Аңдатпа
The article presents analytical results on the estimation of the steady-state flux of tritium ions released from lithium-based ceramics Li 2 TiO 3 under reactor irradiation conditions at the WWR-K reactor. Neutron transport calculations were performed in MCNP6 using the ENDF/B-VII.1 library and compared with in-situ gas release recorded at a ceramic temperature of 650 ℃ and a thermal neutron flux of 5·10 13 n/(cm 2 ·s). The developed calculation model does not contain fitting coefficients and takes into account the ion mean free path, the self-shielding effect in ceramics (δ = 0.32) and recombination losses. The obtained values of T⁺ flux from ceramics of different isotopic composition are 3.07·10 −13 to 1.96·10 −12 mol/s. The maximum decrease in the flux due to self-shielding in ceramics does not exceed 32%. The discrepancy between the calculation and the experiment for the HT component does not exceed 15%, which confirms the adequacy of the approach. The results are of practical importance in assessing the tritium balance in fusion reactor blankets and allow optimizing the composition and size of ceramic pebbles taking into account the activation-free gas evolution of tritium.