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"The Li-CPS Plasma Face Components for steady-state tokamak experiment“
S. Mirnov
TRINITI Troitsk Moscow Reg.2003
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LI7 characteristics Specific weight (g/cm3) - 0,5 Electric conductivity (ohm cm)-1 - 2,2 104
Heat conductivity (W/cm grade) - 0,43 (1830С) - 0,53 (6270С) Heat capacity (kal/cm3grade) - 0,5 Melting temperature - 180,50С Evaporation temperature - 13170С Evaporation heat (eV/at) - 1,5 Energetic “price” of Li+++ ionized by electrons with Те=30eV (eV/at) -1200 Temperature range of D, T recovery -320-5000С
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Tritium recovery
V.A. Evtikhin et al. Fus. Eng. Design 56-57 (2001),p.363 M. J. Baldwin et al. 43 An. Meeting APS, Divis. Plasma Phys.(2001) Rep. N
CP1 25. Y. Furuyama et al. 15 PSI Conf. 2002 J. Nucl. Mat. 313-316 p.288
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Li Capillary Porous Structures
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Li Divertor Concept
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T- 11M Li- experiment
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The radial radiation distribution in experiments with C and Li limiters
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Lithium flux (erosion) from 200 to 650 C
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Summary (1)
1.Lithium, as lowZ-material, is compatible with tokamak plasma (TFTR, T-11M, CDXU).
2.The surface tension forces in CPS may be used to solve the problem of ponderomotive forces (splashing suppress) and regeneration problem of PFC.
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Summary (2)3.Experiments with hydrogen (deuterium) and helium plasmas on T-11M tokamak with Li – CPS limiter have shown:-No spontaneous bursts of lithium ejection under heat flux to limiter at the level about 10 MW/m2 have been observed. -Total lithium erosion close to level of hydrogen and lithium ions sputtering has been measured.- The lithium radiation protected the limiter from high power load during disruptions. -The solid basis of CPS limiter had no damages after more than 2103 of plasma shots. -The recovery temperature of hydrogen isotopes from Li is 320-500oC (for helium 50-100oC). Therefore, at high PFC temperatures (4000-5000 C) a tritium capture can be minimized. -It should be provided, that separation of helium and hydrogen isotopes will be possible in lithium circuit with lower PFC temperatures.
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Summary (3)
4.These results are making a convincing basis for the advance of the liquid lithium PFC for steady state tokamak. The following problems of such tokamak might be decided:- wall and divertor plates erosion,- “dust” accumulation and redeposition,- tritium recovery,- low Zeff(0),- heat removal in stationary conditions and during disruptions.
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Summary (4)
5.The progress of the considered approach needs further experimental, calculation, design research and technological developments. The following studies seem necessary:- experiments with lithium in divertor tokamaks,- calculations of lithium behavior in divertor region and SOL on the basis of existing codes, including evaporation, condensation, ionization processes etc.