Pulsed Field Stability and AC Loss of ITER PF Joints by Detailed Quantitative Modeling

For fusion tokamaks, such as International Thermonuclear Experimental Reactor (ITER), the magnet systems are based on superconducting cable-in-conduit conductors (CICCs). Due to the limited production length of the CICCs and the necessary electrical and thermal connections, joints are widely used to connect the CICC units within the magnet coils, as well as the electrical and thermal interfaces outside the magnet. The joints dominate the current redistribution in the coils and, thus, are critical for the magnet stability. In this study, an ITER poloidal field coil joint operating in pulsed mode is chosen to study the current nonuniformity and relevant stability issues. An optimized joint design by applying high resistive polyimide layers, called "masks" to actively block the large induced coupling currents, is tested with the numerical code JackPot-AC/DC. The current nonuniformity, power dissipation, and thermal evolution of the joints with different configurations are quantitatively analyzed.