Nobuo Tachikawa

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

Experimental and analytical studies on residual stress in the tungsten-copper duplex structure for a divertor application

Abstract Residual stresses that developed during cooling of the tungsten-copper duplex structure were measured by the strain gauge method and compared with those by thermoelastic-plastic analyses. Good agreement was obtained for both residual stress and displacement, even when the creep effect of the copper heat sink was neglected in the analyses. The residual stress on the tungsten top surface decreased with increase in the ratio of copper thickness ( t c ) to tungsten diameter ( D ). The effect of t c / D on the residual stress was large in the range of t c / D t c / D >1. The effective thickness of the plastic region in the copper heat sink was reduced in the same manner as the residual stress. The copper heat sink plastic developed first from the bonding interface and then from the center part of the bottom surface. The calculated edge stresses on the tungsten side surface were quite sensitive to the finite element mesh size near the interface edge, while stress on the tungsten top surface did not depend so much on the mesh size.

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.

Effects of interface edge configuration on residual stress in the bonded structures for a divertor application

Abstract Residual stresses in the interface region, that developed at the cool down during the brazing, were evaluated for several bonded structures to assess the mechanical strength of the bonded interface, using thermoelasto-plastic stress analysis. Normal stress components of the residual stresses around the interface edge of graphite–copper (C–Cu) bonded structures were compared for three types of bonded features such as flat-type, monoblock-type and saddle-type. The saddle-type structure was found to be favorable for its relatively low residual stress, easy fabrication accuracy on bonded interface and armor replacement. Residual stresses around the interface edge in three armor materials/copper bonded structures for a divertor plate were also examined for the C–Cu, tungsten–copper (W–Cu) and molybdenum alloy-copper (TZM–Cu), varying the interface wedge angle from 45° to 135°. An optimal bonded configuration for the least value of residual stress was found to have a wedge angle of 45° for the C–Cu, and 135° for both the W–Cu and TZM–Cu bonded ones.

Optimization studies on interfacial mechanical strength in the graphite-copper bonded structure for a divertor application

Residual stresses in the interface region were evaluated for the graphite-copper bonded system to assess the mechanical strength of the bonding interface. The normal stress components of the residual stresses around the interface edge were compared for three types of bonded structures such as a net-type, monoblock-type and saddle-type ones and were calculated to be 44 MPa, 7 MPa and 13 MPa, respectively. Consequently, the saddle-type structure was found to be favorable in the views of its mechanical integrity, fabrication ease and maintenance. The residual stresses around the interface edge in the saddle-type structures with the wedge angles of 45/spl deg/ to 135/spl deg/ were also examined. As the results, an optimal bonded configuration of the graphite-copper saddle-type structure was found to have wedge angle of about 60/spl deg/ for the least value of residual stress.

Neutron and plasma irradiations of fusion reactor materials using fusion plasma neutron sources

Abstract A compact and steady-state DT plasma with parameters relevant to a fusion-reactor, the FEF (fusion engineering facility), has been considered as a neutron source for the development of fusion reactor materials providing both. (1) 14 MeV fusion neutron irradiation testing, and (2) high charged particle flux for testing of plasma-facing materials during neutron irradiation. The 14 MeV neutron wall loading is about 1–2 MW/m 2 , with a test section area of 1 m 2 . One version of the test section is near the edge of the plasma in the vacuum chamber. Another is for high heat flux testing of plasma facing materials, where the charged particle flux leads to heat loads of several 10 kW/cm 2 . New versions of FEF, with localized high neutron fluxes, are also proposed by introducing a sloshing ion distribution into the plasma.

Fabrication and test of hydraulic jacks for the application to divertor support system of Fusion Experimental Reactor

Two different types of hydraulic jacks, consisting of hydraulic piston and bellows were developed for an application to locking and lifting systems for divertor segments for the Fusion Experimental Reactor. For the lifting system where a minimum stroke of 170 mm was required, a telescopic-type hydraulic jack with a length of 300 mm was fabricated and tested under an operation pressure of 1 MPa. For the locking system, a hydraulic jack with a conical cotter was developed that worked at 6.5 MPa under a radial offset of 2 mm. These jacks satisfied their specifications without any leakage, failure, or abrasion. The performance of several hard facing materials for the cotter, including alumina and metal alloys, was also examined. The Colmonoy No. 5 or Stellite alloy No. 6 was found to be excellent as a hard facing material for the cotter.<<ETX>>

Mechanical characteristics and seismic response of divertor support structure with locking system

Abstract A full-scale divertor support structure with hydraulic jacks was fabricated, and mechanical and seismic loading tests were carried out. This mock-up structure has four hydraulic jacks for locking the divertor plate and three jacks for lifting. The weight of the divertor plate model is 1.2 tons, similar to practical plate. The mechanical test results show that the divertor locking system provides high positioning tolerance within 0.3 mm and 0.2 mm in the toroidal and vertical directions respectively. To confirm the seismic response of the hydraulic cotter, a dynamic loading test was carried out. The fundamental frequency of 27 Hz is the mode of toroidal direction vibration, and is relatively high compared with the predominant earthquake frequencies. No damage of the hydraulic jacks appears under seismic loads. According to the results of mechanical and dynamic loading tests, the feasibility of the divertor support structure is verified.

Engineering aspects of the fusion engineering test facility based on mirror confinement (FEF)

An engineering feasibility study on the mirror-based fusion plasma neutron source as a fusion engineering test facility (FEF) has been carried out. The engineering issues and conditions to meet the requirement are presented. The radius and length of the plasma are about 10 cm and about 4 m, respectively. The wall loading of 14 MeV neutrons is about 1–2 MW/m2, and the required electric power is about 100–150 MW. The critical issues are as follows: (i) the heat lead to the first wall of the central cell and plug cell, which depends strongly on charge exchange neutrals. Methods to reduce the charge exchange in the plug cell are discussed. (ii) The blistering on the Faraday shield and the guard limiter of the ICRF antenna due to the bombardment of the alpha particles is the second critical issue. Some optimization has been considered.