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The paper presents a numerical study on electrodynamic behavior of the inter-pancake twin-box joints of the Poloidal Field (PF) coils of the ITER Magnet System. This work was carried out with the THELMA code, developed at the University of Bologna. The THELMA model was adapted to describe both the superconducting wires and the resistive saddle of the joint in a set of reference configurations representative of a set of experimental tests performed on a joint sample at the SULTAN facility in Villigen, Switzerland. The model was applied to analyze both the DC joint resistance measurements and the AC loss vs frequency curve, in order to obtain a full representation of the joint electromagnetic behavior. The paper reports a comparison between the numerical and experimental results and parametric studies on the impact of the main model parameters on the loss vs frequency curve. [ABSTRACT FROM AUTHOR]

In the frame of the conceptual design studies for the Toroidal Field (TF) coils of DEMO, a solution based on a layer-wound magnet, rectangular-shaped Cable-in-Conduit conductors and W&R manufacturing approach, is being developed. The feasibility and performance of a large-size Nb3Sn conductor operating at about 82 kA in a 13 T field has been proven in the past. Another key technology to be demonstrated for a layer-wound TF coil, is that of a joint between two different conductor grades, to be possibly manufactured in-line during winding. The proposed joint solution would provide the minimum room occupancy, with the joint embedded within the winding pack volume. A joint between two low-field conductor grades, i.e., constituted of a small number of superconducting strands and a high number of stabilization copper wires, has been designed and manufactured. The two conductor lengths are characterized by a different number of superconducting wires and different outer dimensions. A joint sample has been assembled and instrumented, in the configuration allowing testing at the SULTAN facility of the Swiss Plasma Center. Both DC and AC performance of the joint has been characterized at the DEMO TF operating conditions. The present paper reports the main manufacturing steps for the joint and its main test results. The implications on the performance and design approach of the TF coil are also discussed, based on the outcome of such tests. [ABSTRACT FROM AUTHOR]

Partial element equivalent circuit-based methodology appears very attractive for the analysis of fast voltage transients, which may appear across and inside the magnets in fusion devices in case of off-normal events. In this paper, we aim at illustrating how adaptive cross approximation coupled with hierarchical matrix arithmetics can provide an effective method to allow the solution of large-scale problems typical of real devices. [ABSTRACT FROM AUTHOR]

The purpose of the effort described here is to model and monitor the insulation shear bonds between the three conductors of the toroidal field (TF) outer legs. Mechanical failure of the insulation could be a precursor to an electrical failure that could damage the TF inner leg central column, which is more difficult to repair. The shear stress in these bonded layers is proportional to the TF outer-leg out-of-plane (OOP) bending. This is a part of a larger program at National Spherical Tokamak Upgrade to instrument and monitor the structural performance of the machine. Components of the instrumentation plan are shown in this study. Bending of the outer leg due to out-of-plane loads is supported partly by shear in the bond between the three conductors that are bonded together to form the outer leg. Bending stress in the outer conductors provides an indication of the integrity of the shear bond. As a part of the orderly planned increase in operational parameters to 0.8 T (as of Feb 2016 it was at 0.61 T). Ten new FISO gauges were purchased and installed in March 2016. They yielded useful data prior to the forced shut-down due to the failure of PF1aUpper coil. The analytic process used to split out the thermal, in-plane and out-of-plane bending strains is described and compared with the measured results. The measured results provide reasonable benchmarks for the analysis. Ultimately, the main purpose of the instrumentation is to compare coil-to-coil behavior, to watch for consistency. If a coil starts to deviate from the others, the coil can be inspected for possible debonding between the three conductors that make up the outer leg. To properly monitor the full TF system, more channels are needed than Fabry–Pérot systems can provide. The planned fiber-Bragg grating (FBG) system is introduced in this paper. Monitoring and evaluation of the TF outer leg strains are also related to computed quantities that are monitored and protected by the digital coil protection system. Currently, the bending strain is expected to be adequately represented by upper–outer leg global moment sums. This relationship is discussed in this paper. [ABSTRACT FROM AUTHOR]

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 coilWPhas 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 ofmajor 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. [ABSTRACT FROM AUTHOR]

The cable-in-conduit conductors (CICCs) that will be adopted for the ITER central solenoid (CS) coil of the ITER machine have been extensively characterized in the SULTAN facility in Villigen, Switzerland, by means of DC and AC tests. The AC measurements were performed superimposing a sinusoidal magnetic field, with the amplitude of 0.2-0.3 T and frequency in the range of 0.1-5 Hz, to a background constant magnetic flux density of 2 T and 9 T. This paper describes the analysis of the AC loss SULTAN tests of one CICC, identical to that used for the manufacturing of the CS Insert tested in 2015 in the CSMC facility in Naka, Japan. The numerical analysis of these experiments was performed by means of the THELMA code, by developing three models at different levels of discretization (sub-cables of different cabling stages and strands). The comparison between measured and computed losses allows one to estimate the interstrand contact conductances of the tested sample. The model provides a useful insight on the time and space distribution of the power dissipated in the conductor. [ABSTRACT FROM AUTHOR]

