The KSTAR (Korea Superconducting Tokamak Advanced Research) project is the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state-capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for the first plasma by mid 2002. The key design features of KSTAR are: major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, and flexible plasma shaping (elongation 2.0; triangularity 0.8; double-null poloidal divertor). Both the toroidal and the poloidal field magnets a r e superconducting coils. The device is configured to be initially capable of 20s pulse operation and then to be upgraded for operation up to 300s with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beam, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with a full radliation shielding.

The KSTAR team is carrying out the design and research and development for a steady-state-capable advanced superconducting tokamak to establish the basis for an attractive fusion reactor as a future energy source. The physics requirements are driven by the plasma control and exhaust capabilities needed to extend the performance and pulse length of tokamak plasmas. The tokamak has major radius 1.8m, minor radius 0.5 m, toroidal field 3.5 T and plasma current 2 MA, a strongly shaped plasma cross-section shaping (elongation 2.0 and triangularity 0.8), and a double-null poloidal divertor. The initial pulse length is 20 s, long enough to study physics on confinement timescales, but short enough to permit economical plasma-facing component technology. The pulse length can be increased to 300 s through upgrades. The machine will be operable in either hydrogen or deuterium, but neutron yields will be constrained to avoid the cost and inconvenience of remote maintenance and low-activation materials. The magnet system provides an inductively driven 20 s pulse with full current, beta, and shaping. With non-inductive current drive, steady-state plasmas can be sustained over a wide range of profile shapes and plasma pressures. Passive structures are provided to stabilize the vertical instability and high-beta modes and internal coils are provided for fast position control. The divertor structures are designed for particle removal, recycling control, impurity control, and flexibility for advanced divertor operation. The plasma heating system is designed for heating, current-drive, profile control, and flexibility. It will deliver power via neutral beams (8 MW), ion-cyclotron waves (6 MW), and lower-hybrid waves (1.5 MW), each of which can be expanded through upgrades. A comprehensive set of diagnostics is planned for plasma control, performance evaluation, and physics understanding.