Authors: M. Ono, R. Majeski, M.A.
Jaworski, R. Kaita, C.H. Skinner and the NSTX-U Research Team
Abstract:
Steady-state fusion power plant designs present major divertor
technology challenges, including high divertor heat flux both in
steady-state and during transients. In addition to these concerns,
there are the unresolved technology issues of long term dust
accumulation and associated tritium inventory and safety issues.
It has been suggested that radiation-based liquid lithium (LL)
divertor concepts with a modest lithium-loop could provide a
possible solution for these outstanding fusion reactor technology
issues, while potentially improving the reactor plasma
performance. The application of lithium (Li) in NSTX resulted in
improved H-mode confinement, H-mode power threshold reduction, and
reduction in the divertor peak heat flux while maintaining
essentially Li-free core plasma operation even during H-modes.
These promising results in NSTX and related modeling calculations
motivated the radiative liquid lithium divertor (RLLD) concept and
its variant, the active liquid lithium divertor concept (ARLLD),
taking advantage of the enhanced Li radiation in relatively poorly
confined divertor plasmas. To maintain the LL purity in a 1
GW-electric class fusion power plant, a closed LL loop system with
a modest circulating capacity of ~1 liter/second (l/sec) is
envisioned. We examined two key technology issues: 1) dust or
solid particle removal and 2) real time recovery of tritium from
LL while keeping the tritium inventory level to an acceptable
level. By running the LL-loop continuously, it can carry the dust
particles and impurities generated in the vacuum vessel to the
outside where the dust / impurities can be removed by relatively
simple dust filter, cold/hot trap and/or a centrifugal separation
systems. With a 1 l/sec LL flow, even a small 0.1% dust content by
weight (or 0.5 g per sec) means that the LL-loop could carry away
nearly 16 tons of dust per year. In a 1 GW-electric (or ~ 3 GW
fusion power) fusion power plant, about 0.5 g / sec of tritium is
needed to maintain the fusion fuel cycle assuming ~ 1 % fusion
burn efficiency. It appears feasible to recover tritium (T) in
real time from LL while maintaining an acceptable T inventory
level. Laboratory tests are also planned to investigate the Li-T
recover efficiency with the SCT concept and also to assess the
viability of the centrifugal Li-T separator with consultation with
the manufacturer.
Submitted to: Nuclear Fusion
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