Initial assessments of ignition devices based on the spherical torus concept1 suggest that an ignition spherical torus (IST) can be highly cost-effective and exceptionally small in unit size. Assuming advanced methods of current drive and confinement and beta scalings with plasma current, a D-T IST with a toroidal field of 2 to 3 T is estimated to have a major radius ranging from 1 m to 1.6 m, and a fusion power less than 60 MW. For the nominal IST (at 2 T and 1.6 m), the direct cost of the nuclear island is estimated to be about $120 M with a total direct cost about $340 M in mid-1984 dollars based on the Fusion Engineering Design Center (FEDC) cost algorithm2. For ISTs with higher field and smaller size (e.g., at 3 T and 1 m), further reductions of the cost of the nuclear island are estimated. In case of confinement scaling with the plasma size only, strong plasma paramagnetism (self-generated magnetic field) in the spherical torus may still serve to compensate for the projected confinement shortfall. Because of the modest field strength, only conventional engineering approaches are needed in the IST concepts, leading to dramatic engineering simplifications in comparison with the conventional high-field ignition designs3. A free-standing TF coil/vacuum vessel structure is assessed to be feasible and relatively independent of the shield structure and poloidal field coils. The direct cost of this “stand-alone” torus of the nominal IST is estimated to be $70 M. These highly attractive projections of the IST result directly from a combination of the possible exceptional features of the spherical torus plasma4: high beta, low beta poloidal, naturally large elongation, high plasma current, strong paramagnetism, and tokamak-like confinement, which also place the spherical torus in a plasma regime distinct from tokamaks of conventional aspect ratios. Experimental testing of the viability of the spherical torus concept is suggested.