The ITER project development has shown that considerable difficulties are encountered when already known engineering solutions and materials are used for divertor and divertor plates for tokamaks of such a scale. We offer to use a Li capillary-pore system (CPS) as a plasma facing material for tokamak divertor. Evaporated Li serves as a gas target and redistributes thermal load. The heat flux from the plasma is transported to the first wall by Li radiation in the plasma periphery. This allows the divertor plate to reduce the heat flux. A solid CPS filled with liquid lithium has a high resistance to surface damage in the stationary mode and during plasma transitions (disruptions, ELMs, VDEs, runaways) to assure normal operation of the divertor target plates. These materials are not the sources of impurities giving rise to Z eff and they will not be collected as dust in the divertor area and in ducts.Experiments with lithium CPS in a steady-state mode (up to 25 MW m −2 ) and in plasma disruption simulation conditions (∼5 MJ m −2 , ∼0.5 ms) have been performed. High stability of these systems have been shown. Li limiter tests on T-11M tokamak have revealed the lithium CPS compatibility with the edge plasma for energy loads of up to 10 MW m −2 . In a stable discharge mode at lithium limiter temperature of 20-600˚C, no Li abnormal erosion and injection to plasma have been detected. A high sorption of D + and H + ions on the vessel walls was the main substantial result of the replacement of a graphite limiter by lithium one. He and D sorption was terminated by wall heating up to 50-100˚C and above 350˚C, respectively. T-11 tests on helium discharge allowed to reduce limiter heat load by a factor of two due to lithium radiation.All the experimental results have shown considerable progress in the development of lithium divertor.

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To date there is no adequate solution for high heat load plasma facing components of the next step fusion reactor among solid material options. A lithium-filled capillary porous systems (CPS) was proposed as a plasma facing material and experimental work on this subject is now in progress. Steady-state experiments with CPS-based target and lithium supply systems have shown successful operation at heat fluxes of 1-10 MW/m 2 during several hours. Experimental data is obtained on lithium CPS stability at heat flux up to 25-50 MW/m 2 . The lithium CPS behaviour in contact with real tokamak plasma is considered for normal discharge condition at 10 MW/m 2 and for plasma disruption at 15 MJ/m 2 . Erosion mechanism of lithium under tokamak plasma impact was analysed. Stability of lithium CPS in tokamak conditions was shown.

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