The National Spherical Torus Experiment (NSTX) has made considerable progress in advancing the scientific understanding of high performance long-pulse plasmas needed for future spherical torus (ST) devices and ITER. Plasma durations up to 1.6 s (five current redistribution times) have been achieved at plasma currents of 0.7 MA with non-inductive current fractions above 65% while simultaneously achieving beta(T) and beta(N) values of 17% and 5.7 (%m T MA(-1)), respectively. A newly available motional Stark effect diagnostic has enabled validation of current-drive sources and improved the understanding of NSTX 'hybrid'-like scenarios. In MHD research, ex-vessel radial field coils have been utilized to infer and correct intrinsic EFs, provide rotation control and actively stabilize the n = 1 resistive wall mode at ITER-relevant low plasma rotation values. In transport and turbulence research, the low aspect ratio and a wide range of achievable in the NSTX provide unique data for confinement scaling studies, and a new microwave scattering diagnostic is being used to investigate turbulent density fluctuations with wavenumbers extending from ion to electron gyro-scales. In energetic particle research, cyclic neutron rate drops have been associated with the destabilization of multiple large toroidal Alfven eigenmodes (TAEs) analogous to the 'sea-of-TAE' modes predicted for ITER, and three-wave coupling processes have been observed for the first time. In boundary physics research, advanced shape control has enabled studies of the role of magnetic balance in H-mode access and edge localized mode stability. Peak divertor heat flux has been reduced by a factor of 5 using an H-mode-compatible radiative divertor, and lithium conditioning has demonstrated particle pumping and results in improved thermal confinement. Finally, non-solenoidal plasma start-up experiments have achieved plasma currents of 160 kA on closed magnetic flux surfaces utilizing coaxial helicity injection.

Extensive lithium wall coatings and liquid lithium plasma-limiting surfaces reduce recycling, with dramatic improvements in ohmic plasma discharges in the Current Drive eXperiment – Upgrade (CDX-U). Global energy confinement times increase by up to 6×. These results exceed confinement scalings such as ITER98P(y,1) by 2-3×, and represent the largest increase in confinement ever observed for an ohmic tokamak plasma. Measurements of Dα emission indicate that global recycling coefficients decrease to approximately 0.3, the lowest documented for a magnetically confined hydrogen plasma. PACS numbers: 52.55.Fa, 52.40.Hf, 52.25.Fi