Plasma–lithium interaction in the CDX-U spherical torus

Antar, G.; Doerner, R. P.; Kaita, R.; Majeski, R.; Spaleta, J.; Munsat, T.; Maingi, R.; Soukhanovskii, V.; Kugel, H.; Timberlake, J.; Krasheninnikov, S. I.; Luckhardt, S.; Conn, R.W.

doi:10.1016/s0920-3796(02)00055-8

Cited by 35 publications

(12 citation statements)

References 17 publications

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“…The observation was done along a tangential viewing, allowing for the observation of a portion of the poloidal plane where the jet injection is performed. The same phenomenon has been observed in CDX-U device with lithium [3]. In our case, the source of the droplet was most certainly the collector since it is seen to travel upward from the lower port of the tokamak vessel.…”

supporting

confidence: 84%

Fast Camera Images of Surviving Gallium Droplets During Plasma–Liquid Gallium Jet Interaction Experiments in ISTTOK

Gomes¹,

Fernandes²,

Silva³

et al. 2008

IEEE Trans. Plasma Sci.

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Liquid metals have received a renewed interest in fusion research due to the advantages that these materials could offer to overcome the erosion problem of plasma facing components presently used in large-size fusion devices. In ISTTOK Tokamak, the plasma-liquid gallium jet interaction is currently being studied. One of the main technical problems faced during the implementation of the experimental setup was to minimize the amount of gallium that remains in the vacuum chamber. Despite the large effort put in fulfilling this task, liquid-metal droplets could not be completely eradicated from the plasma. Here, pictures of one of those remaining droplets traveling through the plasma are presented and discussed. Index Terms-Gallium, limiter, liquid metals, liquid-metal jet, plasma-surface interaction.M ATERIALS currently used in large-size fusion devices are submitted to very high power loads. A frequent replacement of the plasma facing components is to be expected in the present approach of the fusion research program as a result of the high erosion levels and thermal stress produced by such power loads, and one way to overcome that issue is linked to the use of liquid metals as plasma facing surfaces. The possibility to perform a permanent renewal of liquid surfaces has been pointed out as one adequate solution for the protection of solid walls providing, at the same time, an efficient process to exhaust power from fusion plasmas. Among a set of several liquid metals, lithium has shown the best compatibility with fusion plasmas, which is mainly due to its low Z [1]. Another candidate material that has better thermal properties, a wider liquid state range, and a lower vapor pressure is gallium. The interaction of a liquid gallium jet with plasmas is being studied in ISTTOK, which is a tokamak with R = 0.46 m, a = 0.085 m, B T = 0.45 T, n e (0) = 5 × 10 18 m −3 , T e (0) = 150 eV, I P ∼ 6 kA, and V loop ∼4 V. The jets, which have a diameter of 2.3 mm and a flow velocity of 2.5 m/s, are generated by hydrostatic pressure and received in a lower collector after interacting with the plasma. A detailed description of the experimental setup is presented in [2].Due to its high density, gallium flows have typically high Weber number (in our specific case, We∼140 [2]). One of the consequences of this effect is the splashing of gallium droplets Manuscript

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supporting

confidence: 84%

Fast Camera Images of Surviving Gallium Droplets During Plasma–Liquid Gallium Jet Interaction Experiments in ISTTOK

Gomes¹,

Fernandes²,

Silva³

et al. 2008

IEEE Trans. Plasma Sci.

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“…Exceptional properties of pumping hydrogen plasma particles by lithium have been observed in tokamaks during the very first experiments involving a large area of lithium coated wall surface on T-11 [17][18][19][20] and with rail [21] and toroidal liquid lithium limiter [22,23] on CDX-U spherical tokamak.…”

Section: Particle and Power Extraction By Close-fitting Liwallsmentioning

confidence: 99%

Ignited spherical tokamaks and plasma regimes with LiWalls

Zakharov

Gorelenkov

White

et al. 2004

Fusion Engineering and Design

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“…[6] Following the rail limiter experiments, a shallow, heated, stainless steel tray was installed at the bottom of the CDX-U vacuum vessel. The tray has an inner radius of 24 cm, is 10 cm wide and 0.5 cm deep, and exposes 2000 cm 2 of lithium to the plasma.…”

Section: Ex/p5-23mentioning

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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Plasma–lithium interaction in the CDX-U spherical torus

Cited by 35 publications

References 17 publications

Fast Camera Images of Surviving Gallium Droplets During Plasma–Liquid Gallium Jet Interaction Experiments in ISTTOK

Fast Camera Images of Surviving Gallium Droplets During Plasma–Liquid Gallium Jet Interaction Experiments in ISTTOK

Ignited spherical tokamaks and plasma regimes with LiWalls

Liquid Lithium Limiter Experiments in CDX-U

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