TheKarlsruhe Institute of Technology (KIT) agreed to construct and test the high-temperature superconductor (HTS) current leads for the tokamak JT-60SA presently under construction in the frame of the Broader Approach agreement between Europe and Japan. In total 6 HTS-CLs for 25.7 kA for the TF coils and 20 HTS-CLs for 20 kA for the PF and CS coils are needed. In addition, the acceptance tests of the HTS current leads were performed in the test facility CuLTKa at KIT. The manufacturing started in 2014 and the last current leads were tested in July 2017. The main parts of the acceptance test are the determination of the heat load at 4 K and the 50 K He mass flow rate through the heat exchanger as well as the simulation of a loss of flow accident. One test of the PF current leads also included a pulse test at 20 kA to demonstrate their PF operation capability. In this paper, an overview of the manufacture and acceptance tests of the HTS current lead is given. The results for the different current leads are summarized and compared to the specifications. [ABSTRACT FROM AUTHOR]

The International Thermonuclear Experimental Reactor (ITER) superconducting magnet system requires current supplies from 10 kA to 68 kA. The components relating the electrical power and cryogens through the warm-cold barrier to the ITER magnets are the so-called feeders. In total there are 31 such feeders for the ITER magnet system, 9 of them necessary for the toroidal field magnet system are called TF feeders. Each feeder consists of three main units: the In-Cryostat Feeder (ICF), the Cryostat Feed-Through (CFT) and the S-Bend Box (SBB)/Coil Terminal Box (CTB) assembly. This paper presents the latest design of the TF feeder components. [ABSTRACT FROM AUTHOR]

• Ambient temperature Paschen tests of components during assembly are standard test methods. • Global Paschen tests of a superconducting magnet system were executed so far only at ambient temperature. • A cryogenic global Paschen test of a full magnet system allows to validate the quality of the electrical insulation after having passed the thermal stresses during the cooldown. • A cryogenic global Paschen test can be safely executed at a temperature of about 90 K and usig argon as test gas. Superconducting fusion magnets feature large inductances, and operate at high currents and at cryogenic temperatures. In case of a quench, the stored electromagnetic energy need to be immediately discharged, thereby inducing high voltages. High quality electrical insulation withstanding cryogenic temperatures is therefore essential for the safe operation of the magnet system. Local and integral Paschen tests at ambient temperature have become a standard to validate the electrical insulation of electromagnetic components of fusion devices. In order to mitigate the risk from a degradation of the electric insulation during cool-down, a global Paschen test of the complete magnet system at cryogenic temperatures is suggested. This paper focuses on the specific design of JT-60SA fusion device, analyses the boundary conditions for a cryogenic global Paschen test, and demonstrates its feasibility at limited risk during a future cooldown. [ABSTRACT FROM AUTHOR]

In this paper we extend the recently developed CtFD strategy for the analysis of the thermal-hydraulics of a meander flow type heat exchanger, validated against the Wendelstein7-X (W7-X) High Temperature Superconductor (HTS) current lead prototype experiment, to the case of a heat exchanger for a 68 kA HTS current lead. We aim to assess the optimized operative conditions and the influence of different geometrical configurations on the performance of the heat exchanger. [ABSTRACT FROM PUBLISHER]

This paper reflects the status of the manufacturing of 2 Radial Plate prototypes for the EU ITER TF Coils. The production of these prototypes will supply valuable information for the manufacturing of the total required number of 70 Radial Plates, in terms of manufacturing technologies, optimization of cost, manufacturing time and risks. [ABSTRACT FROM PUBLISHER]

The International Thermonuclear Experimental Reactor is an international scientific project with the aim of building a tokamak fusion reactor capable of producing at least 10 times more energy than that spent to sustain the reaction. In a tokamak the fusion reaction is magnetically confined and the toroidal field coil system plays a primary role in this confinement. Fusion for Energy, the European Domestic Agency for ITER, is responsible for the supply of 10 out the 19 toroidal field coils. Their procurement has been subdivided in three main work packages: the production of 70 radial plates (the structural components which will house the conductors), the manufacture of 10 winding packs (the core of the magnet) and cold test and insertion into the coil cases of 10 winding packs. The cold test/insertion work package presents significant technological challenges. These include the cold test of the winding packs 14 m high, 9 m wide and weighing 110 t, the welding and inspection of the 316 LN stainless steel coil case, with welded thicknesses of up to 144 mm accessible only from one side combined with the need to minimize the deformation during the welding process (more than 70 m of weld per coil and up to 90 passes to fill the chamfer) and the resin filling of the coil case after insertion of the winding pack (the total volume to be filled up is about one cubic meter per coil). From 2009 up to mid 2011, F4E has carried out an R&D program in order to investigate the most challenging steps of the manufacturing processes associated to this work package, both to meet the demands of the ITER schedule and to minimize technological risks; in this paper an overview of the results obtained is presented. [ABSTRACT FROM PUBLISHER]

The ITER Magnet System contains 18 Toroidal Field Coils (TFC). These are large D-shaped coils of about 300 t, 17.5-m height and 9-m width. They consist of a Winding Pack (WP) enclosed in a rigid structural steel case, the Toroidal Field Coil Case (TFCC). The WP is a bonded structure of 7 Double Pancakes (DP), each made up of a radial plate (RP) housing the reacted Nb3Sn cable-in-conduit superconductor (CICC), which operate at 4.5 K in supercritical helium. The conductor carries a current of 68 kA in operation to produce a nominal peak field of 11.8 T. The total stored magnetic energy in the 18 TFCs is 41 GJ. While the Japanese and European Domestic Agencies that are in charge of the procurement of the TFCs are progressing with the manufacturing design and the fabrication trials prior to launch the production of the real coils, the ITER Organization (IO) is completing the development and qualification of the most critical items, e.g. cyanate ester and resin blends for the conductor and WP insulation system, the terminal region, the helium inlet, a charged resin system for the filling of the gap between the WP and the TFCC and the general tolerancing especially at the interfaces between the neighboring systems. This paper presents the final design of the TFCs and the results of the developments carried out in the aforementioned areas in the last 2 years. [ABSTRACT FROM PUBLISHER]

