Introduction At present, the National Spherical Torus Experiment (NSTX) facility is be-ing upgraded to new capabilities to enable physics studies of the spherical tokamak (ST) to advance the ST as a candidate for Fusion Nuclear Science Facility. In the NSTX-U device [1], discharges with I p ≤ 2 MA and P NBI ≤ 12.3 MW and up to 5 s duration are projected to produce steady-state peak divertor heat fluxes in the range 20-30 MW/m 2 , thereby challeng-ing thermal limits of divertor graphite PFCs [2]. The leading heat flux mitigation candidates for NSTX-U are considered to be the snowflake (SF) divertor geometry [3] and the impurity-seeded radiative divertor (RD) technique, applied to the lower and upper divertors. Experiments in NSTX, a large spherical tokamak with lithium-coated graphite PFCs and high divertor heat flux (q peak ≤ 15 MW/m 2 , q ≤ 200 MW/m 2 [2]), have produced a basis for NSTX-U projec-tions and initial experiments. In NSTX, operations with both the RD and the SF divertor have been quite successful: a significant reduction of divertor heat flux, from peak values of 4-10 MW/m 2 to 0.5-2 MW/m 2 , simultaneously with good core H-mode confinement characterized by H98(y,2) up to 1, have been demonstrated in 1.0-1.3 s discharges [4, 5, 6, 7, 8, 9]. Owing to a compact divertor with intrinsic carbon radiation, the RD with a partially detached strike point was achieved in NSTX using either (1) divertor D 2 injection in the standard (albeit high poloidal magnetic flux expansion f m 20) divertor geometry, or (2) in the SF divertor geometry, due to its geometric effects. The SF divertor geometry created in NSTX (w.r.t. the standard divertor) offered an up to 100 % higher plasma wetted area (due to a much higher flux expansion), a longer connection length and a larger divertor volume available for volumetric losses [8, 9].