Poloidal magnetics of a divertor compact ignition tokamak
A technique is presented for calculating bounds on the poloidal field (PF) coil currents required to constrain critical plasma shape parameters when plasma pressure and current density profiles are changed. Such considerations are important in the conceptual design of the PF coils for the Compact Ignition Tokamak (CIT) and their electrical power systems in view of the uncertainty in plasma profiles and operating scenarios. Four relatively independent coil groups are sufficient to find a coil current distribution and equilibrium satisfying a prescribed plasma major radius, minor radius, and divertor strike point coordinates. The variation in the coil current distribution with plasma profiles tends to be large for external PF systems and provides a measure by which coil configurations may be compared. 6 refs., 7 figs., 4 tabs.
- Research Organization:
- Oak Ridge National Lab., TN (USA)
- DOE Contract Number:
- AC05-84OR21400
- OSTI ID:
- 5824944
- Report Number(s):
- ORNL/FEDC-87/2; ON: DE88003648
- Country of Publication:
- United States
- Language:
- English
Similar Records
Poloidal magnetics and divertor strike point control in the Compact Ignition Tokamak
Impact of a poloidal divertor in ignition tokamak design
Related Subjects
700202* -- Fusion Power Plant Technology-- Magnet Coils & Fields
CLOSED PLASMA DEVICES
COMPACT IGNITION TOKAMAK
CURRENT DENSITY
DIVERTORS
ELECTRIC COILS
ELECTRIC HEATING
ELECTRICAL EQUIPMENT
EQUILIBRIUM
EQUIPMENT
HEATING
JOULE HEATING
MAGNET COILS
MHD EQUILIBRIUM
PLASMA HEATING
PLASMA PRESSURE
POLOIDAL FIELD DIVERTORS
RESISTANCE HEATING
THERMONUCLEAR DEVICES
THERMONUCLEAR REACTORS
TOKAMAK DEVICES
TOKAMAK TYPE REACTORS