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One of the promising intermetallic compounds for use in nuclear and fusion reactors, as well as in hydrogen energy technology, are intermetallic compounds of beryllium with metals such as Ti, V, Zr and Nb. Beryllium-based intermetallics are not only a promising material for blankets of future fusion reactors, but can also be utilized in other areas of the nuclear industry, including fission reactor reflectors, rocket and space technology. The interest in studies of the interaction of hydrogen isotopes with beryllides is related to the accumulation of tritium and helium in the material, formed as a result of nuclear reactions under neutron irradiation, and is caused by the need for understanding the processes arising from such interaction.In this work, hydrogen sorption and desorption processes of zirconium beryllide ZrBe2 produced by industrial technologies at JSC “Ulba Metallurgical Plant” (JSC UMP) were investigated. The experiments were performed by the Sievert’s and thermal desorption spectrometry (TDS) methods. In TDS experiments deuterium was chosen for samples saturation to reduce the possible error in determining fluxes associated with the release of hydrogen from the elements of the vacuum chamber.An equation for hydrogen solubility in zirconium beryllide ZrBe2 was obtained by processing the experimental data obtained by the Sievert’s method:S=0.46±0.1molm3∙Pa12∙exp-13.3±0.5kJmolRTThe temperature intervals of formation and decomposition of two different hydride phases (deuterides) of zirconium beryllide ZrBe2 at ∼ 600 K and ∼ 900 K have been established in TDS experiments.

This paper presents the results of 4 reactor campaigns on the irradiation of biphasic lithium ceramics containing different ratios of lithium orthosilicate (LOS) and lithium metatitanate (LMT) components (25 and 35 mol% LMT in LOS). The size distribution of pebbles in pebble beds was 250–1250 μm and 500–710 μm, respectively. The studies were carried out sequentially with each type of ceramics. In experiments carried out using the vacuum extraction method, the composition of gases released from lithium ceramic samples was registered in in-situ mode. The absence of purge gas during the experiments minimized the possibility of T2O and HTO formation, significantly simplifying processing of the results and providing more opportunities for results analysis. The main goal of the present paper was to identify the general patterns of the release of gases with mass numbers M2 (H2), M4 (He + HT), M6 (T2) and M18 (H2O) from ceramic samples throughout the entire irradiation experiment in 4 campaigns. Release trends of main gases with mass numbers M2, M4, M6 and M18 for all four campaigns are presented and their comparative analysis was performed. The average partial pressure of tritium release in the form of T2 and HT molecules for all campaigns was determined as 5.8 × 10-7 Torr. The dependences of formation rates of helium release peaks on the irradiation time were plotted. The nature of peak emissions does not have a monotonic relationship; upon irradiation, both an increase in the frequency of peaks and a decrease in it are observed. During irradiation, the process of peak helium release does not stop.The simulation was carried out by the finite element method, assuming that tritium release from the sample is determined by diffusion and desorption processes from the sample surface. The experimental curves are satisfactorily described by a number of sets of desorption and diffusion parameters. One way or another they lie in the range of specified values. The Arrhenius dependences of the effective diffusion coefficient and desorption coefficient obtained for lithium ceramics 35 LMT are equal to:D=5,2×10-11m2sexp-21kJmoleRT,K=1.21×10-4m2sexp-64kJmoleRT.The values of the effective diffusion coefficient and tritium desorption coefficient in 25 LMT ceramics were 15 and 20% lower than in 35 LMT ceramics.

Selection of materials for fusion reactors and, in particular, for blanket systems is one of the most important tasks that determine the development of fusion energy. Lithium ceramics Li2TiO3 is considered as prospective candidate for solid breeders of future fusion reactors' blankets. The paper presents the results of high-temperature tests of Li2TiO3 + 5 mol% TiO2 ceramics, performed under conditions of blowing with argon containing impurities of water vapor of different isotopic compositions (H2O, HDO, D2O) in the range of temperature from 100 °C to 1400 °C. The vapor–gas mixture for the experiment was prepared in a special mixing tank. The interaction of Li2TiO3 + 5 mol% TiO2 ceramics with chemically active gases and water vapors was investigated by thermogravimetry (TG), differential scanning calorimetry (DSC) and mass spectrometry (MS) methods. As a result of experiment, the dependences of the sample mass change during heating under single purge gas composition was obtained. Mass spectra that characterize the changes in the qualitative composition of the gas mixture in the reaction chamber with the investigated sample were recorded and subsequently analyzed. An isotope effect is observed during the experiment: the release of D2 gas begins significantly earlier than the release of HD gas. The observed effect appears only in the high-temperature region (above 1050 °C), which characterizes the investigated material as almost completely chemically neutral towards water vapor.

