Abstract
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.
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