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Кулешова, Евгения Анатольевна

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Цель ИЯФиТ и стратегия развития - создание и развитие научно-образовательного центра мирового уровня в области ядерной физики и технологий, радиационного материаловедения, физики элементарных частиц, астрофизики и космофизики.

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Кулешова

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Евгения Анатольевна

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Термическая стойкость сталей с повышенными характеристиками прочности для корпусов перспективных реакторов ВВЭР различных проекто
(2023) Мальцев, Д. А.; Крымская, Ольга Александровна; Исаенкова, Маргарита Геннадьевна; Кулешова, Евгения Анатольевна; Федотов, Иван Владимирович; Минушкин, Роман Александрович
Рассмотрены результаты структурных исследований и механических испытаний после длительного теплового воздействия лабораторных плавок металлургически усовершенствованной стали 15Х2НМФА и стали с повышенным содержанием никеля, рассматриваемых в качестве материалов корпусов перспективных реакторов ВВЭР различных проектов. Показано, что как для усовершенствованной стали 15Х2НМФА, так и для высоконикелевой стали отсутствуют признаки зернограничного охрупчивания после проведения термообработки, провоцирующей охрупчивание, что связано с очень низкой зернограничной сегрегацией фосфора в исходном состоянии, обусловленной повышенной дисперсностью структуры, а также достаточно низким содержанием примесей. Кроме того, для усовершенствованной стали 15Х2НМФА не выявлено изменения предела текучести, что согласуется с результатами структурных исследований. Для высоконикелевой стали в результате длительного теплового воздействия выявлена тенденция к небольшому снижению предела текучести на 5 – 10% и закономерному снижению критической температуры хрупкости. Снижение механических характеристик связано с относительно низкой температурой высокого отпуска высоконикелевой стали и, соответственно, возможным протеканием процессов доотпуска структуры при длительном тепловом воздействии, о чем свидетельствуют результаты рентгеноструктурных исследований. Несмотря на протекание процессов доотпуска в высоконикелевой стали при длительном тепловом воздействии сохраняется стабильность основных упрочняющих карбидных фаз, за счет чего предел текучести остается на достаточно высоком уровне, превышающем значения для используемых сталей корпусов реакторов ВВЭР даже при тепловом воздействии с параметрами, значительно превосходящими предполагаемые условия эксплуатации корпусов перспективных ВВЭР. При этом снижение критической температуры хрупкости при тепловом воздействии лишь способствует повышению стойкости стали против хрупкого разрушения при эксплуатации.
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Thermal resistance of steels with increased strength properties for pressure vessels of advanced VVER reactors of various designs
(2023) Maltsev, D. A.; Исаенкова, Маргарита Геннадьевна; Крымская, Ольга Александровна; Федотов, Иван Владимирович; Кулешова, Евгения Анатольевна; Минушкин, Роман Александрович; Kuleshova, E. A.; Fedotov, I. V.; Isaenkova, M. G.; Krymskaya, O. A.; Minushkin, R. A.
The paper considers the results of structural studies and mechanical tests after a long-term thermal exposure of laboratory heats of the metallurgically improved 15Kh2NMFA steel and steel with an increased content of nickel considered as materials for the pressure vessels of advanced VVER-type reactors of various designs. It has been shown that, both for the improved 15Kh2NMFA steel and the high-nickel steel, there are no signs of grain boundary embrittlement after an segregation provoking embrittlement heat treatment. This is explained by the extremely low grain boundary segregation of phosphorus in the initial state caused by a high degree of the structure dispersity as well as by rather a low content of impurities. Besides, no changes have been found in the yield strength value for the improved 15Kh2NMFA steel, which agrees with the structure investigation results. For the high-nickel steel, a tendency towards a minor yield strength decrease by 5 to 10% and a regular reduction of the critical brittleness temperature has been revealed. A decrease in the mechanical properties has been caused by a relatively low temperature of tempering for the high-nickel steel and, accordingly, by the potential occurrence of the structure recovery during long-term thermal exposure, as evidenced by the results of an X-ray diffraction analysis. Despite the structure recovery in the high-nickel steel under the long-term thermal exposure, the main strengthening carbide phases remain stable. Due to this, the yield strength value remains at a relatively high level that exceeds the values for the modern VVER-type vessel steels, even in the case of a thermal exposure much in excess of the expected operating conditions for advanced VVER reactors. The observed decrease of critical brittleness temperature during the long-term thermal exposure contributes to an increase in the steel resistance to brittle fracture.
