2023, Умеренкова, А. С., Гаспарян, Ю. М., Арутюнян, З. Р., Ефимов, В. С., Сергеев, Н. С., Сорокин, И. А., Остойич, Н., Остойич, Никола, Гаспарян, Юрий Микаэлович, Умеренкова, Анастасия Сергеевна, Ефимов, Виталий Сергеевич, Сорокин, Иван Александрович, Арутюнян, Зорий Робертович, Сергеев, Никита Сергеевич

Helium isotope exchange in tungsten during sequential irradiation by 3He and 4He ions at room temperature was investigated. Tungsten samples were irradiated in two ways: by 3 keV mass-separated ion beams, or by low energy ions in plasma discharge. The He amount in W after irradiation was measured using ex-situ (up to 2500 K) thermal desorption spectroscopy. Despite the very high binding energy of He with lattice defects in W, a very high efficiency of He isotope exchange was observed.

2023, Efimov, N. E., Sinelnikov, D. N., Wang, Y., Harutyunyan, Z. R., Gasparyan, Y. M., Grishaev, M. V., Nikitin, I. A., Tan, X., Синельников, Дмитрий Николаевич, Ефимов, Никита Евгеньевич, Арутюнян, Зорий Робертович, Никитин, Иван Андреевич, Гаспарян, Юрий Микаэлович, Гришаев, Максим Валерьевич

One of the challenging problems which arise in the controlled nuclear fusion is related to the design and material choice of plasma facing components for the future reactors. Tungsten is considered to be one of the most suitable candidates due to its high melting point, thermal conductivity and relatively low erosion rate, and, therefore, it is planned to be used in nextgen facilities like ITER and DEMO. However, under high neutron fluxes its stable isotopes may form radioactive ones. Being not so hazardous while it is inside the reactor, in case of a loss of coolant accident (LOCA) a volatile oxide of W and of its transmutation products may appear, which is undesirable. A possible solution to avoid the release of the radioactive oxides is the use of self-passivating W-Cr-Y alloys [1], which under LOCA scenarios forms on the surface a chromium oxide, preventing the formation of tungsten oxide. Such alloys are of the great interest now, especially when it comes to analyzing the dynamics of the chromium release to the outermost layers [2,3]. In this work, capabilities of low energy ion scattering spectroscopy (LEIS) with small angle scattering to the characterization the surface morphology of W-11,4Cr-0,6Y after pre-annealing at different temperatures are revealed.

2020, Harutyunyan, Z. R., Gasparyan, Y. M., Efimov, V. S., Ryabtsev, S. A., Pisarev, A. A., Арутюнян, Зорий Робертович, Гаспарян, Юрий Микаэлович, Ефимов, Виталий Сергеевич, Писарев, Александр Александрович

© 2020, Allerton Press, Inc.Abstract: The retention of deuterium in tungsten preliminarily irradiated with helium ions is investigated by means of thermal desorption spectroscopy (TDS). Preliminary irradiation is performed with a He+ ion beam having an energy of 3 keV and fluence ranging in the interval 1019–1021 He m−2, in order to produce a defect structure typical of different degrees of damage. The specimens are then irradiated with D+3 ions having an energy of 2 keV and a weak fluence of 1019 D m−2, with subsequent TDS analysis to characterize the interaction between deuterium and helium-induced defects. It is shown that an increase in the deuterium retention is observed as soon as irradiation with the minimal fluence of helium ions, but it falls to the background level at fluences above 5 × 1021 He m−2.

2021, Harutyunyan, Z., Gasparyan, Yu., Pisarev, A., Litnovsky, A., Klein, F., Linsmeier, Ch., Писарев, Александр Александрович, Гаспарян, Юрий Микаэлович, Арутюнян, Зорий Робертович

During operation of the future fusion power plant, the plasma-facing components (PFC) will be exposed to intense fluxes of particles of deuterium, tritium, helium, as well as neutrons arising in the process of the D–T fusion reaction. In this regard, one of the important challenges is to minimize the accumulation of radioactive tritium in the PFC [1, 2], as well as to study the effect on the accumulation of helium impurities in the plasma.

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 Микаэлович

Reduced-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.

2020, Gasparyan, Y., Ryabtsev, S., Efimov, V., Harutyunyan, Z., Aksenova, A., Poskakalov, A., Kaziev, A., Kharkov, M., Ogorodnikova, O., Pisarev, A., Гаспарян, Юрий Микаэлович, Ефимов, Виталий Сергеевич, Арутюнян, Зорий Робертович, Аксенова, Александра Сергеевна, Казиев, Андрей Викторович, Харьков, Максим Михайлович, Огородникова, Ольга Вячеславовна, Писарев, Александр Александрович

Helium (He) is a product of deuterium-tritium (DT)-fusion reaction and will be a natural impurity in DT plasma in future fusion devices. He retention in tungsten irradiated by plasma and mass-separated ions in a wide temperature range (300-1200 K) was investigated by means of thermal desorption spectroscopy (TDS). He retention did not exceed the level of 1.5 x 10(21) He m(-2) for all investigated samples. A significant effect of air exposure on TDS spectra was demonstrated. In contrast to in situ TDS measurements, He release after interaction with the air started from similar to 400 K, even in the case of high temperature irradiation. Changes in surface morphology were analyzed by secondary electron microscopy. Blisters were found at the surface after ion irradiation at low temperatures. Acceleration of surface modification and more complex surface morphology was observed in the case of irradiation at temperatures above 1000 K.

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 Робертович

Helium (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.

2019, Kanashenko, S., Ivanov, Y., Ryabtsev, S., Gasparyan, Y., Efimov, V., Harutyunyan, Z., Aksenova, A., Poskakalov, A., Pisarev, A., Гаспарян, Юрий Микаэлович, Ефимов, Виталий Сергеевич, Арутюнян, Зорий Робертович, Аксенова, Александра Сергеевна, Писарев, Александр Александрович

© 2019Helium (He) retention in re-crystallized tungsten (W) irradiated with He+ ions at elevated temperatures (700–1200 K) and fluences in the range of 1020–1022 He/m2 was investigated by means of thermal desorption spectroscopy (TDS). Corresponding surface modifications were analyzed by scanning electron microscopy. Blisters were observed after irradiation at 700 and 1000 K, while the increase of the irradiation temperature up to 1200 K led to development of a complex sponge-like structure on the W surface. Significant surface transformations correlated with appearance of low temperature peaks in TDS spectra below the irradiation temperature. Possible mechanisms and explanations are discussed.