Effect of the poloidal field coil system on the ramp-up scenario with L/H transition

Improved confinement is indispensable for designing an attractive fusion reactor, and an H-mode is one of the most prominent candidates for a fusion reactor plasma. It is experimentally clear that there exists a minimum heating power so as to access the H-mode confinement region. This threshold power has been investigated in many tokamak devices, and recently compiled with some plasma parameters. We simulated the plasma ramp-up of the International Thermonuclear Experiment Reactor, Engineering Design Activity (ITER EDA) by using a global model simulation based on a power balance equation and a helium particle balance equation with L- to H- and H- to L-mode transitions. A new ramp-up method was adopted, whereby the surface area of the plasma is varied during the ramp-up phase to save the auxiliary heating power for ignition or for operating at an elevated density. This method reduces the heating power from 100 to 30 MW or increases the initial density from 0.5×1020 to 0.9×1020 m−3 when the threshold power is in proportion to the plasma surface. The PF coil system is discussed with this ramp-up method and a plasma equilibrium code, and it is found that when the plasma is ramped up from the outer-side of the torus such as in the ITER-EDA design, the power supply of 1.7 GW is sufficient for PF coil system, being independent of the winding method of the central solenoid (CS) coil. However the inner-side ramp-up, which is preferable for reducing L/H transition power, can be realized only when the flexibility of the CS coil is preserved with a pancake winding.