Biphasic lithium ceramics based on lithium orthosilicate Li4SiO4 and lithium metatitanate Li2TiO3 is one of the most promising materials for breeder blankets of future fusion reactors. One of the important issues of biphasic lithium ceramics application in the fusion reactor blanket is to determine the parameters and mechanisms of tritium transfer within and from the ceramics.This paper continues the analysis of irradiation experiments carried out at the WWR-K reactor (Almaty, Kazakhstan) with a sample of biphasic lithium ceramics Li4SiO4-Li2TiO3 (pebbles of lithium orthosilicate with 35 mol% lithium metatitanate with diameter 250––1250 μm). The section of the experiment in which the reactor was temporarily shutdown for 1.5 h was investigated in detail. During this period of time the sample temperature rapidly decreased from 665 °C to 100 °C, generation of tritium and helium in the lithium ceramic sample ceased, but the desorption of previously generated gases from the ceramic surface continued. The experiments were carried out by the vacuum extraction method.The nature of tritium-containing molecules and helium release for that specified time interval was analyzed. The kinetics of tritium release from ceramics in the experiment during reactor shutdown was simulated and the expression for the effective diffusion coefficient D = 5e-11(m2/s)∙exp(-20(kJ/mole)/RT) was determined. It was suggested that one the most realistic mechanisms for tritium release is the mechanism associated with both diffusion and desorption of tritium from the pebbles surface and release from the open pores of the pebble.This mode of the experiment made it possible to estimate the parameters of tritium release immediately after irradiation, which imitates the conditions of breeding blanket operation in the fusion reactor.

Beryllides have a number of unique properties that make them promising for use in various fields of industry and energy. Since some of the beryllides have the potential to dissolve large amounts of hydrogen (in some cases forming hydride phases), there is some interest in evaluating their possible applications in hydrogen technologies. This paper presents the results of studies of changes in the phase composition of the zirconium beryllide deuteride compound during its linear heating. The studies were carried out using a complex method including thermal gravimetrical analysis (TGA), differential scanning calorimetry (DSC) and mass spectrometry. Based on the results of the experiments, a number of parameters were obtained characterizing the transitions of deuteride from the phase of the Orthorhombic I _mma crystal structure of deuteride ZrBe _2 D into the Orthorhombic C _mcm crystal structure of deuteride ZrBe _2 D _0.67 , as well as the decomposition parameters of Orthorhombic C _mcm crystal structure of ZrBe _2 D _0.67 . The energies of endothermic reactions of the corresponding transitions and decompositions of deuteride phases were estimated, and the activation energy of the interaction of water vapor with the surface of zirconium beryllide samples was also estimated.

