Abstract
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.
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