Recent Progress Of NSTX Lithium Program And Opportunities For Magnetic Fusion Research
Authors: M. Ono, et. al.
Abstract:
Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment
(NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium
injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up
to ~160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium
research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode
divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results.
In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including
the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal
physics, and enhanced plasma performance. To fill the LLD with lithium, 1300g of lithium was evaporated into
the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly
improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this
paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing
fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and
associated R&D required for use of lithium in future magnetic fusion facilities including ITER.
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Submitted to: Fusion Engineering and Design (May 2011)
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Download PPPL-4632 (pdf KB pp)
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