Experimental and calculated basis of the lithium capillary system as divertor material

Antonov, N. V.; Belan, V. G.; Evtihin, V.A.; Golubchikov, L. G.; Khripunov, V.I.; Korjavin, V.M.; Lyublinski, I.E.; Maynashev, V.S.; Petrov, V.B.; Pistunovich, V.I.; Pozharov, V.A.; Podkovirnov, V.I.; Shapkin, V.V.; Vertkov, А.V.

doi:10.1016/s0022-3115(97)80219-x

Cited by 14 publications

(5 citation statements)

References 3 publications

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“…Lithium limiter experiments have also been performed on the T-11M device, [4] where a capillary porous system was used to form a "self-restoring" liquid lithium limiter surface. [5] CDX-U is a small spherical torus, with a major radius R 0 =34 cm, minor radius a=22 cm, aspect ratio = 1.5, elongation κ = 1.6, toroidal field B T = 2.1 kG, and ohmic current I p ≤ 90 kA. With the exception of the capacitor banks for the OH system and the field null formation coils, the power supplies are preprogrammed and controlled by digital to analog waveform generators.…”

Section: Introductionmentioning

confidence: 99%

Liquid Lithium Limiter Experiments in CDX-U

Majeski

Jardin²,

Kaita³

et al. 2004

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Abstract. Recent experiments in the Current Drive eXperiment -Upgrade provide a first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R=34 cm, a=22 cm, B toroidal = 2 kG, I P =100 kA, T e (0)~100 eV, n e (0)~ 5 × 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium tray limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U. Tokamak discharges which used the liquid lithium limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z effective <1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced.

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Section: Introductionmentioning

confidence: 99%

Liquid Lithium Limiter Experiments in CDX-U

Majeski

Jardin²,

Kaita³

et al. 2004

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“…3.1, the capillary action and high surface tension of LL keeps the CPS porous metal surface replenished, through the capillary action, from the reservoir while LL evaporates under intense heat flux from plasma. In a test stand with an electron beam, the CPS surface was observed to survive a very high heat flux of 50 MW/m 2 for a several seconds [25]. The CPS surface essentially survived the high heat flux as long as the capillary-porous material was kept saturated with LL.…”

Section: Protective Effects Of Lithium As Pfcsmentioning

confidence: 99%

“…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 surface essentially survived the high heat flux as long as the capillary-porous material was kept saturated with LL. With a fully functioning CPS, the incoming main divertor power is radiated in the vicinity of the divertor plates (CPS) by the lithium neutrals/ions, which screen the target, substantially reducing the effective power flow density onto the CPS target [25]. The calculations show that the thermal loading onto the divertor CPS target is essentially reduced without addition of any heavy impurities, at an electron density at the separatrix of 6 x 10 13 cm -4 .…”

Section: Protective Effects Of Lithium As Pfcsmentioning

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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“…Lithium limiter experiments have also been performed on the T-11M device, [4] where a capillary porous rail limiter system was used to form a "selfrestoring" liquid lithium surface. [5] The T11-M limiter is relatively small, and evaporated lithium wall coatings are thought to be a factor in the experiments. [4] In this paper, we focus on experiments in which a substantial fraction of the plasma-facing surface is liquid lithium.…”

Section: Introductionmentioning

confidence: 99%

Recent Liquid Lithium Limiter Experiments in CDX-U

Kaita¹,

Majeski²,

Gray³

et al. 2005

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Abstract. Recent experiments in the Current Drive eXperiment -Upgrade (CDX-U) providea first-ever test of large area liquid lithium surfaces as a tokamak first wall, to gain engineering experience with a liquid metal first wall, and to investigate whether very low recycling plasma regimes can be accessed with lithium walls. The CDX-U is a compact (R=34 cm, a=22 cm, B toroidal = 2 kG, I P =100 kA, T e (0)~100 eV, n e (0)~ 5 10 19 m -3 ) spherical torus at the Princeton Plasma Physics Laboratory. A toroidal liquid lithium pool limiter with an area of 2000 cm 2 (half the total plasma limiting surface) has been installed in CDX-U.Tokamak discharges which used the liquid lithium pool limiter required a fourfold lower loop voltage to sustain the plasma current, and a factor of 5-8 increase in gas fueling to achieve a comparable density, indicating that recycling is strongly reduced. Modeling of the discharges demonstrated that the lithium limited discharges are consistent with Z effective <1.2 (compared to 2.4 for the pre-lithium discharges), a broadened current channel, and a 25% increase in the core electron temperature. Spectroscopic measurements indicate that edge oxygen and carbon radiation are strongly reduced.

show abstract

Experimental and calculated basis of the lithium capillary system as divertor material

Cited by 14 publications

References 3 publications

Liquid Lithium Limiter Experiments in CDX-U

Liquid Lithium Limiter Experiments in CDX-U

Lithium As Plasma Facing Component for Magnetic Fusion Research

Recent Liquid Lithium Limiter Experiments in CDX-U

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