Toshinari Ando

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.

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.

Divertor experiment on particle and energy control in neutral beam heated JT-60 discharges

The divertor characteristics in particle and energy control in neutral beam (NB) heated discharges on JT-60 have been studied with injection powers of up to 20 MW. The essential divertor functions are achieved successfully. In ohmically heated discharges, the minimum clearances between the separatrix magnetic surface and the fixed limiter for sufficient divertor action are 1.5 cm for e = 1.5 × 1019 m−3 and 2.5 to 3 cm for e = 4 × 1019 m−3. Global power balance studies show that, in NB heated divertor discharges, about 5% to 10% of the total absorbed power, PABS, is radiated from the main plasma, while 50% to 60% is radiated in the limiter discharge. For e = 6 × 1019 m−3, 50% of PABS flows to the divertor plate. The radiation loss in the divertor chamber is 15% of PABS. According to the spatial distribution of the temperature rise on the divertor plate, the half-width of the heat load is less than 1 cm for e = 1.5 to 4.4 × 1019 m−3. The neutral pressures in the divertor chamber and around the main plasma increase in proportion to e2. The compression ratio is about 45. The effectiveness of the divertor pumping system in particle control is demonstrated for NB pulses of, at least, 1 s. Reduction of evaporation by separatrix swing is also shown.

Development of 10 kA Bi2212 conductor for fusion application

Recently, superconducting properties of high Tc superconductors (HTS) have been highly improved and using performance of HTS a conceptual design of a fusion device was performed by Japan Atomic Energy Research Institute (JAERI). HTS has a capability to produce a magnetic field of higher than 16 T, which is required in such a fusion power reactor. Aiming at development of the conductor, a trial fabrication of a 10-kA 12-T conductor was started using round Ag-alloy sheathed Bi-2212 strands, which has best performance at 4.2 K, 16 T at present. The conductor has about 34-mm diameter, and is composed of 729 HTS strands. Operating temperature is designed at not only 4 K but also 20 K. The cable of the conductor is solder-coated on the surface to use specific heat of the lead as much as possible, which at 20 K is almost comparable with specific heat of SHe at 4.5 K, 0.6 MPa. From the tests of the conductor, the fabrication of large HTS conductor and 10 kA operation at 12 T, 18 K /spl sim/ 20 K were successfully performed and the first step of developing work of HTS conductor for fusion application was achieved.

Present status of JT-60 upgrade

In the JT-60 tokamak, the original outer single-null plasma containment vacuum vessel and poloidal field coils, will be exchanged with those of large D-shaped cross sections. The basic dimensions of the new vessel will allow plasmas of 6-MA current, 70 to 100 m/sup 3/ volume, and vertical elongations of 1.4 to 1.7. Installation of a large vessel is made possible at the cost of high toroidal ripple volume in the space occupied by the plasma and at the cost of high electromagnetic force acting on the coils and the vessel. A prime requirement for the modification is maximizing the plasma performance with minimum expense, while ensuring access to the plasma for heating and diagnostics. The plasma configuration is designed to allow the largest D-shape discharge with a lower single null, within the boundary of the original toroidal field coils. The D-shaped vessel is designed according to a new concept of Inconel 625 all-welded continuous chamber in a double skin construction. Rectangular tubes are sandwiched by three-dimensionally curved skins to form a totally everywhere rounded vacuum vessel. The gas is being changed to deuterium from hydrogen, which will allow neutral beams of 40 MW. The lower hybrid and ion-cyclotron systems will be provided with larger antennas.<<ETX>>

Development of Nb3Al/Cu Multifilamentary Superconductors

The Nb3A1 superconductor has been expected to the application under the high field, such as the one for International Thermo-nuclear Experimental Reactor (ITER), on the point of superior characteristics of Ic under stress. The present status of the development of Nb3A1 superconducting wires manufactured by the Jelly-Roll process is presented in this paper. The recent development has achieved the Nb3A1 superconductor stabilized by the copper matrix with a high critical current density and a low hysteresis loss. In the development, the microstructure of Nb-Al compound phases, each identified by EDS, are observed. The volume of each Nb-Al phase calculated by the Image Analysis System is discussed in terms of its dependence of Jc. With a non-copper critical current density of more than 550 A/mm2 at 12T, the wire features excellent high field characteristics, for example μ0Hc2 is 21.5T. Also, this wire has less degradation of Ic by strain compared to Nb3Sn. Hysteresis losses in the wire was measured by the magnetization method. For the field perpendicular to the wire length, it was shown that the effective filament diameter is almost equal to the actual filament diameter (21μm). On the other hand, the hysteresis loss for the parallel field was around one-fourth as large as that for the perpendicular one, corresponding to an effective filament diameter of approximately 8 µm. The 40kA-class cable-in-conduit conductor was fabricated to demonstrate its applicability to fusion magnets. The critical current test of the conductor proves that the capacity of 40kA is attained at 11.2T. The practical 10kA-–100m class long length cable-in-conduit conductor, of which conduit is Titanium, was fabricated first in the world.

Test results of high temperature superconductor current lead at 14.5 kA operation

High temperature superconductor (HTS) current leads have been developed for the International Thermonuclear Experimental Reactor (ITER) magnet system, which are required not only to reduce the lead heat leak but also to maintain safety in a fault condition. A pair of 10-kA class HTS current leads was fabricated and tested. The lead consists of a copper part and an HTS part. The HTS part is composed of 192 Bi-2223 silver-alloy sheathed tapes in a cylindrical array on a stainless steel tube. Thermal performance and stability were tested. The current leads could carry up to 14.5 kA by placing magnetic materials between the HTS elements, which were installed to reduce the perpendicular magnetic field in the HTS elements.

Design of a 60-kA HTS current lead for fusion magnets and its R&D

A 60 kA HTS current lead has been designed for large fusion magnets such as the ITER magnet. The actual refrigeration input power required to cool the current lead is specified to be reduced to one third that of the conventional copper lead. The HTS part of the 60 kA lead consists of 48 units installed with cylindrical array into the outer surface of a stainless steel tube with a diameter of 146 mm. Each unit is composed of six Bi2223/Ag-10at%Au tapes, and its cross-sectional dimension is 6.5 mm/spl times/2.7 mm. The HTS part is cooled by conduction, and the warm and cold end temperature conditions of the HTS part are 50 K and 4.5 K, respectively. The copper part is cooled by helium gas, a flow rate of 3.9 g/s and the inlet temperature of 35 K. The 60-kA lead has been designed in consideration of safety under the long discharge time condition of ITER-TF coil with a detection time of 2 sec, and a discharge time constant of 15 sec. For the purpose of verifying the reliability of the design for the long discharge time, one unit sample has been fabricated and tested. The result indicates that the maximum temperature rise of the HTS part is less than 150 K for the ITER like-discharge from 1.25 kA corresponding to 60 kA of the full lead with 48 units.

Development of the prototype conductors and design of the test coil for the Fusion Experimental Reactor

For the Prototoroidal Coil project, a prototoroidal coil will be fabricated and tested electromagnetically and mechanically under conditions similar to the operating conditions of the toroidal coils of the Fusion Experimental Reactor (FER). Three types of 30-kA, 12-T, force-cooled conductors have been developed as the prototype conductors of the prototoroidal coil. Trial fabrication of these conductors and evaluation testing of their elements are under way. Design of the prototoroidal coil and its background field coil system, which consists of two LCT (Large Coil Task) coils and two LCT backup coils, is also in progress.<<ETX>>