This paper presents the results of an experiment on high-temperature oxidation of SiC coating of a prototype HTGR fuel in an oxygen environment under the following experimental conditions: the maximum temperature of the samples with linear heating is 1400 °C; heating rate – 10 °С/min. According to the results of the experiments, dependences of the change in the composition of the gas mixture in the corrosion chamber with linear heating of the sample were obtained. A qualitative analysis of corrosion data and microstructural research data was carried out before and after the corrosion experiment. The results obtained allow us to estimate the corrosion resistance and the loss of the operational properties of SiC coatings on prototype HTGR fuel.

V-Ti-Cr vanadium alloys have many perspective applications, particularly, they are considered as a main structural material of the first wall and blanket of fusion reactor with lithium coolant. Also, they are used as a high-temperature membrane for hydrogen isotopes separation and effective sorbents for hydrogen purification installations by PSA (Pressure Swing Adsorption) adsorption method (under the control of sorbent surface composition). The use of vanadium alloys can be extremely interesting because of the possibility of hydrogen isotopes separation in hydrogen systems. The fact is that vanadium has unique properties with respect to hydrogen: protium atoms typically occupy octahedral interstices in vanadium bcc lattice, while the deuterium atoms - mainly tetrahedral interstices. That all considerably affects diffusion and solubility of hydrogen isotopes in vanadium. This paper presents the results of experiments on sorption of hydrogen isotopes with V4Cr4Ti vanadium alloy from a mixture of hydrogen isotopes. According to the obtained time dependence of hydrogen isotopes partial pressure change in gas mixture the following parameters were determined: rate of dissolution and heat of dissolution of hydrogen and deuterium in vanadium alloys V4Cr4Ti, V10Cr5Ti. The conclusion about the prospects of using vanadium alloys in storage systems, separation and purification was made.

This paper presents the results of experiments on hydrogen isotopes sorption with V4Cr4Ti vanadium alloys from a mixture of hydrogen isotopes. The studies were carried out at temperatures of 353 K, 393 K, 423 K; and pressures of 10(3)-10(4) Pa in gas mixture of hydrogen isotopes. The alpha-phase domain of V-H (D) system was studied, where the concentration of hydrogen isotopes atoms should not exceed 0.015H (D) atoms per metal atom. The separation parameters were derived for several saturation conditions accordingly to registered time dependences of hydrogen isotopes partial pressure drop.The conclusion was made about the prospects of using vanadium alloys in hydrogen isotopes separation and purification systems. (C) 2016 Elsevier B.V. All rights reserved.

The work presents study of helium and tritium interaction with irradiated S-200F beryllium grade. The goal of research is related to solving the problem of detritiation of stored beryllium material. Also it is aimed to explain the mechanisms of processes occurring during tritium interaction with beryl-lium which is planned to be used in the future fusion-type reactor (FTR). The study consists of following stages: thermodesorption xperiments on gas release from irradiated beryllium, calculation of helium and tritium generated in beryllium; processing and analysis of experimental data; determination of basic pa-rameters of tritium interaction with irradiated beryllium. These data is needed for estimation of high-temperature degassing procedure such as detritiation method in case of S-200F eryllium grade as well as for calculation of tritium distribution in beryllium materials.

Corrosion in air and water remains a problem for graphite as an important structural material of nuclear reactors. Changes in the gas composition in corrosion chamber with samples of RGT graphite have been studied. The sample was linearly heated to 1100C with 10C /min rate, and held at this temperature for 1800 s. Then the sample was cooled down within the same rate. During the experiment changes in the gas composition in the chamber were registered within the mass range 1-70. The time and temperature dependencies of steam, oxygen and gaseous corrosion products of RGT graphite have been obtained. Arrhenius dependencies and activation energy of rate constants of graphite interaction with the following products were derived: oxygen-graphite (mole/(sm2Pa)),water-graphite (mole/(sm2Pa).

Proposed mathematical and software analysis of reactor experiments allowed interpretation of the experimental results of a tritium release study. Tritium was continuously generated by the reaction of lithium-6 with thermal neutrons for various thermal conditions of lithium metatitanate (Li2TiO3). The main gas release parameters were calculated in order to assess the potential use of lithium metatitanate in tritium breeders. These parameters were: gas release rate, tritium retention, retention time, activation energy for thermal desorption as HT, activation energy for volume diffusion as Т+, and the corresponding pre-exponential (frequency) indexes.

ДИФФУЗИЯ ТРИТИЯ, ГЕНЕРИРУЕМОГО В МЕТАТИТАНАТЕ ЛИТИЯ Li 2 TiO 3 В ПРОЦЕССЕ ОБЛУЧЕНИЯ ТЕПЛОВЫМИ НЕЙТРОНАМИ В РЕАКТОРЕ ВВР-КИ.Н.Бекман, И

The paper describes the results of fusion reactor materials R&D activities in the Republic of Kazakhstan in the framework of the fusion material national program and the results of the activities on construction of material testing spherical tokamak in the RK.

