"The Li-CPS Plasma Face Components for steady-state tokamak experiment“

S. Mirnov

TRINITI Troitsk Moscow Reg.2003

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

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

Li Capillary Porous Structures

Li Divertor Concept

T- 11M Li- experiment

The radial radiation distribution in experiments with C and Li limiters

Lithium flux (erosion) from 200 to 650 C

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.

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.

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.

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.