The Karlsruhe Institute of Technology (which is a merger of former Forschungszentrum Karlsruhe and Karlsruhe University) is responsible for the design, construction and testing of the high temperature superconductor (HTS) current leads for the stellarator Wendelstein 7-X (W7-X) which is presently under construction at the Greifswald branch of the Max-Planck-Institute for Plasma Physics. The current leads are of the binary type, the HTS part covering the temperature range between 4.5 K and 60 K while the heat exchanger covers the range between 60 K and room temperature being cooled by 50 K He. In total 2 prototypes and 14 series current leads are required with a nominal current of 14 kA and a maximum current of 18.2 kA. The paper describes the design and first test results of the prototype HTS current leads. [ABSTRACT FROM AUTHOR]

Forschungszentrum Karlsruhe has taken over the responsibility for the design, construction and testing of the High Temperature Superconductor (HTS) current leads for two fusion experiments, i.e. the stellarator WENDELSTEIN 7-X. (W7-X) and the satellite tokamak JT-60SA. W7-X is presently under construction at the Greifswald branch of the Max-Planck-Institute for Plasma Physics and consists of 50 non-planar and 20 planar coils with a maximum conductor current of 17.6 kA. In total 14 current leads are required with a nominal current of 14 kA that are mounted upside down with the warm end at the bottom. In the frame of the Broader Approach Agreement between Japan and the EU and concomitantly to the ITER project, the satellite tokamak project JT-60SA has been agreed in 2006. The magnet system of JT-60SA consists of 18 toroidal field coils, 4 central solenoid modules and 7 poloidal field coils. In total 26 leads mounted in vertical, normal position are required. For W7-X and JT-60SA a common basic design will be used which will be adapted to the special needs of the machines. All current leads will be of the Cu-HTS binary type. The HTS part covers the range between 4.5 K and 60 K and is cooled by heat conduction from the 4.5 K end, only. The Cu heat exchanger is cooled with 50 K He and covers the range between 60 K and room temperature. The paper describes the status of the HTS current lead development for W7-X and JT-60SA. [ABSTRACT FROM AUTHOR]

For the superconducting magnet system of the International Thermonuclear Experimental Reactor, ITER, 60 current leads for a total current of more than 2500 kA are needed. The use of High Temperature Superconductor current leads (HTS-CL) will reduce the resultant large refrigerator load considerably. Within the EU Fusion Technology Programme, the Forschungszentrum Karlsruhe and CRPP have developed and built a 70 kA HTS-CL demonstrator optimized for 50 K Helium operation. In 2004-2005, the CL was successfully tested with 50 K as well as 80 K Helium, and finally with LN2. In the meantime, ITER has decided to use HTS-CLs for the magnet system and China has taken the responsibility to provide the sc feeders for ITER including the current leads. In this paper, the performance of a 70 kA HTS-CL in fault conditions will be analysed using the FZK-CRPP demonstrator as a model. In particular, the quench performance and the safety behavior in case of a loss-of-flow accident (LOFA) will be described and an extrapolation of the leads performance in real ITER conditions will be given. Finally the effect of different shunt materials is discussed. [ABSTRACT FROM AUTHOR]

The large-scale use of superconductivity continues to be dominated by applications for which there is generally no conventional option. In these cases, superconductivity has enabled new science and technology that could not exist without it. The fields of fusion energy and high-energy physics (HIEP) have particularly benefited from the application of superconductivity. High magnetic fields are absolutely necessary to achieve the required performance parameters. Even though superconductivity is an enabling technology for these fields, it comes with a number of challenges and market incursion has been very slow. However, thanks to persis- tent ongoing research, improvements in ancillary technology, and the development of new materials, the field is far from exhausted. On the contrary, many new applications in addition to high field magnets for scientific research and medical applications are on the horizon. Several applications are on the verge of accomplishing the much more difficult task of replacing existing conventional technology. This paper summarizes the current status of large-scale applications of superconductivity, using as an example magnet technology for fusion and HEP, and examines the prospects and challenges for new large-scale applications in the future. [ABSTRACT FROM AUTHOR]

For the superconducting magnet system of the International Thermonuclear Experimental Reactor, ITER, 60 current leads for a total current of more than 2500 kA are needed. To reduce the resultant large refrigerator load at 4.5 K, High Temperature Superconductor current leads (HTS-CL) could be used. Therefore, EFDA CSU Garching had launched a development program for a 70 kA HTS-CL demonstrator. The Forschungszentrum Karlsruhe and CRPP developed and built this CL optimized for 50 K Helium operation. In 2004, the CL was successfully tested in the TOSKA facility at the Forschungszentrum Karlsruhe. The very encouraging results lead to testing this CL with 80 K Helium because ITER provides a large 80 K Helium cooling capacity for the thermal shields. At the end of last year, the test could be successfully performed demonstrating that high current capacity current leads can be stably operated at about 80–85 K. Recently, the CL was retested using liquid nitrogen which would be an interesting alternative option. In this paper, the test results for the 80 K He cooling operation as well as for the operation with LN2 are presented and compared to the results obtained before for nominal conditions. [ABSTRACT FROM AUTHOR]

