Abstract
Experiments in the National Spherical Torus Experiment ( NSTX) have shown beneficial effects on the performance of divertor plasmas as a result of applying lithium coatings on the graphite and carbon-fiber-composite plasma-facing components. These coatings have mostly been applied by a pair of lithium evaporators mounted at the top of the vacuum vessel which inject collimated streams of lithium vapor toward the lower divertor. In neutral beam injection (NBI)-heated deuterium H-mode plasmas run immediately after the application of lithium, performance modifications included decreases in the plasma density, particularly in the edge, and inductive flux consumption, and increases in the electron and ion temperatures and the energy confinement time. Reductions in the number and amplitude of edge-localized modes (ELMs) were observed, including complete ELM suppression for periods of up to 1.2 s, apparently as a result of altering the stability of the edge. However, in the plasmas where ELMs were suppressed, there was a significant secular increase in the effective ion charge Z(eff) and the radiated power as a result of increases in the carbon and medium-Z metallic impurities, although not of lithium itself which remained at a very low level in the plasma core, <0.1%. The impurity buildup could be inhibited by repetitively triggering ELMs with the application of brief pulses of an n = 3 radial field perturbation. The reduction in the edge density by lithium also inhibited parasitic losses through the scrape-off-layer of ICRF power coupled to the plasma, enabling the waves to heat electrons in the core of H-mode plasmas produced by NBI. Lithium has also been introduced by injecting a stream of chemically stabilized, fine lithium powder directly into the scrape-off-layer of NBI-heated plasmas. The lithium was ionized in the SOL and appeared to flow along the magnetic field to the divertor plates. This method of coating produced similar effects to the evaporated lithium but at lower amounts.