The National Spherical Torus Experiment (NSTX) has as one of its primary goals the demonstration of the attractiveness of the spherical torus concept as a fusion power plant. Central to this goal is the achievement of high plasma β (=2μ
0<p>B
2 a measure of the efficiency of a magnetic plasma confinement system). It has been demonstrated both theoretically and experimentally that the maximum achievable β is a strong function of both local and global plasma parameters. It is therefore important to optimize control of the plasma. To this end a phased development plan for digital plasma control on NSTX is presented. The relative level of sophistication of the control system software and hardware will be increased according to the demands of the experimental program in a three phase plan. During Day 0 (first plasma), a simple coil current control algorithm will initiate plasma operations. During the second phase (Day 1) of plasma operations the control system will continue to use the preprogrammed algorithm to initiate plasma breakdown but will then change over to a rudimentary plasma control scheme based on linear combinations of measured plasma fields and fluxes. The third phase of NSTX plasma control system development will utilize the rtEFIT code, first used on DIII-D, to determine, in real-time, the full plasma equilibrium by inverting the Grad-Shafranov equation. The details of the development plan, including a description of the proposed hardware are presented
