Small tokamaks for fusion technology testing

Conference · Fri May 01 04:00:00 EDT 1992

OSTI ID:10139346

Peng, Y K.M.; Galambos, J D; Shipe, P C

Small steady-state tokamaks for testing divertors and fusion nuclear technologies are considered. Based on present physics and technology data and explanation to reduce R{sub 0}/a, H-D-fueled tokamaks with R{sub 0} {approximately} 0.6--0.75 m, R{sub 0}/a {approximately} 1.8--2.5, and B{sub t0} {approximately} 1.4--2.2 T can be driven with P{sub tot} {approximately} 4.5 MW to maintain I{sub p} {approximately} 0.5 MA and produce the ITER-level plasma edge and divertor conditions. Given an adequate steady-state divertor solution and Q{approximately}1 operation based on fusion through the suprathermal component, D-T-fueled tokamaks with R{sub 0} {approximately} 0.8 m, R{sub 0}/a {approximately} 2, and B{sub t0} {approximately} 4 T can be driven with P{sub tot} {approximately} 15 MW to maintain I{sub p} {approximately} 4.6 MA and produce an peak neutron wall load W{sub L} {approximately} 1 MW/m{sup 2}. Such devices appear possible if the plasma properties at the power R{sub 0}/a remain tokamak-like and, for the D-T case, can unshielded center core is feasible. The use of a single conductor as the inboard leg of the toroidal field coils for this purpose is discussed. The physics issues and the design features are identified for such tokamaks with a testing duty for factor goal of 10--20%.

• Research Organization: Oak Ridge National Lab., TN (United States)
• Sponsoring Organization: USDOE, Washington, DC (United States)
• DOE Contract Number: AC05-84OR21400
• OSTI ID: 10139346
• Report Number(s): CONF-920607--9; ON: DE92011790
• Country of Publication: United States
• Language: English

Related Subjects

70 PLASMA PHYSICS AND FUSION TECHNOLOGY
700420
DESIGN
DIVERTORS
PLASMA HEATING
PLASMA SCRAPE-OFF LAYER
PLASMA-FACING COMPONENTS
TESTING
TOKAMAK DEVICES