29aA01
Plasma-wall boundary control with lithium divertor and 	 associated plasma confinement improvements in NSTX*
Masayuki Ono1, M.A. Jaworski1, R. Kaita1, and the NSTX-U Research Team1

Princeton Plasma Physics Laboratory, PO Box 451, Princeton, NJ 08543, USA.

Developing a reactor compatible divertor has
been identified as a particularly challenging
technology/physics
problemfor
magnetic
confinement fusion. While tungsten has been
identified as the most attractive solid divertor
material, many challenges including surface
cracking and deleterious modification of thesurfaces by the plasma must be overcome todevelop robust plasma facing components (PFCs).
Alternately, a possibility of plasma-wall boundary
control with lithium divertor and associated plasma
confinement improvements has been receiving
increasing interest in the recent years [1-2]. On
NSTX, the effects of lithium coatings (both in solidand liquid phases) have been investigated on thedivertor in H-mode plasmas. A liquid lithium
divertor plate system was tested with about 1300 gof lithium evaporated into the NSTX vacuum vessel.
The overall NSTX lithium coating results suggest
attractive opportunities for future magnetic
confinement research including H-mode power
threshold reduction, the control of Edge Localized
Modes (ELM), electron energy confinement
improvements,
impurity
controlduringnon-inductive plasma start-up, and improving radio
frequency wave heating and current drive
efficiencies. Another intriguing result from NSTX
is that the surprisingly low levels of lithium ionsobserved in the NSTX H-mode plasma core (below
0.1% dilution) compared to higher Z impurities(such as carbon) even when the plasma isessentially surrounded by the lithium coated walls.
This observation bodes well for lithium based
plasma facing component and divertor applications.
Importantly, with lithium coating, measurements
showed a significant ~ 50% reduction in heat loadon the LLD, indicated by enhanced bolometricradiation above the divertor surface. Based on theexperimental results from NSTX with its liquid
lithium divertor (LLD) and experimental resultsfrom other devices and test stands, promising
divertor solutions based on liquid lithium areemerging [2].
In terms of handling the very high heat fluxexpected on the tokamak reactor divertor strikepoint, a liquid lithium (LL) based closed radiative

divertor concept (RLLD) has
proposed as shown in Fig.1 [3].

been

recently

Fig.1. A possible RLLD configuration in a fusion
power plant.

Fig. 2. A schematic for the RLLD concept.

With an LL coating, the lithium is evaporated
from the divertor strike point surface due to theintense heat as the lithium evaporation increases
rapidly with the surface LL temperature. Theevaporated lithium is readily ionized by the plasma
due to its low ionization energies, and the ionizedlithium ions can radiate strongly, resulting in a
reduction in the divertor heat flux. A schematic ofthe RLLD concept is shown in Fig. 2. This radiative
process has the desirable function of spreading thenarrowly focused divertor heat load to the entire
divertor chamber, which facilitates the divertor heatremoval. Since the projected Li radiation can bevery high (i.e., ~100 MJ / mole of Li ions), it should
be readily feasible to provide an adequate amountof LL in the high heat flux divertor region even forsteady-state operation. For example, to radiate an
estimated 200 MW of influx power into divertor for
a 1 GW-electric class fusion power plant, it would
only take about two moles of Li per sec to handle
the divertor heat load. One can readily provide
higher evaporation rate if necessary. A 1-D model
calculation of RLLD has been performed with a
two-point model in a cylindrical geometry [3]. With
Li radiation level of 10-26 W-cm3, which assumes
relatively poor confinement of neτ = 109 – 1010 in
the divertor region, the model predicts only ~ 1018
Li particles would be needed for an ST-based FNSF
with R= 1.5 m and ΔR = 2 cm. If the Li radiative
mantle could reduce the direct divertor heat flux
down to the vicinity of 5 MW /m2, then the solid
substrate material (perhaps made out of tungsten)
could become viable as long as it is coated with Li.
The LL surface can also provide a “sacrificial”
surface to protect the substrate solid material fromtransient high heat loads such as the ones caused bylarge ELMs. If the transient heat flux is high, thatmuch more Li would be evaporated and ionized,
which would then increase the radiative cooling
until an equilibrium condition is reached. For
example, in an ITER scale tokamak reactor, withthe enhanced radiative process, only a modest
amount (~ 1cc) of LL is needed in principle toradiate the expected heat pulse of ~ 10 MJ for an
exceptionally large ELM event.
The closed radiative LLD concept has thefollowing additional attractive features: 1. Some
degree of shielding for LL is provided from fastdisruptive events. Placed within the chamber,even if LL is “splashed”, it can be largely contained
within the chamber. 2. Particle partition is provided
from the main plasma chamber due to the strong
divertor lithium retention. 3. Strongly divertor
pumping action is provided from the divertor
chamber wall as the vaporized and ionized lithium
ions coat the entire divertor chamber wall. Thestrong pumping would access a low recycling
regime for advanced high performance plasma
operations. 4. Lastly, by operating the divertor
lithium condensing surfaces at lower temperature(200-450°C) than the first wall (500-700°C), theLLD can serve as a purifying / pumping system forthe entire reactor chamber as impurities generally
migrate toward lower temperature surfaces inaddition to the natural divertor flow action. Tomaintain the LL purity, a closed LL loop systemwith only a modest circulating capacity (~1 liter/secfor ~ 1% level “impurities”) will be needed in a
steady-state reactor operation. A schematic for a
possible LL purification loop is shown in Fig. 3.

Fig.3. A schematic for the RLLD purification loop.

Finally, it should be emphasized that Li application
should be quite compatible with various divertor
geometry and magnetic confinement configurations
providing the same benefits of Li. Application of Li
may also be considered for protecting the tungsten
based solid PFC surfaces such as the ones for ITER,
as long as a means to purity/refresh LL can beprovided. In summary, a radiative mantle based LL
divertor solution have the exciting prospect ofproviding a cost effect flexible means to improve
the fusion reactor performance, while providing apractical solution to the highly challenging divertor
heat handling issue confronting steady-statemagnetic fusion reactors.
*This work was supported by DoE Contract No.
DE-AC02-09CH11466.
References:[1] Hirooka Y., Mazzitelli G., Mirnov S.V. et al., Nucl.
Fusion 50, 077001 (2010).
[2] M. Ono, M.G. Bell, Y. Hirooka, et al., Nuclear
Fusion 51, 037001 (2012).
[3] M. Ono, et al., 24th IAEA Fusion Energy conference(FEC 2012).