Cable-in-conduit superconductors for fusion magnets: electro-magnetic modelling for understanding and optimizing their transport properties

Cable-In-Conduit conductors feature large current-carrying capacity and stability against local and transient heat deposition. As such they are suitable for application in superconducting magnets for nuclear fusion, as the ones of the International Thermonuclear Experimental Reactor (ITER). Due to the high cost connected to sample assembly and test, simulation is an essential tool for studying and optimizing the performance of CIC conductors. However, the task is made complex by the cabling of some 1000 strands in multiple stages, resulting in an entangled network of electrical and thermal contacts. The numerical code JackPot-ACDC allowing the simulation of CIC conductors and their lap-type joints with strand-level detail is validated under DC and AC conditions. The code is used to investigate the influence of sample and joint layouts on the current distribution in large NbTi-type CIC conductors. For the first time the heating power and current non-uniformity are quantified and visualized. The maximum values for the design of the ITER Poloidal Field coils are specified. The relation between twist pitches sequence and inter-strand coupling loss induced by time-varying magnetic field is parametrically explored in multi-stage ITER-type CIC conductors. The analysis shows that inter-strand coupling loss can be controlled, even in the case of long twist pitches, by keeping the stage-to-stage twist pitch ratio close to 1. The prediction is successfully confirmed by experimental measurements. The performance of four conductors proposed for the ITER Central Solenoid featuring different twist pitches and cabling patterns is characterized under operating plasma scenario conditions. Plasma scenario behaviour is also analysed for the lap-type joints of the ITER Toroidal and Poloidal Field coils. The effect of possible design optimization options is assessed and discussed