Kunihiro Matsui

Progress of the ITER central solenoid model coil programme

The worlds largest pulsed superconducting coil was successfully tested by charging up to 13 T and 46 kA with a stored energy of 640 MJ. The ITER central solenoid (CS) model coil and CS insert coil were developed and fabricated through an international collaboration, and their cooldown and charging tests were successfully carried out by international test and operation teams. In pulsed charging tests, where the original goal was 0.4 T/s up to 13 T, the CS model coil and the CS insert coil achieved ramp rates to 13 T of 0.6 T/s and 1.2 T/s, respectively. In addition, the CS insert coil was charged and discharged 10 003 times in the 13 T background field of the CS model coil and no degradation of the operational temperature margin directly coming from this cyclic operation was observed. These test results fulfilled all the goals of CS model coil development by confirming the validity of the engineering design and demonstrating that the ITER coils can now be constructed with confidence.

Development of Nb3Al superconductors for International Thermonuclear Experimental Reactor (ITER)

Abstract A fabrication technique of a Nb3Al strand using mass-production billets was developed for the framework of International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). The weight of a multifilament billet is about 70 kg, which corresponds to the total length of the wire of about 16 000 m. The critical current density is over 600 A mm−2 at 4.2 K, 12 T, and the hysteresis loss for the field parallel to the wire axis is lower than 600 mJ cm−3 at ±3 T. The residual resistivity of the chrome-plated wire is drastically improved to be lower than 1.6×10−10 Ωm with setting the large copper area at the centre, which reduces the chrome diffused into the copper. The yield of the superconducting strand has been improved, resulting in a yield of 70% for wire longer than 1500 m. The total volume of fabricated Nb3Al superconducting strand for ITER Nb3Al insert coil, which is fabricated in the ITER-Central Solenoid (CS) model coil program, is about 210 000 m, i.e. 1000 kg. This is the first experience of mass-production for Nb3Al superconductors in the world.

Critical current test results of 13 T–46 kA Nb3Al cable-in-conduit conductor

In the framework of ITER-EDA, a 13 T-46 kA Nb3Al conductor with stainless steel jacket has been developed in order to demonstrate applicability of an Nb3Al conductor with react-and-wind technique to ITER-TF coils. Using a 3.5 m sample consisting of a pair of conductors with 0% and 0.4% bending strain, the critical current performances of the Nb3Al conductors were studied to verify that the conductor achieves the expected performance and the bending strain of 0.4% does not originate degradation. The critical currents were measured at background magnetic fields of 7, 9, 10 and 11 T at temperatures from 6 to 9 K. The expected critical currents were evaluated taking into account the variation of the strain in the cross-section due to the bending strain as well as self-field and non-uniform current distribution as results of an imbalance in the joint resistance and inductances. The calculation results indicate that the current distribution is almost uniform and the experimental results showed good agreement with the expected critical currents. Accordingly, we can conclude that the fabrication process of this conductor is appropriate and the react-and-wind technique using the Nb3Al conductor is applicable to ITER-TF coils. In addition, the critical current of the Nb3Al conductor is expected to be 108 kA at 13 T and 4.5 K, resulting in a sufficient margin against the nominal current of 46 kA. Furthermore, it was found that the decrease in the critical current by thermal strain can be made small by applying the bending strain to the conductor so as to reduce the compressive strain at higher fields, i.e. inner side of the coil, in the conductor cross-section

AC loss measurement of 46 kA-13T Nb/sub 3/Sn conductor for ITER

AC losses of Nb/sub 3/Sn conductor samples with various void fractions for the ITER Central Solenoid Model Coil (CSMC) were measured by using calorimetric and magnetization techniques. The CSMC is designed to generate the magnetic field of 13 T at the operating current of 46 kA. The conductor consists of the multi-stage cable, having 1152 Nb/sub 3/Sn strands, and Incoloy 908 square jacket with circular hole. The strands are coated by chrome plating with 2 /spl mu/m layer. The last sub-cables are wrapped with Inconel tape, having high electric resistivity, to reduce the coupling current loss. The optimum void fraction for pulse coils is obtained from the relation between the coupling time constant and the void fraction. It is indicated that the sub-cable wrapping is very effective in limiting the coupling current between the sub-cables, as expected. The AC losses of the CS Insert were measured in various operating modes. From these obtained results, the validity of conductor design is demonstrated.

Performance of Japanese Conductors for ITER Toroidal Field Coils

The cable-in-conduit conductors for the ITER TF coils are fabricated using the latest high performance strands. The strands made by bronze and internal-tin methods are used for the conductors with a void fraction of 29% and 33%, respectively. Superconducting performance of the conductors was measured at the operating condition of the TF coils. The measured current sharing temperatures Tcs are 6.3-6.6 K for the bronze and 5.6-6.1 K for the internal-tin. The Tcs of the conductor with void fraction of 29% is 0.1-0.3 K higher than the conductor with a void fraction of 33%. It is shown from the results that the strain on the cable is between 0.7% and 0.75% and the n-values are between 4 and 6, much smaller than the n-values of strands.

