n-Ohm.

In the International Tokamak Experimental Reactor (ITER), six Poloidal Field (PF) ring coils are employed to initiate, stabilize, and shape the plasma. The Russian Domestic Agency for ITER is responsible for the supply of the PF1 Coil. The manufacturing process, which includes design, tooling procurement, building preparation, and component qualification, spanned over 10 years and concluded in 2022. This article primarily focuses on the final assembly activities following the vacuum pressure impregnation of the coil's winding pack, with specific emphasis on risk mitigation measures to prevent the failure of the coil's electrical insulation and instrumentation. Further, the factory acceptance tests of the completed coil and their outcomes are described and discussed in comparison with the established requirements. These tests include measuring the electrical insulation performance under partial vacuum conditions, checking for leak tightness of the pressurized hydraulic circuit, and conducting dimensional inspections of the coil's interfaces critical for integration into the ITER tokamak. Finally, the article addresses the transportation of the coil and the upcoming cold test to be conducted at the ITER site.

In the ITER machine, two sets of three pre-compression rings will be installed at the top and bottom of the 18 toroidal field coil structures. The rings will tightly hold the coils with a radial force of 4,800 t per coil against the operating forces. Four other pre-compression rings will serve as spares. Weighing 3.4 t each, with an inner diameter of 5 m, the rings made of pultruded S-fiberglass composite are possibly the largest and most highly stressed composite structures to be designed for a cryogenic environment. All 10 rings have been delivered to ITER and tested at the ITER pre-compression ring facility at a double nominal stress level. The facility is used to measure the stress at rupture of sub-size rings and for acceptance tests of the full size rings. These tests are supplemented by the stress relaxation test at nominal load on a ring equipped with acoustic sensors, to confirm that the creep performances are also adequate. The paper discuss the design of the rings, the manufacturing route, and the test results including the analysis of the acoustic emission signals.

Design studies of the steady-state Korean fusion demonstration reactor (K-DEMO) magnet system have been conducted since 2013. The maximum magnetic field of the K-DEMO toroidal field (TF) coil is expected to be over 16 T. A high field superconducting magnet and a facility for testing the K-DEMO TF conductor samples are also being designed to prepare for the development of the K-DEMO TF conductor. To test straight fusion conductor samples compatible with the K-DEMO TF conductor, it is essential to have a large, 150 mm wide by 120 mm high sample space in the magnet. The required maximum magnetic field of the fusion conductor test magnet is expected to be over 15 T. The field consists of about a 15 T background field and some self-field. Both high current density Nb 3 Sn strands and high temperature superconductor (HTS) tapes are candidates for the superconducting materials for the magnet. In this work, two conceptual design study models of the high field magnet are introduced - a planar racetrack dipole based on HTS tape (I c > 200 A/4-mm at 4.2 K, 18 T) cable-in-conduit conductor (CICC), and a non-planar flared end saddle dipole based on Nb 3 Sn strand (J c > 2600 A/mm 2 at 4.2 K, 12 T) CICC. 2-dimensional magnetic and mechanical analyses of both magnet models were performed and the main features of the models are presented.

The 16 T superconducting magnet for the next generation fusion reactor, called K-DEMO would be a great engineering challenge with current technology. The project on the test facility, named SUCCEX (Super Conducting Conductor Experiment) is prerequisite for the development of K-DEMO superconducting magnet. It would generate an external magnetic field of 16 T for developing and testing K-DEMO superconductors. Also, this project will give a guideline to fabricate future superconducting conductors for the next generation nuclear fusion reactor construction. It will be an essential facility for the production of superconducting magnets for K-DEMO. In 2021, the Korean government has approved this project and the project is in its early stages of construction. The construction strategy and the current progress are discussed in this article.