In the frame of the European Fusion Technology Programme, the Forschungszentrum Karlsruhe and the CRPP Villigen have designed and built a 70 kA current lead for the ITER TF Coils using High Temperature Superconductors (HTS). At the beginning of 2004 the HTS current lead was installed and tested in the TOSKA facility of the Forschungszentrum Karlsruhe. The scope of the experiment was to characterize the current lead in steady state conditions and to explore the operation limits as well. For this, the temperature profile, the contact resistances, the heat load at 4.5 K, the required 50 K He mass flow rate, and the temperature margin were evaluated. The safety margin in case of a loss of He mass flow was studied, too. The paper describes the experimental results as well as the thermal and electrical models developed. [ABSTRACT FROM AUTHOR]

In this paper we illustrate a self-consistent procedure to design the superconducting Toroidal Field (TF) and Central Solenoid (CS) magnet systems starting from the basic plasma dimensioning parameters. By defining the plasma major radius and minor radii, as well as the inboard TF coil neutron shielding radial built, we proceed by sizing the TF Winding Pack (WP), its case and the CS modules. The ITER magnets design philosophy and criteria are adopted throughout the study. A key outcome of this work is to illustrate the inter-play between machine size, magnetic field and availability of inductive flux for burn. In doing so we investigate the advantages brought by high field tokamak design compared the more conventional (ITER-like) Nb3Sn based magnet design solutions. [ABSTRACT FROM AUTHOR]

The European Roadmap to Fusion Electricity (Federici et al., 2018) [1] details the path to complete within the next three decades the DEMOnstration power plant, DEMO, aiming to a net gain of Energy Q=40. The 2018 DEMO baseline considers a 2 GW tokamak device, a target performance of 6 T on the plasma, with a major radius of 9 m. Considering the plasma aspect ratio and elongation, and the radial build of the thermal shield, the outer radius of the TFC inner leg is about 4.3 m. In the current designs, the TFC winding pack operates in a range of B = 12−13 T and is based on low temperature superconductors (LTS). As an alternative, High Temperature Superconductors (HTS) could be employed to extend the coil operation beyond the typical J/T/B range of Nb3Sn. The rationale behind the exploration of a fully HTS or Hybrid HTS/LTS design of the TFC Winding Pack (WP) is twofold: on the one hand, it is possible to maintain the target performance of the machine redesigning the radial build of each magnet as a function of a more efficient cable in terms of critical current density; on the other hand, it is possible to keep the machine power constant, increasing the field on the plasma axis, and consequently define a new radial build. This work presents and describes a combined algorithm implemented to enhance the DEMO magnet systems and optimize the DTT central solenoid, capable of accounting for the main electromagnetic and structural constraints, and consistently reshape the machine. • This paper describes the MAgnet Design Explorer algorithm (MADE). • The aim of this tool is improving the capability of the existing design codes. • Characterize the magnet radial-builds inside the reactor. • Intertwine physical, mechanical and manufacturing aspects of the fusion magnets. • Compare different superconducting technologies highlighting advantage and disadvantage. [ABSTRACT FROM AUTHOR]

The structural integrity of superconducting magnets that are key elements of a fusion reactor must be ensured. At an early design stage relatively simple calculation tools can greatly facilitate design optimization. The main objective of this paper is the mechanical pre-dimensioning of the tokamak toroidal field coils by simple means prior to the global 3D numerical modeling. A semi-analytical calculation tool that reasonably estimates the static strength of the toroidal field coil under the electromagnetic forces at the critical location (inner leg equatorial plane) is described. The novelty of the approach is that it treats not only the massive coil casing but also the winding pack conductor jacket under an essentially 3D stress state. The calculation tool features pre-optimization of the coil winding for graded layered winding layouts. The minimum space (radial built) required for the coil inboard portion that is a key design parameter is defined after possible winding pre-optimization. The procedure has been successfully benchmarked against numerical solutions and has been used for pre-dimensioning the toroidal coils in the frame of the current 2015 DEMO activity. [ABSTRACT FROM AUTHOR]

For the European DEMO fusion reactor, several options of the Toroidal Field (TF) coil design were proposed. The winding of the TF coil option 1 is based on React&Wind (RW), Nb 3 Sn, graded conductors in order to optimize the amount of superconductor according to the amplitude of the magnetic field in individual layers, as well as to limit the stress applied to the conductor using thicker jacket where the Lorentz forces are higher. In order to splice two different grades of Nb 3 Sn conductors, a joint is made to insure the adequate thermal and electrical contact between both conductors ends. For the good operation of the magnet, both electrical resistance and AC loss of the joint must be sufficiently low. A novel conductor splicing technique has been developed at Swiss Plasma Center (SPC) based on diffusion bonding and copper spray coating. The first joint prototype based on this technology was manufactured and tested successfully in 2019. The scopes of this paper are as follows: a) to present the design, the manufacture and the test of the second joint prototype based on diffusion bounding, b) to address the issue of joint resistance degradation during cyclic loading, experienced by the first prototype and c) to investigate on the reproducibility of the joint technique testing a second prototype. [ABSTRACT FROM AUTHOR]