Advanced fusion technologies developed for JT-60 superconducting tokamak

Modification of JT-60 as a full superconducting tokamak (JT-60SC) is planned. The objectives of the JT-60SC programme are to establish scientific and technological bases for steady-state operation of high performance plasmas and utilization of reduced-activation materials in an economically and environmentally attractive DEMO reactor. Advanced fusion technologies relevant to the DEMO reactor have been developed for the superconducting magnet technology and plasma facing components of the JT-60SC design. To achieve a high current density in a superconducting strand, Nb3Al strands with a high copper ratio of 4 have been newly developed for the toroidal field coils (TFCs) of JT-60SC. The R&D to demonstrate the applicability of the Nb3Al conductor to TFCs by a react-and-wind technique has been carried out using a full-size Nb3Al conductor. A full-size NbTi conductor with low ac loss using Ni-coated strands has been successfully developed. A forced cooling divertor component with high heat transfer using screw tubes has been developed for the first time. The heat removal performance of the carbon fibre composite target was successfully demonstrated on an electron beam irradiation stand.

Evaluation of strain applied to strands in a 13 T–46 kA Nb3Al cable-in-conduit conductor

A 13 T–46 kA Nb3Al coil, a Nb3Al insert, has been developed to demonstrate the applicability of the react-and-wind technique to toroidal field (TF) coils of a fusion reactor, such as ITER. It is estimated that a conductor is subjected to ≈0.4% bending strain after heat treatment when the react-and-wind method is applied to the TF coil fabrication. Therefore, 0.4% bending strain was artificially applied to the Nb3Al insert conductor after the heat treatment. Since a stainless steel conduit was used in this conductor, the strands are subjected to the compressive strains due to thermal stress. The effective strains applied to the strands are estimated by comparing the critical current test results of the Nb3Al insert at 10–13 T and calculation. Those due to thermal stress and conductor bending were evaluated to be ≈−0.4% and ≈0%, respectively. The effective strain by the conductor bending is sufficiently small so as not to affect the critical current performance of the Nb3Al conductor. In addition, the evaluated strain of the Nb3Al conductor is compared with those of same scale Nb3Sn conductors, in which there is an unexpected strain, which is proportional to the electromagnetic force. Such strain was not observed in the Nb3Al conductor. One of the explanations is higher rigidity of the Nb3Al strand than the Nb3Sn one. This shows that a Nb3Al conductor is suitable for application to large magnets, such as the TF coil of the fusion machine, which experiences large electromagnetic force in the conductor.

Fabrication of ITER central solenoid model coil-outer module

The central solenoid (CS) model coil-outer module being fabricated to demonstrate the justification of the CS design for the ITER, was almost completed except for epoxy impregnation to concrete whole layers. All the wound and heat treated layers have been assembled symmetrically with the insulation on the same axis, and for layer-to-layer joints the newly developed butt joint, has been installed.

Fabrication and Test of a 60‐kA HTS Current Lead for Fusion Magnet System

A 60‐kA HTS current lead was fabricated and tested in the frame of the R&D work for large fusion magnets such as the ITER magnet. The design of the current lead was characterized by its safe thermal protection for the current decay of about ten seconds after quench as well as lower electrical power consumption for its cryogenic equipment. The 60‐kA current lead is composed of a lower temperature HTS part using HTS and a high temperature copper part using a conventional copper cable. The HTS part consists of 48 HTS units installed in the cylindrical array into the grooves provided on the outer surface of a stainless steel cylinder with a diameter of 146 mm. The unit was composed of six Bi2223/Ag‐10at%Au tapes and its cross‐sectional dimension is 6.5 mm × 2.7 mm. The stainless steel cylinder and Ag alloy sheath have a share in thermal protection for quench. The current lead was tested under the cooling condition that the bottom of the HTS part was immersed into 4.2 K liquid helium, and the bottom of the cop...

Development of advanced superconducting coil technologies for the National Centralized Tokamak

Advanced technologies for fabrication of superconducting coils have been developed for the National Centralized Tokamak which is based on modification of JT-60. One of the technologies developed is the application of the react-and-wind (R&W) method of fabrication of a Nb3Al D-shaped coil. The bending strain of 0.4% due to the R&W method did not affect the critical current characteristics. This finding indicates the possibilities that the manufacturing cost of large size coils can be reduced further by downsizing the heat treatment furnace, and large complicated shape coils can be manufactured by using the Nb3Al conductor. Another technology is an advanced winding technique for the reduction of the ac losses of Nb3Sn coils by loading bending strain on the conductor. It was found that 0.2% bending strain is enough to reduce the ac losses to one-fifth at the virgin state. The newly developed NbTi conductor attained both (i) low ac loss of 116 ms in coupling time constant and (ii) low cost owing to the stainless steel wrap of the sub-cables and Ni plated NbTi strands with 11 µm filaments.