Design of a toroidal plasma confinement device with a levitated super-conducting internal coil
- High Temperature Plasma Center, University of Tokyo, Bunkyo-ku, Tokyo 113-8656 (Japan)
- School of Frontier Science, University of Tokyo, Bunkyo-ku, Tokyo 113-8656 (Japan)
- National Institute for Fusion Science, Toki, Gifu, 509-5292 (Japan)
A toroidal device has been constructed and nonneutral plasma experiments have been intensively promoted, where an internal ring coil with a copper conductor has been employed. We are now designing a toroidal plasma trapping device with a levitated superconducting internal coil, so as to avoid plasma loss through current-lead and support structures of the internal coil. Typical machine parameters are as follows; the major radius of the internal ring coil is 40 cm and the coil current is 500 kA. Concerning to the levitated coil, the high-temperature (high-Tc) super-conducting coil is preferable for plasma experiments, because long pulse and/or high power heating experiments might be available due to the good property for the thermal stability and large heat capacity of the high-Tc super-conducting coil. Our primary candidate is Bi-2223 super-conducting cable. Since the maximum magnetic field strength is around 2 T in our device, the deterioration of the critical current is not so severe up to 40 K. We are now promoting a detailed design of the toroidal device with a high-Tc super-conducting internal coil.
- OSTI ID:
- 21210382
- Journal Information:
- AIP Conference Proceedings, Vol. 498, Issue 1; Conference: Workshop on non-neutral plasma physics III, Princeton, NJ (United States), 2-5 Aug 1999; Other Information: DOI: 10.1063/1.1302143; (c) 1999 American Institute of Physics; Country of input: International Atomic Energy Agency (IAEA); ISSN 0094-243X
- Country of Publication:
- United States
- Language:
- English
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