Impact of a poloidal divertor in ignition tokamak design
System design studies were performed to assess the effect of assuming a poloidal divertor instead of a limiter as a means of impurity control for ignition tokamak configurations. Results show that for the nominal Tokamak Fusion Core Experiment (TFCX) device with superconducting TF coils, a feasible poloidal divertor configuration can be obtained without increasing the major radius. In the TFCX nominal copper TF coil device, however, field limits at the PF coils are exceeded when the effects of asymmetry associated with a poloidal divertor are included. It was found that a 12% increase in the major radius of this device is necessary to simultaneously satisfy the plasma-shaping requirements of a poloidal divertor and the magnetics constraints at the superconducting PF coils.
- Research Organization:
- Oak Ridge National Lab., TN (USA); Grumman Aerospace Corp., Bethpage, NY (USA); Science Applications International Corp., Oak Ridge, TN (USA); McDonnell Douglas Corp., St. Louis, MO (USA); Tennessee Univ., Knoxville (USA)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 5665256
- Report Number(s):
- CONF-850310-106; ON: DE85012892
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
700202* -- Fusion Power Plant Technology-- Magnet Coils & Fields
DESIGN
DIVERTORS
ELECTRIC COILS
ELECTRICAL EQUIPMENT
ELECTROMAGNETS
EQUIPMENT
MAGNET COILS
MAGNETS
PLASMA
POLOIDAL FIELD DIVERTORS
SHAPE
SUPERCONDUCTING DEVICES
SUPERCONDUCTING MAGNETS
SYMMETRY
THERMONUCLEAR REACTORS
TOKAMAK TYPE REACTORS