This paper summarizes the work on the manufacturing and testing of conductor samples for ITER poloidal field 1&6 coils. The work was carried out by the IO and multiple institutions in the EU, Switzerland and Russia. Five samples were manufactured and tested in the SULTAN test facility. Analyses of dc performances, ac, and mechanical losses test results have been performed. All measurements were carried out before and after cyclic mechanical load created by the electromagnetic forces. DC test results of all samples tested demonstrated that the conductor performances are not affected by cycling mechanical load and meet the required value of current-sharing temperature $T_{{\rm{cs}}}\geq $ 5.8 K at magnetic field $B_{{{\max}}}\,= $ 6.4 T and operating current $I_{{\rm{op}}}\,= $ 48 kA. Measurements of the ac and mechanical losses were performed by gas flow calorimetric method. AC loss measurements were carried out at different operating currents and background magnetic fields. AC loss of tested samples showed no change under cycling mechanical load, but showed a correlation of ac loss versus background magnetic field. Mechanical loss, as had been expected, decreased significantly after first ten cycles then reaches a stabile value. The value of mechanical loss even in virgin condition does not affect workability of a magnet. [ABSTRACT FROM PUBLISHER]

Abstract: The High Temperature Superconductor (HTS) current leads (CL) for the Wendelstein 7-X stellarator (W7-X) with a maximum current of 18.2kA are designed and manufactured by the Karlsruhe Institute of Technology (KIT). In addition the acceptance tests of the W7-X HTS CLs are performed at KIT. Therefore the existing TOSKA facility has been extended by a test cryostat connected to the main vacuum vessel. After the extensive prototype CL test campaigns in 2010 the final acceptance tests of 14 series CLs started in 2011. The estimated completion of the routine test campaign is in December 2012. The main parts of each acceptance test are the determination of the heat load at the 4.5K level, of the necessary 50K He mass flow rate through the heat exchanger as well as the simulation of a loss of flow accident of the 50K He mass flow at full current (18.2kA). The tests also include a long-time operation at the maximum current of 18.2kA to demonstrate the steady state operation capability of the HTS CLs. In the present paper an overview of all conducted HTS CL acceptance tests is given. The results for the different CLs are summarized and compared to the specifications. [Copyright &y& Elsevier]

Abstract: The Karlsruhe Institute of Technology (KIT) is responsible for design, production and test of the High Temperature Superconductor (HTS) current leads for the stellarator Wendelstein 7-X (W7-X). In total 14 current leads with a maximum current of 18.2kA are required. Special feature is the upside-down orientation of the current leads because of the location of the power supplies in the basement of the experimental area of W7-X. One further important requirement is the Paschen tight electrical insulation of current leads and the connection to the bus bar system. Due to some very specific manufacturing steps, budget and time restrictions, it has been mutually decided between the project partners to manufacture most of the components in house, except the HTS stacks which have been produced and delivered by industry. As the semi-finished parts were manufactured in the central workshop of KIT, the assembly of the current leads was performed in the ITEP (Institute for Technical Physics). The final acceptance test of the current leads is performed at KIT, using a dedicated test cryostat assembled beside and connected to the main vacuum vessel of the TOSKA facility. The paper describes the status of the manufacturing of the current leads. In addition attention is given to specific problems that occurred during the manufacturing and testing. [Copyright &y& Elsevier]

Abstract: In case of a magnet quench, the primary quench detection system is classically based on voltage detection. In addition, a secondary quench detection is required and it could rely on signals of thermohydraulic nature. A model based on the coupled GANDALF [1] and FLOWER [2] codes has been developed in previous papers for the Central Solenoid [3,4], and for the Toroidal Field Coils [5,6] of ITER. This study focuses on Poloidal Field Coil PF5 as it has the highest magnetic field value along the conductor. Results of some parametric studies are presented and the influence of the initiation parameters of the quench (energy, location and length of heat deposition, time duration for detection). The possibility of a secondary thermohydraulic detection and the main impact on cryogenic system are analyzed and discussed. [Copyright &y& Elsevier]

Abstract: All large fusion experimental machines presently under construction like the tokamak ITER, the stellarator Wendelstein 7-X (W7-X), and the satellite tokamak JT-60SA will use high temperature superconductor (HTS) current leads to reduce the cryogenic load and operational cost. The Karlsruhe Institute of Technology (KIT which is a merger of former Forschungszentrum Karlsruhe and University Karlsruhe) designed, constructed and successfully tested a 68kA HTS current lead demonstrator for ITER which led to the decision by ITER IO to use HTS current leads for the ITER magnet system, and the Chinese Domestic Agency is in charge of providing these leads as a part of the superconducting feeder system. KIT is in charge of design, construction and test of HTS current leads for W7-X and JT-60SA. The paper discusses the status of the current lead development and describes the results of the W7-X prototype test which has been performed recently. [Copyright &y& Elsevier]

Abstract: An extensive acceptance test program for the Wendelstein 7-X (W7-X) confinement coils is presently being carried out in the CEA Saclay cryomagnetic test facility. Over half of the 50 non-planar coils and 20 planar coils have already been subject to a cool-down to liquid helium temperature, allowing current tests in the superconducting state. This paper presents a description of the cooling protocol observed at CEA. In-depth background information about the helium refrigeration technology limitations, the coil active cooling procedure and its control are given. The cryogenic power extraction is estimated through mass flow rate and enthalpy balance of the winding and casing helium circuits, which are derived from various coil and facility sensors. Coils geometry and material thermal properties are given, as well as a simple modelling of the coil cooling. The observed coil thermal behaviour can help to better understand the role of the casing cooling loop on the cooling inertia and on the total cooling time. Finally, the cooling down process is projected into the future operation of the W7-X stellarator under construction. More generally, data and experience gained from the cooling tests operated at the CEA Saclay cryomagnetic test facility provide practical knowledge to foresee the thermal behaviour and cryogenic challenges of other large magnets for thermonuclear fusion. [Copyright &y& Elsevier]