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Complex study of grain boundary segregation in long-term irradiated reactor pressure vessel steels
(2020) Fedotova, S. V.; Maltsev, D. A.; Saltykov, M. A.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2019 Elsevier B.V.Grain boundary (GB) embrittlement of reactor pressure vessel steel occurs at the long-term operation through the radiation-enhanced segregation of impurities. All the known techniques of studying the GB segregation have their advantages and disadvantages, and complete information can be obtained only through a comprehensive analysis by different methods. In this paper the level of GB segregation in VVER-1000 RPV steels with high nickel content was assessed by combined methods of Auger electron spectroscopy (by the P monolayer coverage and atomic concentration of other elements); atom probe tomography (by the Gibbsian interfacial excess) and fractographic analysis (by the maximum fraction of brittle intergranular fracture). The results obtained by different methods are in good agreement and show that phosphorus GB segregation increases with the increase of fast neutron fluence, GB concentration of Ni and Mn, the bulk concentration of Ni and Mn, and correlates with the fraction of brittle intergranular fracture.
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Grain Boundary Embrittlement of Steels of Vver-1000 Reactor Vessels Under Long-Term Operation
(2019) Mal'tsev, D. A.; Fedotova, S. V.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2019, Springer Science+Business Media, LLC, part of Springer Nature.Results of fractographic analysis and Auger electron spectroscopy of materials of vessels of VVER-1000 reactors obtained at the “Kurchatov Institute” Research Center are systematized. Comparative analysis of the effect of the operating factors on the level of grain boundary embrittlement in the matrix metal and welded joints is performed. The contribution of grain boundary embrittlement into the total radiation-induced embrittlement of the weld metal of the vessels is determined.
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Evolution of the Structural-Phase State of E110 Alloy Fuel Rod Claddings at High Temperatures and Stresses
(2020) Frolov, A. S.; Gurovich, B. A.; Maltsev, D. A.; Safonov, D. V.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2020, Pleiades Publishing, Ltd.The results of microstructural studies of the fragments of E110 alloy fuel rod cladding specimens based on sponge and electrolytic zirconium after the operation as a part of VVER-1000 fuel assemblies with the subsequent creep tests under axial loading are presented. It was shown that, during the creep tests, the studied specimens showed no changes in the chemical composition, average size, and bulk density of the second phases, including radiation-induced ones. It was found that, during the creep tests, dislocation loops were annealed, i.e., an increase occurred in their average size with a simultaneous decrease in bulk density. It was shown that the specimens of fuel rod claddings made of an alloy based on electrolytic zirconium demonstrated more significant creep resistance compared to the sponge-based zirconium alloy specimens, which is related to a higher density of globular β-Nb precipitates in the irradiated electrolytic zirconium specimens.
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Comparison of the high Ni VVER-1000 weld microstructure under the primary irradiation and re-irradiation
(2020) Gurovich, B. A.; Fedotova, S. V.; Zhuchkov, G. M.; Frolov, A. S.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2020 Elsevier B.V.This paper presents the first comparative experimental assessment of the microstructures of the same surveillance specimens of VVER-1000 high Ni weld: after the primary irradiation, recovery annealing and re-irradiation in the same industrial NPP (Balakovo NPP unit 1) to assess the possibility of its lifetime extension after annealing up to 60 years. Therein, studies of both the initial (carbides) and radiation-induced hardening phases (precipitates and dislocation loops), and the behavior of grain boundary segregation as well are presented. It has been shown that the microstructures of high Ni VVER-1000 weld after the primary irradiation and re-irradiation after recovery annealing up to the close values of fast neutron fluences are identical: the characteristics of both initial and radiation-induced hardening phases, as well as grain boundary segregation, are the same.
