The poloidal field (PF) coil system on ITER, which provides both feedforward and feedback control of plasma position, shape, and current, is a critical element for achieving mission performance. Analysis of PF capabilities has focused on the 15 MA Q = 10 scenario with a 300–500 s flattop burn phase. The operating space available for the 15 MA ELMy H-mode plasma discharges in ITER and upgrades to the PF coils or associated systems to establish confidence that ITER mission objectives can be reached have been identified. Time dependent self-consistent free-boundary calculations were performed to examine the impact of plasma variability, discharge programming and plasma disturbances. Based on these calculations a new reference scenario was developed based upon a large bore initial plasma, early divertor transition, low level heating in L-mode and a late H-mode onset. Static equilibrium analyses for this scenario, which determine PF coil currents to produce a given plasma configuration, indicate that the original PF coil limitations do not allow low li(<0.8) operation or plasmas with lower flux consumption, and the flattop burn durations were predicted to be less than the desired 400 s. This finding motivates the expansion of the operating space, considering several upgrade options to the PF coils. Analysis was also carried out to examine the feedback current reserve required in the central solenoid and PF coils during a series of disturbances, heating and current drive sources for saving volt-seconds in rampup, a feasibility assessment of the 17 MA scenario was undertaken, and the rampdown phase of the discharge is discussed. Results of the studies show that the new scenario and modified PF system will allow a wide range of 15 MA 300–500 s operation and more limited but finite 17 MA operation.
H-mode plasmas in the pre-fusion power operation 1 phase of the ITER research plan