Abstract: The discovery of an upward counter flow of helium in the outer annulus of the vertically oriented and top-to-bottom cooled ITER PF-FSJS (Poloidal Field Coil-Full Size Joint Sample) in 2002 led to closer investigations of the effect because it may lead to a reduction of the operational margin of the superconductor used in the ITER environment. Recently, further thermo-hydraulic experiments were carried out on the TFAS2 sample (Toroidal Field Advanced Strand sample 2) with the intent to asses the effect in detail. First investigations confirmed the initial assumption that the origin of the effect lies in the buoyancy of the heated, and thus less dense, helium in the outer annulus of the cable. The helium there is in good contact with the superconducting strands heated by neutron irradiation, ac losses or heat influx, but is thermally and hydraulically less well coupled to the downward flowing helium in the central channel. This paper presents an analysis of the TFAS2 experiments using the simulation program THEA™, specifically extended with a term for gravitational forces acting on helium of varying density. With the experience gained, the simulation of the thermo-hydraulic behavior of the Toroidal Field Coil inner leg shows the operational limits and boundary conditions of this coil in ITER. [Copyright &y& Elsevier]

Abstract: This paper describes a new method to determine the equivalent transverse heat transfer coefficient between the helium flow in the cable bundle and the flow in the central space of a CICC’s with parallel cooling channel. The method is based on the analysis of the temperature traces during a heat pulse experiment. The equations for the average temperature distribution in the cable are solved analytically to obtain a closed-form expression of the characteristic rise time of the temperature front as a function of the transverse heat transfer coefficient. The value of the equivalent transverse heat transfer coefficient from the bundle to the hole is then obtained as the best fit of the experimental rise time. We show the results of the method by application to short cable sample experiments in SULTAN. [Copyright &y& Elsevier]

Various tests, either on full-size joint samples or on model coils confirmed that current distribution may play a crucial role in the electrical behaviour of CICC in operating conditions. In order to evaluate its influence, CEA developed a code (ENSIC) the main feature of which is a CICC electrical model including a discrete resistive network associated with superconducting lengths. Longitudinal and transverse resistances are also modeled, representing either joint or conductor. In our paper we will present the comparison of experimental results with ENSIC calculations for one International Thermonuclear Experimental Reactor (ITER) sample prototype relevant to poloidal field (PF) coils: the PF-full-size joint sample (PF-FSJS). In this purpose, the current distribution has been measured thanks to a segmented Rogowski coils system. Current distribution effects on the basic characteristics (TCS, n-value …) of the cable compared to single strand will be discussed. This study aims at putting light on the global strand state in a conductor and is also useful to evaluate some intrinsic parameters hardly measurable (effective interpetal transverse contact resistance for example) allowing further application in coils. [Copyright &y& Elsevier]

Two types of model coils (MC) were recently tested in order to assess possible designs of the international thermonuclear experimental reactor (ITER) conductor and Joints for the Toroidal field (TF) coils. One of these MC is wound using a stainless-steel-jacketed conductor, the Toroidal field model coil (TFMC), and the other a Ti-jacketed conductor, the Toroidal field coil insert (TFCI). This paper is dedicated to the electrical behaviour of these conductors through the V–T characteristics. In a first part we present the analysis of experimental results with a global thermo-electromagnetic model in which the current distribution is supposed uniform among the strands. Emphasis is put on all elements making the operating conditions of strands in a coil different from those of a single strand test (global strain, magnetic field and temperature gradients). The second part is dedicated to current distribution non-uniformity effects, either considered at the strand scale (TFMC) or at the subcable scale (TFCI). For this purpose, CEA developed a code (ENSIC) based on an electrical model network simulating conductor lengths and the associated joints. This model is applied to both coil tests. The last part includes a discussion on the validity and precision of the results provided by both models, compared to the experimental measurements. [Copyright &y& Elsevier]

The analysis of current distribution and re-distribution in superconducting cables requires the knowledge of the electric coupling among strands, and in particular the interstrand resistance and inductance values. In practice both parameters can have wide variations in cables commonly used such as Rutherford cables for accelerators or Cable-in-Conduits for fusion and SMES magnets. In this paper we describe a model of a multi-stage twisted cable with arbitrary geometry that can be used to study the range of interstrand resistances and inductances that is associated with variations of geometry. These variations can be due to cabling or compaction effects. To describe the variations from the nominal geometry we have adopted a cable model that resembles to the physical process of cabling and compaction. The inductance calculation part of the model is validated by comparison to semi-analytical results, showing excellent accuracy and execution speed. [Copyright &y& Elsevier]

The International Thermonuclear Experimental Reactor Toroidal Field Model Coil, a large (2.7 m × 3.8 m × 0.8 m) superconducting (Nb3Sn) DC coil designed and constructed in collaboration between EU industries and laboratories coordinated by European Fusion Development Agreement, has been tested during 2001 in the TOSKA cryogenic facility at Forschungszentrum Karlsruhe, Germany, achieving the nominal 80 kA at 7.8 T peak field and 86 MJ stored energy as a standalone coil (Phase I). Possibly the highlight of the tests was the measurement of the current sharing temperature (TCS) at different transport currents I in the self-field of the coil. The measurement method is discussed, based also on the two-year long predictive work, which preceded it. The results of the full set of TCS measurements at I=80, 69 and 57 kA are presented here, and evaluated in a companion paper (part 2). [Copyright &y& Elsevier]