Fusion for Energy (F4E), the European Domestic Agency for the International Thermonuclear Experimental Reactor (ITER), is responsible for the supply of 5 out of the 6 Poloidal Field (PF) Coils: PF2-PF6. One coil, PF1, is being manufactured by the Russian Federation Domestic Agency (RFDA). While the 10 m diameter PF6 was manufactured by the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and tested in the cold test facility at Cadarache under a collaboration agreement with F4E; coils PF2-PF5 are currently being manufactured on site, close to the Tokamak building, their size ranging from 17 to 24 m diameter and weights from 200 to 400 T. This article describes the final acceptance tests performed on the coils PF5 and PF6, the testing setup, paying special attention to the tests performed before, after and during the cool-down at 80 K. The tests cover a wide range of aspects of the operation at cryogenic temperatures: ranging from the high voltage electrical insulation performance during the potential fault conditions during plasma operation, leak tightness under vacuum and pressure drop behavior of its hydraulic system during operation with forced flow helium. In addition, we will briefly introduce the results for the current center line (CCL) calculation obtained for PF5 and PF6.

The poloidal magnetic field of ITER is provided by 6 Niobium–Titanium (NbTi) coils mainly for plasma shaping and position control. The fabrication of the two lower coils PF6 from China ASIPP and PF5 from European F4E has been completed and the coils installed in the bottom of the Tokamak pit in 2021. This paper presents the challenge of the manufacturing process, which involves manufacturing of several dozens of superconducting double pancakes, requiring bending, insulating and welding steps to keep accuracy of a few mm on dimensions up to 25m. The reproducibility and reliability of the manufacturing processes to manage magnet winding/insulation and cryogenic skills are essential for final quality and schedule robustness. After having successfully completed final thermal cycles for site acceptance in Cadarache, the PF6 and PF5 have been delivered simultaneously and handed over to IO. This paper demonstrates the assembly scenario of the two coils, including assembly tools for transportation, preparation for lifting, lifting tooling design, temporary supports design in cryostat and alignment of positioning on supports from assembly hall to the tokamak pit. The coils are sensitive components and preservation steps are needed to protect them in the pit during other assembly operations. In future, the two coils will sit in pit for several years until mounted onto the TF coils through the support then connected to feeders at the coil terminals. PF6, PF5 will firstly rest on temporary supports, and later shift up by jacks and conical guides to align onto TFC. The alignment & shimming processes and tightening will be presented to show the practical functionality and avoidance of interferences with the other components of the ITER tokamak which achieving high accuracy to avoid error fields.

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.

The development of new generation superconducting magnets for fusion research, such as the ITER experiment, is largely based on coils wound from so-called “Cable-In-Conduit” Conductors (CICCs). CICCs consist of various types of stainless steel jackets, densely filled with compacted superconducting strands, which are cooled by supercritical helium. The design of the various magnet systems, and in particular the ITER Poloidal Field (PF) coils, imposes the use of electrical joints to connect unit lengths of the CICCs. The electrical joints are delicate, electrical resistive components, carefully designed to provide efficient high current transfer while avoiding heat generation. The PF joints are subjected to fast varying magnetic fields that induce currents which, combined with the Joule heating in the resistive joints due to transport current, increase the temperature of the helium. Various characteristics, including electrical performance and mechanical behavior, have been addressed in the past in order to optimize manufacturing for satisfactory joint operation. Here an extensive post-mortem characterization of pre-qualification full-size PF joints is reported. Void fraction, twist pitch, and the current path connection are investigated in order to understand their effect on electrical performance and tune the manufacturing processes.

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.

The ITER PF6 coil is manufactured by winding, insulating the superconducting conductors, and further connecting them electrically and hydraulically. To qualify the insulation design and manufacturing process to be applied to the ITER PF6 coil, two beam-shaped mock-ups taking the form of a 3 x 3 array of conductors combined with the same insulation as in the production coil have been tested in simulated Tokamak operating conditions. All the results of the electrical and mechanical tests, including DC and AC high-voltage withstanding at both room and cryogenic temperature, AC partial discharge, Paschen tests, cryogenic compressive fatigue, push-out, and thermal cycling, met the ITER requirements. Performance of the mock-ups confirmed that the insulation design and manufacturing process are suitable for the ITER PF6 coil, based on that the qualification of the insulation manufacturing process for the double pancake can be started.

