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Three types of high temperature superconducting current leads (HTSCL) are designed to carry 68 kA, 55 kA or 10 kA to the ITER magnets. Before the supply of the HTS current lead series, the design and manufacturing process is qualified through mock-ups and prototypes. Seven mock-ups, representing the critical technologies of the current leads, were built and tested successfully in the Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) in 2013. After the qualification some design features of the HTS leads were updated. This paper summarizes the qualification through mock-ups. In 2014 ASIPP started the manufacture of the prototypes. The preparation and manufacturing process are also described. [ABSTRACT FROM AUTHOR]

• The ITER current lead insulation consist of interleaved pre-preg and polyimide tapes. • The acceptance high voltage test for current lead includes DC, Paschen and PD test. • ASIPP has completed the insulation qualification of ITER current lead. The ITER high temperature superconducting current lead is a critical component for the magnet system, which has the benefit of reduction in the heat load of the cryogenic system compare with the conventional lead. The current lead is located in the coil terminal box and dry box at the end of the ITER feeder system. As a warm to cold transition section, the current leads feed the huge current from the power supply system into the coils. At the outer surface of the current lead, one layer of a composite insulation is applied to isolate the high voltage potential of the current lead to the ground potential of the environment. The main body of the current lead is that of a long cylinder, but at the cold termination and the cooling inlet, the local geometry is much more irregular. So the insulation wrapping and curing technology of the current lead had to be developed to acquire its uniform mechanical and electrical performance. Now, the multi-stage autoclave curing technology has been qualified in ASIPP and the series ITER current leads are being manufactured based on the qualified procedure. In this paper, the latest insulation progresses for the ITER current lead are introduced, the high voltage testing results as part of the formal qualification are presented and discussed. [ABSTRACT FROM AUTHOR]

• A pair of 100 kA high temperature superconducting (HTS) current leads for CFETR is being designed at ASIPP. • The analysis of the maximum current carrying capacity of HTS module with self-field and external field has been performed. • The simulation of LOFA and burnout time for the 100 kA HTS current lead are also discussed. • The analysis results indicate that maximum current carrying capacity, LOFA and burnout time meet CFETR requirement. Based on the Agreement of Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Ministry of Science and Technology of China during the Thirteen Five-Years Plan (so called 13∙5 plan), a dedicated test facility of superconducting magnet for China Fusion Engineering Test Reactor (CFETR) will be constructed in Hefei, China. A pair of high temperature superconducting (HTS) current leads rated at 100 kA for the CFETR magnet test facility is being designed at ASIPP. Because magnetic energy stored in the toroidal field magnet is 134 Giga Joules which are 3 times higher than the ITER case, HTS current leads must work at nominal current without performance degradation during the HTS current lead quench. In order to verify the safety and reliability of HTS current lead, the analysis of static magnetic field is done by ANSYS iteration to predict the maximum current carrying capacity of HTS module. This paper also give the numerical calculation of loss of flow accident(LOFA) of HTS current lead and temperature margin analysis of HTS module. The analysis results indicate that the HTS module has a current carrying capacity of 165 kA without external field at 65 K and a current carrying capacity of 149 kA with external 30 mT vertical field at 65 K. The LOFA time is about 580 s and the burnout time is about 30 s, which is better than the requirement of the toroidal field coil fast discharge time of 24 s. [ABSTRACT FROM AUTHOR]