In simulating reference scenarios proposed for ITER operation, we also explore performance of the poloidal field and central solenoid coil systems using a controller to maintain plasma shape and vertical stability during the discharge evolution. We employ a combination of techniques to evaluate system constraints and stability using time-dependent transport simulations of ITER discharges. We have begun the process of benchmarking these simulations with experiments on the DIII-D tokamak. Simulations include startup on the outside limiter, X-point formation and current ramp up to full power, plasma burn conditions at 15 MA and 17 MA, and ramp down at the end of the pulse. We also simulate perturbative events such as H-to-L back transitions. Our results indicate the viability of proposed ITER operating modes. 1. S imulation techniques In solving for free-boundary equilibria with thermal transport in the CORSICA [1] code, we use a combination of transport simulation techniques [2] to assess coilsystem performance and stability. In fast simulations to explore parameter variations, we use a series of prescribed shapes to evolve the discharge. After each time step, coil currents consistent with the shape evolution are obtained from a subsequent freeboundary solution. For forward control, the ITER controller (JCT-2001, VS1) maintains the plasma via feedback control of the reference shape at the gap (difference between a reference separatrix and the boundary obtained) locations. In this case, vertical stability is provided by the fast feedback loop response controlling the equilibrium centroid velocity. In these forward simulations, only the controllerpredicted poloidal field (PF) and central solenoid (CS) coil voltages are fed back to the equilibrium code to maintain control. The coil currents obtained are sensitive to the transport assumptions used in simulations with the resistive flux evolution altering the overall coil current demands. Using scaling of ITER size and time scale, a series of startup similarity experiments are being conducted on the DIII-D tokamak [3,4] and are providing data to validate the controller transport modeling described here. In modeling ITER during the high auxiliary heating phase (burn), we generate an H-mode-like temperature pedestal from the transport model and with the prescribed density profile calculate the edge bootstrap current peak from a neoclassical model. The presence of this edge current alters the internal inductance [l i (3) used here] and is important for exploring stability and control system performance. Coil current limits stemming from the magnetic field strength and forces at the conductors are monitored in these simulations to further assess the adequacy of the coil system. During plasma evolution, we calculate the vertical instability growth rates. Equilibria are saved during the scenario simulations and available for additional evaluation of the stability and for shape optimization studies. We have explored H-to-L and L-to-H transitions, beta (normalized stored energy) and l i collapses, and controller-induced disturbances by suspending control and allowing the plasma to drift vertically unstable.