Numerical simulation of the stability in a cable-in-conduit conductor developed for fusion magnet applications

The stability margins of the US-Demonstration Poloidal Coil (US-DPC) and the International Thermonuclear Experimental Reactor (ITER) toroidal field (TF) coils have been modeled numerically using the computer program CICC. The computed US-DPC limiting current, I _lim, compares favorably with the values determined experimentally. Using the detailed program CICC output, the authors investigated the DPC quench initiation mechanism in each of the three stability regions. In the ill-cooled region, the imposed heat pulse heats the conductor to the current-sharing temperature. In the transition region, the resistance heating after the pulse must be strong enough to overcome the induced flow reversal. In the well-cooled region, good heat transfer heats the helium during the pulse. After the pulse, these high helium temperatures along with poor heat transfer cause the conductor to quench. Changes in I_lim agree with Dresner´s relationship. I_lim can be improved by decreasing the copper resistivity, the helium fraction, or the conductor diameter. Preliminary results show the ITER TF coil operating point is in the well-cooled region