[en] The description of algorithm to design in-pipe experimental ampoule devices (IPAD) is presented here, including description of IPAD design for irradiation tests of highly enriched lithium ceramics at WWR-K reactor. The description of the system for registration of tritium release from ceramics during irradiation is presented as well. Typical curve of tritium release from the IPAD during irradiation under various temperatures of the samples is shown here

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[en] Full text of publication follows: Lithium ceramics is planned to be used in tritium breeding systems of future fusion reactors. To provide effective tritium generation while obeying the ecological and safety restrictions on tritium processing it is necessary to investigate tritium interaction with elements of proposed breeding systems. Therefore tritium-ceramics interaction is of most interest in such systems. Presented work describes experimental studies of tritium yield from lithium ceramics (Li₂TiO₃+5 mol.% TiO₂ ) after long-term neutron irradiation. Initially ceramics was 96% enriched with Li⁶ and irradiated with neutrons (about 220 days) in research water-water reactor of Kazakh National Nuclear Center (WWRK) till the 20% burn-up of Li⁶. Examinations of residual tritium yield from irradiated lithium ceramics were conducted using thermodesorption method with linear heating rates from 2 to 10 K/min up to ceramics melting point temperature. The experiments were carried-out under continuous pump-out and mass-analysis of desorbed gases in experimental chamber. As the result the data on tritium (and other gases) release rates from irradiated ceramics are obtained. Preliminary results on estimations of residual tritium content in irradiated lithium ceramics and its thermodesorption data are presented in given report. (authors)

[en] Vanadium alloys are most promising materials being considered for lithium blanket-breeder in future fusion reactors. The primary reason for these stems from good combination of physical-mechanical and radiation properties of vanadium alloys. In operational conditions of fusion reactors the very important issue is behavior of vanadium alloy with respect to hydrogen isotopes under neutron and gamma irradiation. This paper shows results of the experimental studies of reactor irradiation influence on parameters of hydrogen release from vanadium alloys. Experiments were carried out for various levels of reactor irradiation and showed the effect of irradiation on parameters of hydrogen release from vanadium alloy V4Cr4Ti

[en] Highlights: • 819 GBq/mg of ¹⁷⁷Lu specific activity would be produced at the WWR-K reactor at the thermal neutron flux of 1.2×10¹⁴ cm^-2s^-1. • The activity of the parasitic radioisotope Lu-177m during this irradiation time is about 2.5% of Lu-177 activity. • K-factor is 1.5 the best match corresponds to experimental data. Production of lutetium-177 using direct nuclear reaction ¹⁷⁶Lu(n,γ)¹⁷⁷Lu by WWR-K reactor neutrons on enriched LuCl₃ (up to 82% of ¹⁷⁶Lu) is described. Calculations were performed by MCNP6 transport code. Two different irradiation positions of the WWR-K research reactor were considered. Estimates of the maximum specific activity of the luthetium-177 are obtained for the reactor irradiation positions located: (a) in the reactor core centre, (b) in the core periphery. In these positions, thermal neutron flux is two times different. Experimental data was shown that k-factor is 1.5 for considered irradiation positions. The study shows that for the position located in the core center, the estimated maximum specific activity of lutetium-177 is 819 GBq/mg, is to be achieved after 15 days of irradiation. For the position located in the core periphery, specific activity of lutetium-177 is 561 GBq/mg, is to be achieved after 20 days of irradiation. Ratio of Lu-177m to Lu-177 specific activity is not more than 0.025 for both irradiation positions.