Poloidal Field Coil Configuration and Plasma Shaping Capability in NCT

This paper describes the latest design of poloidal field coil configuration in the national centralized tokamak (NCT), that is a new design based on the former superconducting coil tokamak JT-60SC. The most notable design change was made for the outer equilibrium field (EF) coils to increase a plasma shaping parameter
$ S(= q_95ast I_ p/ aast B_ t)$
and to make off-axis heating using N-NBI possible. As a result, the maximum plasma shaping parameter of
$sim$
7 was obtained by double null divertor configuration with a lower aspect ratio of A
$sim$
2.6. It was confirmed that two additional EF coils are useful for the plasma squareness control, but only one is finally adopted for ITER plasma simulation and to keep compatibility of off-axis N-NBI heating.