[en] One of the key challenges for the conventional divertor tokamak is the plasma wall interaction interface and its materials. High divertor heat loads and material erosion in the spherical torus (ST) are of particular concern because of the compact divertor, and as a result, small plasma-wetted surfaces. The implications of the toroidal plasma physics at low aspect ratio and high β for edge energy and particle transport, properties of the scrape-off layer (SOL) and divertor are being studied on the National Spherical Torus Experiment (NSTX)--a medium size ST (R = 0.85 m, a = 0.67 m, A ≅ 1.27, β_t < 32 %, β_N < 5 %). NSTX operates routinely with stationary outer target plate peak heat loads up to 6 MW/m² for up to 1 s in the 6 MW NBI heated H-mode regime with type I, III, V ELMs , with the largest peak heat flux measured to date q_out = 10 MW/m² A detached divertor is an effective heat flux mitigation technique which has been developed in large aspect ratio tokamaks. Heat flux at the plate is reduced in the detached divertor through volumetric momentum and energy dissipative processes--the ion-neutral elastic collisions, recombination and radiative cooling