K. Nakamoto

Stability of cable-in-conduit superconductors for Large Helical Device

The stability of cable-in-conduit superconductors has been experimentally investigated as part of a poloidal field coil program for the Large Helical Device (LHD) project. A new conductor was designed and fabricated, focusing on the stability. As a result of a zero-dimensional stability analysis, it was found that the conductor had a high stability, 5*10/sup 5/ J/m/sup 3/, at the design condition of 20.8 kA and 6.5 T. Current transfer performance after partial quenching has been investigated by using a short sample of the conductor for the poloidal field coil. The effects of the current transfer among the strands on the conductor stability are discussed.<<ETX>>

Experimental results of the R&D forced-flow poloidal coil (TOKI-PF)

Abstract As research and development (R&D) of poloidal-field coils for the Large Helical Device (LHD), a forced-flow cooled cable-in-conduit-type NbTi superconducting coil (TOKI-PF) has been tested. The success of excitations assured us that the NbTi cable-in-conduit conductor can be adopted for the LHD poloidal coils. During cool-down, mainly the hydraulic characteristics were measured. The friction factor could be expressed by using an empirical formula. In the DC operations, training behavior was observed, like in a pool-cooled coil. The friction factor was also affected by the number of excitations, which may be related to strand movement. The stability margin and propagation velocity were also measured using an inductive heater. It became evident that the stability margin had a lower value when the operation current was higher than 15 kA. This current seems to correspond to the limiting current.

Development of the Magnetically Floating Superconducting Dipole in the RT-1 Plasma Device

A high-temperature superconducting dipole floating in the magnetic field of a normal conducting coil has been developed for the Ring Trap (RT)-1 plasma confinement apparatus at the University of Tokyo. The RT-1 device consists of the floating superconducting dipole, a levitation coil made of normal conductor, a vacuum vessel for plasma confinement and detachable services for the floating dipole. This paper describes the design concept and operating scenario of the floating dipole as well as the required services such as a HTS persistent current switch necessary to realize this scenario

Excitation test results on a single inner vertical coil for the Large Helical Device

Excitation experiments on a single inner vertical coil for the Large Helical Device (LHD) were carried out to confirm its performance. The coil is one of the LHDs poloidal field coils and consists of a forced-flow Nb-Ti cable-in-conduit conductor (CICC). After cooldown for 250 hours, the superconducting transition of the whole coil was confirmed. Pressure drops were measured during the cooldown to determine the coils hydraulic characteristics. Then, the coil was successfully energized up to the specified current, 20.8 kA. In the experiments, heat generation of joints, radial displacement and acoustic emission (AE) were measured.

Design and fabrication of forced-flow coils as an R&D program for Large Helical Device

Two forced-flow cooled NbTi superconducting coils (TOKI-TF, PF) have been designed and fabricated. The helical coil (TOKI-TF) is a 1/4-scale model of the Large Helical Device (LHD). It has a major radius of 0.9 m, a minor radius of 0.25 m, and a pitch number of 4. Nominal current and maximum field were designed to be 8 kA and 2.8 T, respectively. Another coil (TOKI-PF) was fabricated for the demonstration of LHD poloidal field coils. It consists of two double pancakes with an inner radius of 0.6 m and an outer radius of 0.82 m. The nominal current of 25 kA simulates that of LHD poloidal field coils. Cable-in-conduit-type conductors were used for the both coils. The test facility was also constructed with a vacuum vessel, a liquid nitrogen shield, 30-kA power leads, a heat exchanger, and cryogenic supports. Design concepts and details are presented.

First Experiment on Levitation and Plasma With HTS Magnet in the RT-1 Plasma Device

The high temperature superconducting (HTS) floating magnet of the ring trap 1 (RT-1) reached the first experiment on levitation and plasma. The magnet using an HTS coil was levitated stably by levitation coil, and plasma was produced around the ring-shaped HTS magnet by electron cyclotron heating with 8.2 GHz microwave. This novel plasma device was constructed at the University of Tokyo to explore means of achieving the advanced-fuel fusion. The plasma confinement mechanism is based on the concept of high-beta relaxed state that is self-organized within flowing plasma. The HTS magnet is operated in a persistent-current mode and magnetically levitated in a plasma vacuum chamber. The weight of the HTS magnet is about 110 kg. Initially the HTS coil is cooled below 20 K by an external cooling system with detachable transfer tubes. After the transfer tubes are detached, an experiment of levitation and plasma is conducted while the HTS coil temperature remains within the range of 20 K-32 K without cooling. This paper describes the HTS coil design and test results of the HTS magnet as follows; an initial cooling, a persistent-current operation without cooling and the first levitation and the first plasma experiment.

Development of a superconducting joint technique between CIC conductors for poloidal coil of large helical device (LHD)

A solid state bonding technique been developed for the large superconducting joint of LHD poloidal coils. Low electrical resistance, high superconducting stability and compactness are required for the joints between pancake coils. All NbTi strands of cable-in-conduit (CIC) conductors were jointed directly and at one time by the proposed technique. From the critical current test with the real scale joint model, it is shown that quench current of this joint is about 1/3 times of the conductor critical current. This paper describes the main points in manufacturing the joint and evaluation of the quench current.<<ETX>>

Fabrication of the R&D forced-flow poloidal coil (TOKI-PF)

Abstract A forced-flow NbTi superconducting poloidal coil (TOKI-PF) using a cable-in-conduit conductor (CICC) was fabricated as an R&D project for the LHD coils. The main objective of TOKI-PF is to gather engineering data for the superconducting poloidal field coil. The major parameters are as follows: inner and outer radii 0.6 and 0.82 m, respectively, operating current 25.6 kA, stored energy 1.04 MJ, conductor NbTi CICC sized 10.0 × 22.5 mm. The mechanical coil configuration and nominal current of 25.6 kA simulate those of the LHD poloidal coils. Useful technical data were obtained and the feasibility of fabricating the LHD poloidal coils was confirmed. In this paper, preliminary test results will also be described.

Measurements of coupling losses in superconducting cable-in-conduit conductors affected by internal transverse electromagnetic-forces

In order to elucidate the dependence of the internal electromagnetic-forces on the interstrand coupling loss in superconducting cable-in-conduit conductors, we develop a new measuring system. This system enables us to measure the coupling loss of large scale conductors with transport current of 20 kA and the transverse magnetic field of 3.22 T bias in the wide frequency range of the superposed ripple field of 0.1-300 Hz. The data of the preliminary measurements show that the change of the coupling loss is only a little in the low frequency range to the extent of the electromagnetic-force of 0.44 MPa in the case of IS-conductor of the poloidal field coil for LHD.

Development of forced-cooled superconducting coil for Large Helical Device

A force-cooled superconducting coil has been studied for the Large Helical Device (LHD), which will be a main experimental apparatus of the Japanese National Institute for Fusion Science (NIFS). To achieve the main requirements for the LHD, a large-current, high-current-density NbTi cable-in-conduit-type conductor of 10 kA and 47 A/mm/sup 2/ at 8 T was developed. The test coil was wound in the form of a one-layer solenoid and tested at an operating current of 10 kA in a superposed field of about 7 T, at a hydraulic condition of 1.0 MPa, 4.5 K, and 10 g/s supercritical helium. The stability, normal propagation velocity, and pressure rise during a quench for the high-current-density case were also studied.