Experience and technical issues of liquid lithium application as plasma facing material in tokamaks

Lyublinski, I.E.; Vertkov, А.V.

doi:10.1016/j.fusengdes.2010.08.036

Cited by 24 publications

(19 citation statements)

References 9 publications

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“…In Figure 10, the Active Capillary-Porous Systems (CPS) on various fusion devices are shown. The CPS was developed in Russia in 1990's [24,25]. A CPS system utilizes a capillary-porous material to wick the LL from a reservoir to the CPS plasma contacting surfaces.…”

Section: Lithium Delivery Systemsmentioning

confidence: 99%

“…The CPS system principle is also being applied to a new type of high heat flux CPS divertor module being developed, as discussed at the recent lithium workshop [21]. An example for the CPS divertor module planned for the KTM tokamak [24] is shown in Figure 10 (d).…”

Section: Lithium Delivery Systemsmentioning

confidence: 99%

“…There is a question of how to keep the CPS surface free of lithium compounds, which would not flow as well as pure LL. A CPS concept based on a large surface area divertor is now under development to be used in the KTM tokamak [24]. If the CPS surface function is just to provide LL available for evaporation/ionization into the divertor plasma, the CPS surface maybe kept clean by the plasma "scrubbing" action of the plasma.…”

Section: Handling High Divertor Heat Fluxmentioning

confidence: 99%

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Lithium As Plasma Facing Component for Magnetic Fusion Research

Ono¹

2012

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The use of lithium in magnetic fusion confinement experiments started in the 1990's in order to improve tokamak plasma performance as a low-recycling plasma-facing component (PFC). Lithium is the lightest alkali metal and it is highly chemically reactive with relevant ion species in fusion plasmas including hydrogen, deuterium, tritium, carbon, and oxygen. Because of the reactive properties, lithium can provide strong pumping for those ions. It was indeed a spectacular success in TFTR where a very small amount (~ 0.02 gram) of lithium coating of the PFCs resulted in the fusion power output to improve by nearly a factor of two. The plasma confinement also improved by a factor of two. This success was attributed to the reduced recycling of cold gas surrounding the fusion plasma due to highly reactive lithium on the wall. The plasma confinement and performance improvements have since been confirmed in a large number of fusion devices with various magnetic configurations including CDX-U/LTX (US), CPD (Japan), HT-7 (China), EAST (China), FTU (Italy), NSTX (US), T-10, T-11M (Russia), TJ-II (Spain), and RFX (Italy). Additionally, lithium was shown to broaden the plasma pressure profile in NSTX, which is advantageous in achieving high performance H-mode operation for tokamak reactors. It is also noted that even with significant applications (up to 1,000 grams in NSTX) of lithium on PFCs, very little contamination (< 0.1%) of lithium fraction in main fusion plasma core was observed even during high confinement modes. The lithium therefore appears to be a highly desirable material to be used as a plasma PFC material from the magnetic fusion plasma performance and operational point of view. An exciting development in recent years is the growing realization of lithium as a potential solution to solve the exceptionally challenging need to handle the fusion reactor divertor heat flux, which could reach 60 MW/m 2. By placing the liquid lithium (LL) surface in the path of the main divertor heat flux (divertor strike point), the lithium is evaporated from the surface. The evaporated lithium is quickly ionized by the plasma and the ionized lithium ions can provide a strongly radiative layer of plasma ("radiative mantle"), thus could significantly reduce the heat flux to the divertor strike point surfaces, thus protecting the divertor surface. The protective effects of LL have been observed in many experiments and test stands. As a possible reactor divertor candidate, a closed LL divertor system is described. Finally, it is noted that the lithium applications as a PFC can be quite flexible and broad. The lithium application should be quite compatible with various divertor configurations, and it can be also applied to protecting the presently envisioned tungsten based solid PFC surfaces such as the ones for ITER. Lithium based PFCs therefore have the exciting prospect of providing a cost effective

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Section: Lithium Delivery Systemsmentioning

confidence: 99%

Section: Lithium Delivery Systemsmentioning

confidence: 99%

Section: Handling High Divertor Heat Fluxmentioning

confidence: 99%

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Lithium As Plasma Facing Component for Magnetic Fusion Research

Ono¹

2012

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“…In liquid limiter case, limiter represents a non stop flowing liquid metal wall between fusion plasma and first solid wall of the plasma chamber. And one of the main candidates for this solution is lithium [4][5][6]. Lithium also is one of the most advantageous blanket materials not only because of its high heat transfer property to compose a self -cooled fusion reactor system but it also has a high tritium breeding ratio [7].…”

Section: Introductionmentioning

confidence: 99%

Experimental setup for analysis of sorption and desorption of tritium in liquid lithium under different external conditions

Lescinskis

Ķizāne

Vītiņš

et al. 2013

IOP Conf. Ser.: Mater. Sci. Eng.

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“…Later on, different lithium conditioning techniques were developed at other devices, e.g., the capillary porous system (CPS) in the T-10, T-11 and FTU tokamaks [3,4], and the evaporated lithium coating used in NSTX [5] and TJ-II [6,7]. Recently, the CPS concept has been extended to divertor plates, and some examples of implementation in divertor devices are available [8,9]. Thus, lithium is acquiring importance as a material of choice for plasma facing components in a fusion reactor.…”

Section: Introductionmentioning

confidence: 99%