A measurement-based numerical reconstruction technique is proposed to control winding geometry of the poloidal coils manufactured for ITER. A detailed coil model is parameterized in terms of the magnetic characteristics. Deviations from the specified coil shape are evaluated in a comparison of reconstructed "ideal" field and data of magnetic measurements. The technique has been validated in the course of quality inspection of the dummy double pancake similar to the pancakes of the ITER PF1 coil. Experimental results are presented.

An advanced computation model of the Poloidal Field coil #1 (PF1) of ITER tokamak is developed that involves a detailed description of the winding geometry. The model describes all basic current-carrying components, including turn joggles and inter-layer joggles of the coil. The model has been validated in comparative simulations with the simplified models and analytical calculations based on Ampere's law. A high computational accuracy of 7 · 10 -7 is demonstrated. The error field produced by PF1 coil winding was estimated to be an order of magnitude below the threshold level. The model is planned to be applied for the quality control of the manufactured coil. The approach utilized in the PF1 coil model is quite generalized and flexible that makes it applicable for the modeling of other coils of the ITER magnet system.

One of the critical components of the ITER poloidal field (PF) coils is the electrical joint connecting two conductor lengths. The lap "shaking hands" joints will operate under variable field, causing parasitic-induced currents in superconducting strands and temperature rise of the strands. Previously, some design changes for decreasing the induced currents in the joints were proposed and assessed with the JackPot-ACDC model. In this paper, we use the same model to compare the behavior of two designs under the reference operation cycle of the PF coils. It is concluded that the joints with the proposed design changes will have sufficient stability margin against thermal and electromagnetic disturbances.

The ITER poloidal field (PF) coil system provides a magnetic field for plasma shaping and position control together with the central solenoid coils; it needs to operate in a fast pulse mode, leading to induced voltages of up to 14 kV on the coil terminals during operation. The PF magnet system consists of six coils. The cable-in-conduit conductors with niobium-titanium (NbTi) superconducting material are used in the coils. All coils are fabricated by stacking six to nine double pancakes wound by two-in-hand winding scheme. The six PF coils (PF1 to PF6) are attached to the toroidal field coil cases through the flexible plates or sliding supports to allow small radial and vertical displacements. The PF coils will be procured by the European and Russian domestic agencies under separate procurement arrangements. To accelerate the PF6 coil schedule, which is one of the critical paths for the ITER schedule, a cooperation agreement has been placed between F4E and ASIPP in China in October 2013 with the CN-DA support. Before starting the manufacturing of the coil, the component qualification has been started, such as the 3 $\times$ 3 conductor mock-up, turn insulation, and helium inlet with the dummy conductors. Corresponding mechanical and electric tests were carried out at room temperature and 77 K. The PF dummy double pancake is also wound to demonstrate the winding. This paper presents the updated design for manufacturing of components. Their fabrication methods are also described. This paper concludes with a summary state report on PF1 dummy winding.

The ITER poloidal field (PF) magnet system is a set of six circular coils attached to the periphery of the toroidal field coil (TFC) structure. While the manufacturing of the PF coils is being launched in three different countries, the ITER Organization (IO) is looking ahead at assembly of PF coils in the tokamak. The PF coils can be grouped in three pairs with respect to their similarities in design and attachment to the TFC, which leads to preparation of different strategies and tooling for the transportation and assembly. The procedures shall also consider the integration, assembly sequence, and environment of the whole tokamak. For example, the two lower coils PF6 and PF5 must be brought in storage position in the tokamak pit and stay there until all the TFCs are assembled, before being attached to them. This paper presents the assembly scenario of the PF coils, including logistics, analysis of assembly tolerances, aligning strategy, and detailing the particularities of each PF coil, from the moment they are delivered to IO until they are connected to the feeder, waiting for the commissioning of the machine.