T. Terakado

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.

First test results for the ITER central solenoid model coil

Abstract The largest pulsed superconducting coils ever built, the Central Solenoid (CS) Model Coil and Central Solenoid Insert Coil were successfully developed and tested by international collaboration under the R&D activity of the International Thermonuclear Experimental Reactor (ITER), demonstrating and validating the engineering design criteria of the ITER Central Solenoid coil. The typical achievement is to charge the coil up to the operation current of 46 kA, and the maximum magnetic field to 13 T with a swift rump rate of 0.6 T/s without quench. The typical stored energy of the coil reached during the tests was 640 MJ that is 21 times larger than any other superconducting pulsed coils ever built. The test have shown that the high current cable in conduit conductor technology is indeed applicable to the ITER coils and could accomplish all the requirements of current sharing temperature, AC losses, ramp rate limitation, quench behavior and 10 000-cycle operation.

Development of a vacuum switch carrying a continuous current of 36 kA DC

This paper describes the development of a vacuum switch carrying a continuous current of 36 kA DC. This switch consists of three vacuum interrupters connected in parallel. Generally, it is required to reduce the resistive loss and to increase the heat removal capability for increasing the current carrying capability of VCB. Then, maximizing the cross section of the conductor, and shortening the current path are principally important. However, a coil structure which produces an axial magnetic field to extinguish the arc stably, makes it difficult due to the geometric complexity. Then, the authors adopted unique design of the coil structure to solve this difficulty. The newly developed vacuum interrupter is possible to carry a current of 8 kA without any cooling, which is twice that of the largest VCB available at present in the world. Moreover, introducing the forced-air cooling enhanced the performance up to 12 kA by improving the cooling efficiency. They are verified through a heat running test.

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.

Measurement of high frequency impedance of ITER CS Model Coil and the simulation analysis of internal resonance phenomenon

This paper describes the internal resonance phenomenon of a large scale superconducting coil. The frequency dependence of electric impedance defined at the coil terminal and the voltage distribution of the layer winding were measured at the ITER CS model coil under the superconducting state in advance of the pulse operation test. As a result, large change of the impedance and unbalance of the layer voltage were observed. The AC circuit analysis was carried out in order to evaluate the measurement results and to calculate the voltage of layer insulator. It was concluded that the higher voltage, which may appear on the layer insulation, is about a half of the coil terminal voltage in high frequency.

Design and analysis of plasma position and shape control in superconducting tokamak JT-60SC

Abstract The analyses of the plasma position and shape control in the superconducting tokamak JT-60SC in JAERI are presented. The vacuum vessel and stabilizing plates located closely to the plasma are modeled in 3 dimension, and we can take into account the large ports in the vacuum vessel. The linear numerical model used in the design for the plasma feedback control system is based on Grad–Shafranov equation, which allows the plasma surface deformation. For a slower control of the plasma shape, the superconducting equilibrium field (EF) coils outside toroidal field coils are used, while for a fast control of the plasma position, in-vessel normal conducting coils (IV coil) are used. It is shown that the available loop voltages of the EF and IV coils are very limited, but there are sufficient accuracy and acceptable response time of plasma position and shape control.

Development and Operational Experiences of the JT-60U Tokamak and Power Supplies

The design of the JT-60U tokamak, the configuration of the coil power supplies, and the operational experiences gained to date are reviewed. JT-60U is a large tokamak upgraded from the original JT-60 in order to obtain high plasma current, large plasma volume, and highly elongated divertor configurations. All components inside the toroidal magnetic field coils, such as vacuum vessel, poloidal magnetic field coils, divertor, etc., were modified. Various technologies and ideas were introduced to develop these components; for example, a multi-arc double skin wall structure for the vacuum vessel and a functional poloidal magnetic field coil system with taps for obtaining various plasma configurations. Furthermore, boron-carbide coated carbon fiber composite (CFC) tiles were used as divertor tiles to reduce erosion of carbon-base tiles. Later, a semiclosed divertor with pumps, for which cryo-panels originally used for NBI units were converted, was installed in the replacement of the open divertor. These development and operational results provide data for future tokamaks. Major failures experienced in the long operational period of JT-60U, such as water leakage from the toroidal magnetic field coil, fracture of carbon tiles, and breakdown of a filter capacitor, are described. As a maintenance issue for tokamaks using deuterium fueling gas, a method for reducing radiation exposure of in-vessel workers is described.

Preparations for pulse operation tests of the ITER CS model coil using the JT-60 power supply

Some preparations for pulse operation tests of the ITER CS model coil in the JT-60 power supply system have been performed. Since the ITER CS model coil takes a longer current ramp-up time, the pulse duration is extended up to 70 s from 15 s by decreasing the output voltage of the motor-generator. In the protection system a new grounding resistor with higher resistance has been installed and a new protection make-switch has also been developed. An analysis for the local resonances in the coil has been performed, and measurements of the coil impedance is planned to confirm the validity of the analysis results.

Modification of the JT-60 poloidal field power supply for high triangularity divertor operation

This paper describes the modification of the poloidal field power supply for the high triangularity divertor operation in JT-60. The power supply was mainly modified so as to change the combination of the thyristor converters and the poloidal field coils. After this modification. Stable divertor plasmas with a triangularity up to 0.4 at a plasma current of 2 MA were successfully obtained. However, an overcurrent of the thyristor converters in the horizontal field power supply had been often observed due to the oscillation of the voltage control command and natural commutation property of the thyristor converters in conjunction with MHD instabilities. This serious problem has been solved by introducing the rate limit control for the output voltage of the converters. In addition, it was demonstrated that both horizontal and vertical positions of the divertor X-point can be controlled simultaneously, which may be useful for the W-shaped semi-closed divertor operation.

VME and network application to magnet power supply control in JT-60U

The original control system for the poloidal field coil power supply (PFPS) in the JT-60 tokamak was developed in 1984. It was composed of CAMAC-based 16-bit microcomputers and I/Os. Corresponding to the progress in physics experiments, it had been required to improve the control system as well as to modify the power supply. Then, the control system was rejuvenated where advanced technologies of VME, networks and UNIX-workstations were applied. In addition to the upgrade of the system performance, the renewal of the control system has remarkably improved its hardware maintainability and software development environment.