Recently, the conceptual design of Steady State Tokamak (SST-1) has been concluded at Institute for Plasma Research, India. In SST-1, all but one pair of PF coils are superconducting. A NbTi based 10 KA Cable-in-Conduit-Conductor (CICC) having a cabling pattern of 3 X 3 X 3 X 5 has been chosen as the base conductor for the superconducting PF coils. The resistive pair of PF coils placed inside the vacuum vessel will have 20 X 20 mm X mm internally cooled Cu based conductor. The SST-1 PF magnets are designed for running both double null and single null nominal (220 KA) and high current (330 KA) plasmas and to ramp up the plasmas to 220 KA and 330 KA in 3 seconds in conjugation with Lower Hybrid Current Drive (LHCD). A wide range of reference equilibria with elonglation in the range of 1.7-1.9, triangularity in the range of 0.4-0.7, l_i in the range of 0.75-1.4 along with the control of the divertor strike points have been studied with the SST-1 PF coils.

The steady-State operation implies a steady-state heat removal capability from the plasma. Since the power and particle exhaust efficiency is closely linked to the divertor strike point location, the PF magnets are required to have the capability to fix such a location for a variety of l_i and b_p values. Plasmas with higher l_i are increasingly difficult to elongate and are are more prone to vertical instability, on the other hand plasmas with lowerl_i load the coils which produce traingularity. The PF optimization which has been carried out for SST-1 therefore is a complex process where space constraints and operational flexibility clash with each other. Presently, the engineering design and manufacturing details of the SST-1 PF coils are being carried out. The PF coil physical and operational characteristics, winding pack details, insulation and support details, interfaces and hydraulic schemes will be discussed in this paper.