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 amechanical 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. [ABSTRACT FROM AUTHOR]

The toroidal field (TF) system of JT-60SA tokamak comprises 18 NbTi superconducting coils. In each TF coil, 6 cablein-conduit conductor (CICC) lengths are wound into 6 doublepancakes and carry a nominal current of 25.7 kA at a temperature of 4.7 K. Each coil is tested in the cold test facility (CEA Saclay), up to quench. A SuperMagnet (CryoSoft) model has been developed, each of the 12 pancakes being modeled with THEA and cryogenic circuit being modeled with FLOWER. The experiments showed that helium inlet temperature increases until quench triggering at about 7.5 K on C11 and C13, with quench starting on a lateral and on a central pancake, respectively. Each test has been simulated, applying (or not) a realistic heat load from casing to winding pack that was estimated from experimental measurements. A parametric analysis has been performed, considering realistic or null heat flux deposition, variation of friction factor (in fabrication quality range), and CICC critical current density (in strand Jc performance range). This last parameter was found to have the largest impact on the localization of the first quenched pancake (central or lateral). [ABSTRACT FROM AUTHOR]

Strain limit for critical current degradation in various technical rapid-heating, quenching, and transformation-processed Nb3Al strands has been studied with respect to the matrix material and filament diameter. This property is one important factor as well as the strain sensitivity to Jc characteristics, especially when we consider to apply the so-called react and wind technique to large magnets such as DEMO reactors. We compared five strands, including a standard Nb matrix, standard Ta matrix, Nb/Ag/Nb three-layered barrier structure, and other two fine filament diameter strands prepared by the restackingmethod. In normal Nb and Ta matrix strands with a filament diameter of more than 30 μm, the irreversible strain limit is about 0.3%. Refinement of the filament diameter enables us to significantly improve the irreversible strain limit up to more than 0.7%. The Nb/Ag/Nb three-layered barrier structure appearsmore attractive from a practical point of view. The irreversible strain limit was improved up to nearly 0.6%, even if the filament diameter is more than 20 μm. The Ag layer between the Nb3 Al filaments seems to act as a cushion to mitigate the stress concentration in the wire. The Ag barrier also seems to be effective for current bypass in existence of cracks, resulting in suppressing Ic degradation in a higher strain range. [ABSTRACT FROM AUTHOR]

In summer 2015, a new reference baseline is issued for theDEMOEUROfusion tokamak. Thereafter, the toroidal field (TF) coils have been updated with the new layout of the react-andwind conductor proposed by the Swiss Plasma Center (SPC). Each TF coil consists of 12 single layers of graded Nb3Sn conductors connected in series by "invisible" inter-layer joints fully embedded in the winding pack. The high-grade Nb3Sn conductor operates at 63 kA, 12.2 T with Tcs > 6.5 K. The new prototype of the highgrade cable had been manufactured by ENEA at TRATOS, Italy, and delivered to SPC: 20 m of Nb3Sn and about 10 m of copper dummy conductor. The SULTAN conductor sample consisting of two identical, beforehand heat-treated conductors was prepared and tested at SPC in May-June 2017. Upon the completion of comprehensive conductor tests, the one of those conductors was used for manufacturing of the designed at SPC potentially applicable inter-layer TF joint. The conductor was cut, and then, joined again in accordance with the developed at SPC design of inter-layer joint. The SULTAN sample comprising the earlier tested non-destroyed conductor and the earlier tested, but then cut and joined conductor was re-assembled and tested in the SULTAN test facility at SPC in August 2017. [ABSTRACT FROM AUTHOR]

The React&Wind (RW) method for the large Nb3Sn magnets of theEUROfusionDEMOdevice was proposed in 2013 by the Swiss Plasma Center. A first short-length prototype conductor for the toroidal field (TF) coils, named RW1, was tested in the EDIPO test facility during 2015-2016 up to 82 kA at 13 T, with an effective strain matching the expected thermal strain of ≈ -0.30%. After the baseline of the DEMO device was updated in 2015, the new requirements led to an updated conductor design, for 63 kA at 12.2 T. Themanufacturing experience of the first prototype, named RW1, is exploited in a second short-length prototype conductor, named RW2, assembled and tested in 2017: The conductor aspect ratio is reduced, and the segregated copper wires are replaced by a solid block of mixedmatrix stabilizer. Although designed for theTF coils, with dc operation, the moderate ac loss of the flat cable makes the RW2 a good candidate also for the central solenoid conductor. The Nb3Sn strand for RW2 is supplied by WST (PRC), and the flat cable is made at TRATOS (Italy). The rationale of the design, the conductor manufacture, the sample assembly, and the test results in SULTAN are reported. [ABSTRACT FROM AUTHOR]

