The Correction Coils system (CC) within ITER, is intended to reduce the range of magnetic error fields created by assembly or geometrical imperfections of the other coils used to confine, heat, and shape the plasma. The proposed magnet system consists of three sets of 6 coils each, located at the top (TCC), side (SCC) and bottom (BCC) of the Tokamak device and uses a NbTi cable-in-conduit superconducting conductor (CICC) operating at 4.2 K. The ITER Organization (IO) and the Institute of Plasma Physics at the Chinese Academy of Sciences (ASIPP) are jointly preparing the definition of the technical specifications and the upcoming qualification program for the Correction Coils. The proposed design consists of a one in hand conductor winding without internal joint inserted in a structural casing which reacts the electromagnetic loads. The development of major items such as terminal joints, casing manufacture, and vacuum impregnation system, is an essential phase before the series production which will take place at the premises of the supplier. This paper discusses the key technologies on CC coils and future plans for short sample prototypes fabrication.