One of the important areas of research in the energy sector is the study of the prospects for using new types of nuclear fuel, including tritium, which is one of the most promising types of fuel for thermonuclear energy. At the same time, for the production of tritium in the required quantities, the one that is the most optimal is the use of blanket materials based on lithium-containing ceramics. This is where tritium is released from lithium under the influence of neutron irradiation. The paper presents the results of an investigation of the influence of two-phase ceramics based on Li4SiO4–Li2TiO3 compounds on the resistance to external influences (mechanical loads) during the accumulation of hydrogen and helium (He2+) in the near-surface layer. The interest in such studies is primarily related to the search for solutions in the field of creating high-strength materials for tritium generation for its further use as nuclear fuel for thermonuclear fusion, as well as to the study of the mechanisms of the influence of different phases on the changes in the strength properties of ceramics, which provides an opportunity to expand fundamental knowledge in this area. The proposed method of obtaining two-phase ceramics by mechanical-chemical mixing and subsequent sintering into spherical particles enables the production of well-structured, high-strength ceramics of specified geometric dimensions (limited only by the dimensions of the mold) with a controlled phase ratio. During the experiments, it was found that increasing the content of Li4SiO4 phase in ceramics leads to an increase in strength characteristics (hardness, resistance to cracking) by 15–20% compared to single-phase ceramics. The most optimal composition of two-phase ceramics with high resistance to destructive embrittlement is the ratio of phases 0.75Li4SiO4–0.25Li2TiO3. One of the factors explaining the increase in resistance to destructive embrittlement under high-dose irradiation for two-phase ceramics is the increased dislocation density and the presence of interphase or intergranular boundaries, the high concentration of which leads to the creation of additional obstacles to the agglomeration of hydrogen and helium in the near-surface layer.

This work presents the preliminary experimental data on the study of gas release from two-phase lithium ceramics Li4SiO4-Li2TiO3 under neutron irradiation conditions. Experiments were carried out at the WWR-K research reactor (Almaty, Kazakhstan) for ∼4.3 days. The total neutron fluence during the irradiation was ∼1.8·1019n/cm2. In the course of irradiation, two experiments on ceramics heating during irradiation and two experiments with hydrogen isotopes (H2 and D2) supply into the experimental chamber with the sample were performed at a temperature of 680 °C and reactor power of 6 MW. During the entire irradiation, the gas composition in the continuously evacuated ampoule device with samples was recorded. The obtained dependences of the release of tritium-containing molecules and helium during the experiment were qualitatively analyzed.

In the development of fusion energy, an important task is the study and improvement of tritium production technologies. In this case, one of the most promising materials for tritium generation is lithium ceramics. Considering the importance of the task, numerous studies are aimed at solving the problem of determining the parameters and mechanisms of tritium release in lithium-containing materials. This paper presents the results of a study of tritium release processes from two-phase lithium ceramics of Li4SiO4/Li2TiO3 during reactor irradiation when hydrogen and deuterium are injected into the chamber with irradiated samples. The mechanisms regularities of the tritium yield process in the presence of these isotopes were established. The experiments were carried out in the WWR-K research reactor at a neutron flux density of 5∙1013n/cm2∙s and sample temperatures from 650 to 700 °C.

This paper analyzes part of the reactor experiments on the study of tritium and helium release from promising two-phase lithium ceramic (Li2TiO3 and Li4SiO4) of natural lithium enrichment conducted by vacuum extraction.The basis for such an analysis was a more careful study of the time trend of pressure changes of gases in the chamber with the test samples. In a particular case, it was clearly shown that the pressure fluctuations observed during irradiation for gases with mass number M4 (to which both HT and He molecules correspond) are determined only by He, which leaves the intergranular regions of the ceramic through open channels or cracks. The kinetics of changes in the amount of helium that is released during irradiation was traced and both the rate of helium release and the frequency of emissions were determined. It was assumed that the observed emissions correspond to a certain “formation of free paths” from the internal cavities of the irradiated ceramics into the chamber of the facility. The data obtained for the helium emissions were compared with the release of tritium-containing molecules from the ceramics. The quasi-equilibrium levels of the release of tritium-containing molecules and their dependence on the reactor power were estimated. The release of helium and tritium was compared with the calculated values of the tritium generation rate in the test sample.

The relevance of the study is caused by the necessity to develop methodical and instrumental bases for researching hydrogen isotope interaction with different materials under reactor irradiation. The main aim of the study is to develop and to test the technique on determining hydrogen isotope interaction parameters with structural materials under reactor irradiation; to carry out the estimates and experiments confirming the possibility of using the technique. The methods used in the study: thermophysical, neutron calculations methods of ampoule devices and method of hydrogen permeation were used. The results: The paper introduces the principle concepts of material reactor investigation technique using the hydrogen permeation method, estimates of the technique applicability and the results of its application at the reactor experiments with structural materials, such as steel SS316L (N)-IG, vanadium alloy V4Cr4Ti and copper bronze CuCrZr.