Periods of edge localized mode (ELM)-free H-mode with increased pedestal pressure and width were observed in the DIII-D tokamak when density fluctuations localized to the region near the separatrix were present. Injection of a powder of 45 µm diameter lithium particles increased the duration of the enhanced pedestal phases to up to 350 ms, and also increased the likelihood of a transition to the enhanced phase. Lithium injection at a level sufficient for triggering the extended enhanced phases resulted in significant lithium in the plasma core, but carbon and other higher Z impurities as well as radiated power levels were reduced. Recycling of the working deuterium gas appeared unaffected by this level of lithium injection. The ion scale, k θ ρ s ∼ 0.1-0.2, density fluctuations propagated in the electron drift direction with f ∼ 80 kHz and occurred in bursts every ∼1 ms. The fluctuation bursts correlated with plasma loss resulting in a flattening of the pressure profile in a region near the separatrix. This localized flattening allowed higher overall pedestal pressure at the peeling-ballooning stability limit and higher pressure than expected under the EPED model due to reduction of the pressure gradient below the 'ballooning critical profile'. Reduction of the ion pressure by lithium dilution may contribute to the long ELM-free periods.

Reduction or elimination of edge localized modes (ELMs) while maintaining high confinement is essential for future fusion devices, e.g., the ITER. An ELM-free regime was recently obtained in the National Spherical Torus Experiment, following lithium (Li) evaporation onto the plasma-facing components. Edge stability calculations indicate that the pre-Li discharges were unstable to low-n peeling or ballooning modes, while broader pressure profiles stabilized the post-Li discharges. Normalized energy confinement increased by 50% post Li, with no sign of ELMs up to the global stability limit.

Extensive lithium wall coatings and liquid lithium plasma-limiting surfaces reduce recycling, with dramatic improvements in ohmic plasma discharges in the Current Drive eXperimentUpgrade (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.

National Spherical Torus Experiment [which M. Ono et al., Nucl. Fusion 40, 557 (2000)] high-power divertor plasma experiments have shown, for the first time, that benefits from lithium coatings applied to plasma facing components found previously in limited plasmas can occur also in high-power diverted configurations. Lithium coatings were applied with pellets injected into helium discharges, and also with an oven that directed a collimated stream of lithium vapor toward the graphite tiles of the lower center stack and divertor. Lithium oven depositions from a few milligrams to 1g have been applied between discharges. Benefits from the lithium coatings were sometimes, but not always, seen. These benefits sometimes included decreases in plasma density, inductive flux consumption, and edge-localized mode occurrence, and increases in electron temperature, ion temperature, energy confinement, and periods of edge and magnetohydrodynamic quiescence. In addition, reductions in lower divertor D, C, and O luminosity were measured.

Reductions in H content and particle recycling are important for the improvement of ion cyclotron range of frequency (ICRF) minority heating efficiency and the enhancement of plasma performance of the EAST superconducting tokamak. During recent years several techniques of surface conditioning such as baking, glow discharge cleaning/ICRF discharge cleaning, surface coatings, such as boronization, siliconization and lithium coating, have all been attempted in order to reduce the H/(H+D) ratio and particle recycling in EAST. Even though boronization and siliconization were both reasonably effective methods to improve plasma performance, lithium coatings were observed to reduce the H content and particle recycling to levels low enough to allow the attainment of enhanced plasma parameters and operating modes on EAST. For example, by accomplishing lithium coating using either vacuum evaporation or the real-time injection of fine lithium powder, the H/(H+D) ratio could be routinely decreased to about 5%, which significantly improved ICRF minority heating efficiency during the autumn campaign of 2010. Due to the reduced H/(H+D) ratio and lower particle recycling, and a reduced H-mode power threshold, improved plasma confinement and the first EAST H-mode plasma were obtained. Furthermore, with increasing accumulation of deposited lithium, several new milestones of EAST performance, such as a 6.4 s-long H-mode, a 100 s-long plasma duration and a 1 MA plasma current, were achieved in the 2010 autumn campaign.