Взаимодействие ионов с поверхностью (ВИП)
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Просмотр Взаимодействие ионов с поверхностью (ВИП) по Автор "Gasparyan, Yu. M."
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- ПубликацияОткрытый доступDEUTERIUM RE-EMISSION AND THERMAL DESORPTION FROM IRON AND EUROFER(НИЯУ МИФИ, 2017) Ryabtsev, S. A.; Gasparyan, Yu. M.; Ogorodnikova, O. V.; Harutyunyan, Z. R.; Pisarev, A. A.; Арутюнян, Зорий Робертович; Огородникова, Ольга Вячеславовна; Писарев, Александр Александрович; Гаспарян, Юрий МикаэловичReduced-activation ferritic-marthensitic (RAFM) steels, such as Eurofer, are considered as candidates for structural materials in fusion reactors due to the high thermal conductivity, the low thermal expansion coefficient and good resistance to radiation swelling. There are also some concepts of fusion reactors, where RAFM steels also considered as material for plasma-facing components. In this regard, the key aspects of hydrogen (H) isotopes interaction with RAFM steels, such as tritium (T) retention and migration in these materials are particularly important as a point of safety concern.
- ПубликацияОткрытый доступDeuterium trapping in co-deposited layers of ITER-relevant materials(НИЯУ МИФИ, 2021) Krat, S. A.; Prishvitsyn, A. S.; Vasina, Ya. A.; Fefelova, E. A.; Gasparyan, Yu. M.; Pisarev, A. A.; Писарев, Александр Александрович; Гаспарян, Юрий Микаэлович; Крат, Степан Андреевич; Пришвицын, Александр СергеевичHydrogen isotope accumulation in fusion devices is a serious problem. Because deuterium-tritium mixture will be a working gas in future fusion devices, including ITER tokamak, tritium accumulation is an issue from the perspective of radiation safety. In total, only 700 grams of tritium are allowed to be present in ITER vessel at any time, with additional 120 in the cryopumps, and 180 grams allocated to measurement error, to the total of 1000 grams.
- ПубликацияОткрытый доступHELIUM THERMAL DESORPTION FROM TUNGSTEN AFTER ION BEAM IRRADIATION AT ELEVATED TEMPERATURES(НИЯУ МИФИ, 2019) Ryabtsev, S. A.; Gasparyan, Yu. M.; Harutyunyan, Z. R.; Efimov, V. S.; Aksenova, A. S.; Pisarev, A. A.; Писарев, Александр Александрович; Ефимов, Виталий Сергеевич; Гаспарян, Юрий Микаэлович; Арутюнян, Зорий РобертовичHelium (He) is a product of deuterium-tritium reaction, so appearance of helium impurities will be unavoidable. In addition to He implantation from fusion plasma, He can be introduced into material by both neutron irradiation and tritium radioactive decay. Presence of He in plasma-facing materials may significantly influence their mechanical properties and surface morphology [1, 2], as well as hydrogen isotope recycling [3, 4]. Tungsten (W) will be used as a plasma-facing material in ITER divertor [5], and it is considered also for application in future fusion devices. Therefore, investigation of He interaction with W is of great interest.
- ПубликацияОткрытый доступHYDROGEN CO-DEPOSITION WITH METALS IN PLASMA DISCHARGE(НИЯУ МИФИ, 2017) Krat, S. A.; Gasparyan, Yu. M.; Vasina, Ya. A.; Pisarev, A. A.; Писарев, Александр Александрович; Крат, Степан Андреевич; Гаспарян, Юрий МикаэловичDeposition of a single element film is always accompanied by co-deposition of a certain amount of other elements. This can be done properly to improve properties of the coating or due to contamination by impurities. In the field of thermonuclear fusion research, where hydrogen isotopes are used as a fuel, co-deposition with sputtered material from the wall is one of major mechanisms of hydrogen isotopes accumulation in the installation. Since D-T fuel will be used in ITER and future fusion reactors, accumulation of radioactive tritium will limit the lifespan of the installations due to safety concerns. For example, tritium accumulation in ITER is limited by 1 kg. This is why carbon materials were not accepted for the use in ITER. Basing on experiments, it was predicted that the safety limit could be reached after 100 of shots with tritium. Recent experiments in JET [1] demonstrated in the case of “ITER-like” wall (first wall – Be, divertor area - tungsten) accumulation of deuterium fuel in the co-deposits was 20 times lower than in the full-carbon wall campaign. This is both due to smaller amount of co-deposits and smaller concentration of deuterium in them.
- ПубликацияОткрытый доступSURFACE MODIFICATIONS OF W-BASED MATERIALS UNDER HELIUM AND DEUTERIUM ION IMPLANTATION(НИЯУ МИФИ, 2021) Ogorodnikova, O. V.; Klimov, N. S.; Gasparyan, Yu. M.; Harutyunyan, Z. R.; Efimov, V. S.; Kovalenko, D.; Gutarov, K.; Poskakalov, А. G.; Kharkov, M. M.; Kaziev, A. V.; Харьков, Максим Михайлович; Гаспарян, Юрий Микаэлович; Казиев, Андрей Викторович; Ефимов, Виталий Сергеевич; Огородникова, Ольга ВячеславовнаIn a thermonuclear reactor, materials will be irradiated with hydrogen isotopes and helium (He), neutrons, and heat fluxes. Tungsten (W) and dense nano-structured tungsten (CMSII) coatings are used as plasma-facing materials in current tokamaks and suggested to be used for future fusion devices. In this regard, the study of the accumulation of He and deuterium (D) in W based materials and corresponding surface modifications under normal operation conditions and transient events appears necessary for assessment of safety of fusion reactor due to the radioactivity of tritium and material performance and for the plasma fuel balance. Therefore, in this work, irradiation of W-based materials with D and He ions in stationary regime and in quasi-stationary high-current plasma gun QSPA-T below and above the melting threshold has been performed. In QSPA-T, a pulse duration was 1 ms and number of pulses was varied from one to thirty. In stationary plasma loads, ion energy was varied from 20 to 3 keV, temperature 300-1200 K and flux/fluence 1017-1021 at/m2s/1020-1025 at/m2.