Li-CPS limiter in tokamak T-11M

Мирнов, С. В.; Lazarev, V.B.; Sotnikov, S. M.; Evtikhin, V.A.; Lyublinski, I.E.; Vertkov, А.V.

doi:10.1016/s0920-3796(03)00018-8

Cited by 60 publications

(51 citation statements)

References 13 publications

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“…Here, however, ELM disappearance was studied in more detail and with more diagnostics. Our more general observation of confinement improvement with lithium has also been reported in other devices [12][13][14][15], whereas the use of lithium as a PFC material has also been discussed elsewhere [16,17].…”

Section: Introductionsupporting

confidence: 78%

Transition to ELM-free improved H-mode by lithium deposition on NSTX graphite divertor surfaces

Mansfield

Kugel

Maingi

et al. 2009

Journal of Nuclear Materials

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a b s t r a c tLithium evaporated onto plasma facing components in the NSTX lower divertor has made dramatic improvements in discharge performance. As lithium accumulated, plasmas previously exhibiting robust Type 1 ELMs gradually transformed into discharges with intermittent ELMs and finally into continuously evolving ELM-free discharges. During this sequence, other discharge parameters changed in a complicated manner. As the ELMs disappeared, energy confinement improved and remarkable changes in edge and scrape-off layer plasma properties were observed. These results demonstrate that active modification of plasma surface interactions can preempt large ELMs.

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Section: Introductionsupporting

confidence: 78%

Transition to ELM-free improved H-mode by lithium deposition on NSTX graphite divertor surfaces

Mansfield

Kugel

Maingi

et al. 2009

Journal of Nuclear Materials

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“…Lithium limiter experiments have also been performed on the T-11M device, [4] where a Capillary Porous System (CPS) was used to form a "self-restoring" liquid lithium limiter surface. [5].…”

Section: Introductionmentioning

confidence: 99%

Testing of liquid lithium limiters in CDX-U

Majeski

Kaita

Boaz

et al. 2004

Fusion Engineering and Design

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showed some reduction of impurities in CDX-U plasmas with the L3, compared to discharges with a boron carbide limiter. While no reduction in recycling was observed with the L3, which had a plasma-wet area of approximately 40 cm 2 , subsequent experiments with a larger area fully toroidal lithium limiter demonstrated significant reductions in both recycling and in impurity levels. Two series of experiments with the toroidal limiter have now been performed. In each series, the area of exposed, clean lithium was increased, until in the latest experiments the liquid lithium plasma-facing area was increased to 2000 cm 2 . Under these conditions, the reduction in recycling required a factor of eight increase in gas fueling in order to maintain the plasma density.The loop voltage required to sustain the plasma current was reduced from 2V to 0.5V. This paper summarizes the technical preparations for lithium experiments and the conditioning required to prepare the lithium surface for plasma operations. The mechanical response of the liquid metal to induced currents, especially through contact with the plasma, is discussed. The effect of the lithium-filled toroidal limiter on plasma performance is also briefly described.-2 -

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“…21 The first experiment with a liquid, rather than a solid, lithium limiter was conducted on the T11-M device, starting in the late 1990s. 28 The T11-M capillary porous (CPS) system employed multiple layers of molybdenum mesh to retain lithium via wetting and surface tension against motion induced by plasma-generated MHD forces on the liquid metal. A system based on the T11-M development, but larger and more extensive, has been more recently employed in the FTU tokamak, starting in 2006.…”

Section: Lithium Wall Conditioning and Liquid Lithium Pfc Experimentsmentioning

confidence: 99%

Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

Majeski¹

2010

AIP Conference Proceedings

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Abstract. At present, the only solid material believed to be a viable option for plasma-facing components (PFCs) in a fusion reactor is tungsten. Operated at the lower temperatures typical of present-day fusion experiments, tungsten is known to suffer from surface degradation during long-term exposure to helium-containing plasmas, leading to reduced thermal conduction to the bulk, and enhanced erosion. Existing alloys are also quite brittle at temperatures under 700°C. However, at a sufficiently high operating temperature (700 -1000 °C), tungsten is selfannealing and it is expected that surface damage will be reduced to the point where tungsten PFCs will have an acceptable lifetime in a reactor environment.The existence of only one potentially viable option for solid PFCs, though, constitutes one of the most significant restrictions on design space for DEMO and follow-on fusion reactors. In contrast, there are several candidates for liquid metal-based PFCs, including gallium, tin, lithium, and tin-lithium eutectics. We will discuss options for liquid metal walls in tokamaks, looking at both high and low recycling materials. We will then focus in particular on one of the candidate liquids, lithium.Lithium is known to have a high chemical affinity for hydrogen, and has been shown in test stands 1 and fusion experiments 2,3 to produce a low recycling surface, especially when liquid. Because it is also low-Z and is usable in a tokamak over a reasonable temperature range (200 -400 °C), it has been now been used as a PFC in several confinement experiments (TFTR, T11-M, CDX-U, NSTX, FTU, and TJ-II), with favorable results. The consequences of substituting low recycling walls for the traditional high recycling variety on tokamak equilibria are very extensive. We will discuss some of the expected modifications, briefly reviewing experimental results, and comparing the results to expectations.

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Li-CPS limiter in tokamak T-11M

Cited by 60 publications

References 13 publications

Transition to ELM-free improved H-mode by lithium deposition on NSTX graphite divertor surfaces

Transition to ELM-free improved H-mode by lithium deposition on NSTX graphite divertor surfaces

Testing of liquid lithium limiters in CDX-U

Liquid Metal Walls, Lithium, And Low Recycling Boundary Conditions In Tokamaks

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