The ITER Magnet System includes a set of 18 superconducting correction coils (CC) which are used to compensate the error field modes arising from geometrical deviations caused by manufacturing and assembly tolerances. The turn and ground insulation are electrically insulated with a multi-layer fiberglass polyimide interleaved composite, impregnated with epoxy resin using vacuum pressure impregnation (VPI). Adequate high voltage insulation (5 kV), mechanical strength and rigidity of the winding pack should be achieved after impregnation and curing of the insulation system. VPI is an effective process to avoid defects such dry spots and incomplete wet out. This insulation technology has also been developed since several years for application to large superconducting coils and more recently to ITER CC. It allows the coils to be impregnated without impacting on their functional characteristics. One of the critical challenges associated with the construction of the CC is the qualification of the VPI insulation. Sections issued from representative VPI test samples with real scale side correction Coil (SCC) cross-section have been delivered and characterized at CERN. High resolution micro-optical inspections have been carried out on large areas through digital microscopy. The aim was to identify lack of impregnation, areas of pure resin and void entrapments. The areas near the filling fibre glass rope received special attention. High precision dimensional and geometrical assessments have been performed with the help of image analysis. Compression and pull-out tests have been also carried out. Finally, high-resolution 3D-computed tomography has been applied for a full volumetric inspection of the sections, enabling the reconstruction in three dimensions of the VPI samples and allowing to fully detect, confirm, and image the volume defects already identified by micro-optical observations. [ABSTRACT FROM PUBLISHER]

A focused Research and Development program on indigenously developing fusion relevant magnets is presently under way in India at the Institute for Plasma Research in association with Indian industries. A large size magnet winding facility under contractual obligations from Institute for Plasma Research is being developed by Patel's Analog and Digital Measurements Co. Pvt. Ltd. (PADMCO). This “special purpose winding machine” (SPWM) is capable of winding up to 30 mm \times 30 mm square cross section NbTi and Nb3Sn based cable-in-conduit-conductors (CICC) in both circular and non-circular (D shape) winding pack geometries. Magnet winding packs with both layered as well as double pancake configurations can be wound with this SPWM. The winding pack dimensions could be up to five meters in the major dimensions, with a width as high as several hundreds of mm. The Special Purpose Winding Machine (SPWM) facility primarily comprises of a feeding conductor spool with appropriate tensioning mechanisms, straightening and shaping units, cleaning unit, insulation unit, press forming units and tangent tracing units, rotating turn table mandrel, rotation and synchronization arrangements etc. with both hydraulic and electrical control elements and feedbacks. The facility is expected to carry out winding with a precision of 1 mm over a meter of built-up length in various arc lengths of windings. Prototype windings are planned using an indigenously developed 30 kA CICC with these facilities in the second half of 2011. [ABSTRACT FROM PUBLISHER]

A principal part of the ITER fusion reactor is the Toroidal Field magnet system which consists of 18 “D” shaped coils. Fusion for Energy, (F4E), the European Domestic Agency for ITER, is responsible for the procurement of 10 such coils. The completed coils, measuring approximately 14 m \times 9 m and weighing 300 tons, comprise an outer structural case into which a “winding pack”, itself made up of 7 conductor double pancake and radial plate assemblies, is inserted. [ABSTRACT FROM PUBLISHER]

The test of two Wendelstein 7-X (W7-X) prototype high temperature superconductor (HTS) current leads (CL) was extended to dynamic operation, in order to evaluate AC losses in parts of the CL. Trapezoidal and triangular current cycles up to a maximum current of 20 kA with different ramp rates between 0.5 kA/s and 10 kA/s were performed. Temperature changes at the warm end of the HTS-module and the influence on the heat load at the cold end of the CL were studied. Additionally, a DC operation that generates a power loss equivalent to the triangular pulsing run was investigated. AC losses in the low temperature part of the CL are estimated by simplified models and compared to the experimental results of the heat load measurements. The effect of the AC losses on the CL performance is discussed. [ABSTRACT FROM PUBLISHER]

All 21 Procurement Agreements (PAs) related to the ITER superconducting magnet system were signed in March 2011. Six Domestic Agencies (DAs) of the seven, namely China, EU, Japan, Russia, the United States and Korea have to deliver components of the magnet system like TF coils and Feeders. In order to organize the complicated international supply-chain, the ITER Manufacturing Database (MD) has been developed as a sub-project of the ITER Engineering Database (EDB) which aims to become the official Product Lifecycle Management (PLM) system of the ITER project. The first version of the PLM was released in August 2011. One of the advantages of the MD is to provide clear definitions of the work conditions and resultant deliverables to all stakeholders for a common understanding and better mutual communication. The work conditions (e.g. applicable code/standard) are basically agreed by all stakeholders after the preparation phase. They are then prepared in the MD as four customized User I/F (interface) screens: the interactive Parts List, the Manufacturing Inspection Plan (MIP), the Welding Quality Inspection and Production Plan (WQIPP) and the Tooling List. Users only upload the required reports (e.g. the mill certificate for each part, test reports) into the corresponding User I/F. Consequently, the filled up User I/F becomes a deliverable, the Manufacturing Dossier. [ABSTRACT FROM PUBLISHER]

When energized, the ITER Central Solenoid coils experience large pulsed electromagnetic forces that the conductor jacket itself must withstand. The conductor jacket consists of circle-in-square extruded and drawn austenitic stainless steel pipes. The qualification of the production of stainless steel jacket material was carried out on jackets manufactured in both a very low carbon AISI 316LN grade and a high Mn-bearing austenitic stainless steel, called JK2LB. Two different suppliers provided fully representative batch productions of both grades. Extensive metallurgical and dimensional metrology examinations were carried out at different steps of the processing, starting from the forged billets used as semifinished products to be engaged in the extrusion process, to the solution annealed jackets in their final shape. A specific method of Phased Array Ultrasonic Testing (PAUT) was developed and successfully applied for the non-destructive examination of the different jacket productions. PAUT sectorial scan inspections were carried out with probes traveling on the outer surface of the section, allowing almost 95% of the volume to be examined despite the complex geometry of the jacket. [ABSTRACT FROM PUBLISHER]