Introduction The spherical tokamak (ST) is viewed as a candidate concept for future fusion and nuclear science applications [1,2]. Divertor experiments in NSTX, a high-power density medium size ST (R = 0.85 m; a = 0.65 m) with graphite plasma-facing components (PFCs), have demonstrated the features of the inherently compact ST divertor. ITER-scale steady-state peak divertor heat fluxes qpk < 15 MW/m2 and q\\ < 200 MW/m2 have been measured in Ip = l.o -1.2 MA discharges heated by 6 MW NBI . As a result of this and other ST- or NSTX-specific geometry features, e.g., a small in/out SOL power ratio, a small divertor PFC area, an open divertor geometry and reduced divertor volumetric (radiated power and momentum) losses, a reduced operating space of the conventional heat flux mitigation techniques, such as the divertor geometry and radiative divertor, has been found [4,5,6]. Viewing this as the challenge and the opportunity for plasma-material interface (PMI) development, NSTX research is now focusing on developing the PMI for future devices, in particular for NSTX-Upgrade , where steady-state qpk < 25 - 40 MW/m2 are predicted based on the present scalings .