[en] The work was devoted to development of the experimental unit for measurement and comparison of tritium permeation rates through bare and protected martensitic steel tubes. Both out-of-pile deuterium gas driven experiments and in-pile tritium permeation experiments were performed. The permeation reduction factor was estimated to be not less than 50 in both series. A few in-pile experiments of 6 h of duration each were performed in a high flux reactor. No degradation of the protection layer was observed for this period

[en] Reaction of radiative capture of α-particle by ⁷Li nuclei is considered. The reaction leads to excitations of some resonance states of ¹¹B nuclei. Cross-sections for high-energy γ-quanta flight-out are calculated

[en] Beryllides have good properties and are one of the candidate materials for the neutron multiplier of the DEMO reactor in which good performance is required at 600-800 deg. C. Therefore, a good compatibility with structural materials is expected for the beryllides. In this study, a compatibility test between Be₁₂Ti and SS316LN was carried out as first step to evaluate the compatibility between beryllides and structural materials. The thickness of the reaction layer between Be₁₂Ti and SS316LN at 800 deg. C after 1000 h was approximately 30 μm, whereas that of beryllium metal was 300 μm. At 600 deg. C after 1000 h, the thickness of the reaction layers as to Be₁₂Ti and Be was less than 10 and 100 μm, respectively. The compatibility between Be₁₂Ti and SS316LN was evaluated and Be₁₂Ti was perfectly better than that between beryllium metal and SS316LN at high temperature (600-800 deg. C). The advantage of beryllides as a neutron multiplier in the Demo reactor was proved

[en] Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented

[en] Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal

[en] Gas-driven permeation of hydrogen through metal membranes in the surface-limited regime (SLR) is analyzed. An analytical solution for the concentration and permeation flux as a function of time is given for permeation through the asymmetric membrane having different conditions on the inlet and outlet sides. The features of the steady state and transient permeation are discussed. Comparison of calculations with an experiment on deuterium permeation through vanadium, which is available from literature, is performed. It has been demonstrated that the parameters, extracted from the measurements of the permeation rate in SLR are very uncertain. Even the calculations with opposite asymmetries can be equally well adjusted to the experiment. Non-zero initial conditions in SLR experiments are additional source of uncertainty. Measurements of the accumulation in a closed volume instead of the permeation rate can seriously mislead in the interpretation of the experiment. Complementary experiments on permeation in two opposite directions and measurements of permeation decay could increase the reliability of the data obtained

[en] The permeation of deuterium through martensitic steel F82H is investigated in the ranges of 573-873 K and 10²-10³ Pa in two series of experiments with different surface conditions. The experiment is compared with diffusion calculations. The effective values of the diffusion coefficients and solubilities have been obtained. Both the 'permeation' and 'accumulation' curves are analyzed. The later can mislead in judgment about the quality of the experiment and validity of the diffusion approach

No abstract available

[en] Historically in Kazakhstan all disposals of used beryllium and beryllium wasted materials were stored and recycled at JSC ''Ulba Metallurgical Plant''. Since Ulba Metallurgical Plant (beside beryllium and tantalum production) is one of the world largest complex producers of fuel for nuclear power plants as well it has possibilities, technologies and experience in processing toxic and radioactive wastes related with those productions. At present time only one operating Kazakhstan research reactors (EWG1M in Kurchatov) contains beryllium made core. The results of current examination of that core allow using it without replacement long time yet (at least for next five-ten years). Nevertheless the problem how to utilize such irradiated beryllium becomes actual issue for Kazakhstan even today. Since Kazakhstan is the member of ITER/DEMO Reactors Projects and is permanently considered as possible provider of huge amount of beryllium for those reactors so that is the reason for starting studies of possibilities of large scale processing/recycling of such disposed irradiated beryllium. It is clear that the Ulba Metallurgical Plant is considered as the best site for it in Kazakhstan. The draft plan how to organize experimental studies of irradiated beryllium disposals in Kazakhstan involving National Nuclear Center, National University (Almaty), JSC ''Ulba Metallurgical Plant'' (Ust-Kamenogorsk) would be presented in this paper as well as proposals to arrange international collaboration in that field through ISTC (International Science Technology Center, Moscow). (author)

[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