Conceptual design of SST-1 tokamak

Conference · Sun Dec 31 04:00:00 EST 1995

OSTI ID:505898

The superconducting steady-state tokamak (SST) will have major and minor radii, R{sub o} and a, of 1.05 m and 0.20 m respectively. The maximum toroidal field will be 2.9 T at R{sub o} and plasma current I{sub p} {approximately} 220 kA. The D-shaped plasma, in a double-null (DN) divertor configuration, will have {kappa}{sub x} lying in the range 1.7--2.0 and {delta}{sub x} {approx_equal} 0.4--0.7, corresponding to q{sub 95} in the range 5--7. At flattop, {beta}{sub p} and {ell}{sub i} would lie in the ranges 0.7--1.4 and 0.6--1.2 respectively. The TF and PF coils will be superconducting (NbTi), while the ohmic transformer and correction coil system will be made of copper. Steady-state current drive, and also part of the current rampup, will be based on lower hybrid current drive (LHCD). ICRF and NBI will provide auxiliary heating. Fast vertical control will make use of saddle coils placed inside the vacuum vessel, while the PF coils will be used for shape control. The outer divertor will have a slot configuration to permit operation with high recycling and impurity injection. The TF and PF systems and vacuum vessel are designed to allow later upgradation to a high-current, larger plasma with I{sub p} = 400 kA, a = 0.29 m, R{sub o} = 1.16 m, with lower elongations.

• OSTI ID: 505898
• Report Number(s): CONF-950905--; ISBN 0-7803-2970-8
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
BEAM INJECTION HEATING
CONTROL SYSTEMS
CRYOGENICS
DESIGN
DIVERTORS
ICR HEATING
LOWER HYBRID CURRENT DRIVE
PLASMA DIAGNOSTICS
SUPERCONDUCTING MAGNETS
TOKAMAK DEVICES
VACUUM SYSTEMS