The compact nature of the spherical tokamak (ST) presents an economically attractive path to fusion commercialization, but concentrates power exhaust, threatening the integrity of plasma-facing components. To address this challenge, experimentally constrained divertor modeling in the National Spherical Torus Experiment (NSTX) is extrapolated to investigate divertor concepts for future ST devices. Analysis of NSTX Upgrade with UEDGE shows that the secondary snowflake X-point position can be adjusted for favorable neutral transport, enabling stable partial detachment at reduced core densities. For a notional ST-based Fusion Nuclear Science Facility, divertor concepts are identified that provide heat flux mitigation (<10 MW m−2) and low temperatures (<10 eV) compatible with high-Z targets. This research provides guidance for upcoming experiments and a basis for continued development of predictive capability for divertor performance in STs.