Given work presents the results of complex material-science studies of 1mm diameter ceramic pebbles manufactured of Li2TiO3+5mol% TiO2 ceramics before and after long-time neutron irradiation. Ceramic samples were placed in specially ampoules (six items) made of stainless steel Cr18Ni10Ti which were vacuumized and filled with helium. Irradiation of ampoules was carried out in the loop channel of WWRK reactor (Almaty, Kazakhstan) during 223 days at 6MW power. After irradiation light-colored pebbles became grey-colored due to structure changes which generation of grey-colored inclusions (lithium oxide) with low density and microhardness. There is a radiation softening of lithium ceramic and that effect is higher for lower irradiation temperature 760K than for 920K. The value of maximum permissible load (pebble crash limit) at that is low and comprises ∼37.9N. The content of residual tritium is higher for ceramic irradiated at 760K (6.6±0.6×1011Bq/kg) than for ceramic irradiated at 920K (17±3×1010Bq/kg). The size change indicates that pebble increase more after irradiation at 760K than at 920K where the bigger portion of tritium leaves the pebble. X-ray analysis shows radiation modification of Li2TiO3+5mol% TiO2 phase composition and generation of new phases: LiTi2O4, LiTiO2 and Li4Ti5O12.

Y. Chikhraya, V. Shestakov b, S. Afanasyev a, A. Yelishenkova, T. Kulsartovc, A. Kuykabayeba c, Y. Kenzhind, I. Tazhibayeva c, I. Beckmane, H. Kawamuraf and K. Tsuchiyag aKazakh National University, Kourmangazy 15, app.10, 480100 Almaty, Kazakhstan bKazakhstan State University, Almaty, Kazakhstan cInstitute of Atomic Energy NNC RK, Institute of Atomic Energy NNC RK, Kurchatov, Kazakhstan dNational Nuclear Center, Kurchatov, Kazakhstan eMoscow State University, Moscow, Russian Federation f JAEA, Oarai, Japan gDirectorates of Fusion Energy Research, JAEA, 4001 Oarai-machi, Higashi-ibaraki-gun, 311 1393 Ibaraki-ken, Japan john@physics.kz

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.

Lithium ceramics is planned to be used in tritium breeding systems of future fusion reactors. To provide e ective 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 (Li 2TiO3+5mol.% TiO2) after long-term neutron irradiation. Initially ceramics was 96% enriched with Li 6 and irradiated with neutrons (about 220 days) in research water-water reactor of Kazakh National Nuclear Center (WWRK) till the 20% burn-up of Li 6 . 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.

A 220-day irradiation of Li2TiO3 ceramics with 96% enrichment of isotope 6Li was carried out at WWRK reactor. One of the study goals was to examine tritium release behavior during Li burn-up. To achieve this goal three types of ceramics samples were examined simultaneously using a system for in-pile tritium monitoring: one (pebbles) – under constant temperature of 650°C, and two (pebbles and pellets) – within temperature change ranges from 500 to 900°C. Flows of tritium release from ceramics during the various reactor campaigns, as well as tritium generation rates for each ampoule are presented in the paper. The main result is justification of burn-up in 6Li reaching up to 20% of Li2TiO3 ceramics.

In-pile experiments on tritium permeation were conducted for ferritic steel (F82H) with and without a ceramic coating of Cr2O3–SiO2 including CrPO4, using a research reactor, IGV. 1M, in Kazakhstan. The tritium source used was liquid lithium-lead eutectics, Pb17Li, which was poured into a space around a tubular diffusion cell used as a specimen of ferritic steel (F82H) with or without the coating on the inner surface of the cell. The irradiation time was about 4h at 6MW reactor power, which corresponds to a fast neuron fluence of about 2×1021n/m2 (E>1.1MeV). The permeation reduction factor (PRF) was obtained by comparison of the kinetics curve of tritium permeation through the diffusion cell of F82H steel with and without the coating. The PRF at 600°C was 292. This value is close to a corresponding PRF value of 307, which had been obtained at 600°C, in a previous out-of-pile experiment.

Ceramic coating on structural materials has been considered to be used as a tritium permeation barrier for fusion power plants. In the present study, a series of hydrogen and deuterium permeation experiments was performed for ferritic F82H steel with and without a ceramic coating of Cr2O3–SiO2 including CrPO4. First, experiments were made on the permeability of F82H steel without coating at 300–600°C in an atmosphere of 100–1000Pa hydrogen and deuterium. The measured values of diffusion coefficient, permeability and solubility were in good agreement with those published previously. Next, experiments were performed on the permeability of F82H steel with the coating at 400–600°C in an atmosphere of 1000–1500Pa deuterium, and the permeation reduction factor (PRF) of the coating was determined. The obtained PRF at 600°C was about 400, which is of the same order of magnitude as the PRF value of about 1000 previously evaluated for the same coating on an SS316 substrate. A significant decrease in the PRF (down to about 4%) was observed, when the sample temperature was lowered from 600 to 400°C.