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Radiation-Induced Phase Formation in Steels of VVER Reactor Pressure Vessels Containing ~0.3–1.3 wt % Nickel
(2019) Frolov, A. S.; Zhuchkov, G. M.; Fedotov, I. V.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2019, Pleiades Publishing, Ltd.Abstract: The radiation-induced structural elements in the materials of water-moderated water-cooled reactor pressure vessels have been studied by TEM and atomic probe tomography at the National Research Center Kurchatov Institute. The effect of Ni concentration in the range from 0.34 to 1.28 wt % on the formation of phases under fast neutron irradiation has been analyzed. The volume density of radiation-induced phases has been shown to depend on the Ni concentration in steel; the phase sizes and compositions remain almost the same. The volume density of these precipitations affects the radiation hardening, one of the radiation embrittlement mechanisms. The quantitative parameters of radiation-induced phases have been shown to control the service life growth, the enhancement of the thermal stability, and the melting technology of reactor vessel steels bearing nickel in the 0.3–0.7 wt % range.
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Annealing as a Technique for Estimating the Structural Elements Contribution to NPP Materials Service Properties
(2019) Fedotov, I. V.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2019, Pleiades Publishing, Ltd.Abstract: The annealing technique is used to estimate the effect of radiation-induced structural elements on the service properties of the materials used in nuclear power plants: reactor pressure vessel (RPV) and in-vessel internals steels. The WWER-1000 RPV weld seam is affected by a hardening mechanism, the contribution of which is 60–65%, and a nonhardening mechanism resulting in grain-boundary segregation, the contribution of which is 30–35%. Radiation defects 40–45% and G-phase precipitates 35–40% are mainly responsible for the radiation-induced hardening of WWER-1000 internals materials. Radiation-induced structural changes should be eliminated for these materials to be reused.
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Degradation of Fuel Cladding Materials Based on Zirconium after Operation in VVER-Type Reactors
(2019) Frolov, A. S.; Gurovich, B. A.; Maltsev, D. A.; Safonov, D. V.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2019, Pleiades Publishing, Ltd.Abstract: The paper presents microstructural studies of specimens cut from fuel elements made of E110 spongy zirconium-based alloy after operation in a VVER-1000 before reaching the burnup of ~35 MWd/kg U. As a result of exposure to high temperatures and neutron irradiation, significant changes in the phase composition of fuel cladding materials appear: change in the size, density, and composition of β-Nb particles; change in the composition of the Laves phase; formation of dislocation loops of α type, as well as δ and γ hydrides. The main structural elements determining the degradation of the mechanical properties of the E110 alloy under irradiation are dislocation loops and fine-phase precipitates owing to their relatively large density. The data obtained can be used to construct dose dependences of microstructural changes with the aim of predicting the residual life of claddings and fuel assemblies as a whole.
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Structural evolution features of the 42XNM alloy during neutron irradiation under VVER conditions
(2021) Gurovich, B. A.; Frolov, A. S.; Maltsev, D. A.; Safonov, D. V.; Kuleshova, E. A.; Кулешова, Евгения Анатольевна
© 2020The paper presents the results of microstructural studies by transmission electron microscopy and atom probe tomography of the Ni-Cr-Mo (42XNM) alloy in the initial state, as well as after neutron irradiation, in the range of fast neutron fluence (0.9–1.5)•1026n•m−2 (E>0.1 MeV) (5–12 dpa) at ~ 300°C. It was shown that during neutron irradiation, various defects were formed – dislocation loops and pores, the bulk density and size of the second phase precipitates (TiN and α-Cr) remained stable. Quantitative analysis of chemical composition for regions near the boundaries of austenitic grains was made, which showed typical effects for materials with an FCC lattice (which contains Ni and Cr) after neutron irradiation – nickel depletion of grain boundaries and enrichment in chromium. Similar studies were carried out for segregation processes on the Frank loop edges.