Персона: Тихомиров, Георгий Валентинович
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Институт ядерной физики и технологий
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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Руководитель научной группы "Лаборатория виртуальной реальности в области ядерных технологий"
Руководитель научной группы "Лаборатория инженерного компьютерного моделирования"
Руководитель научной группы "Лаборатория инженерного компьютерного моделирования"
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Тихомиров
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Георгий Валентинович
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- ПубликацияОткрытый доступNeutronic modeling of a subcritical system with corium particles and water from international benchmark(2020) Pugachev, P. A.; Saldikov, I.; Takezawa, H. ; Muramoto, T.; Nishiyama, J. ; Obara, T.; Богданова, Екатерина Владимировна; Терновых, Михаил Юрьевич; Тихомиров, Георгий Валентинович; Смирнов, Антон Дмитриевич; Tikhomirov, G. V.; Ternovykh, M. Y.; Bogdanova, E. V.; Smirnov, A. D.Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.After the accident at the Fukushima Daiichi nuclear power station, the attention of the scientific community is riveted on how the consequences are being eliminated. Removing corium - a resolidified mixture of nuclear fuel with other structural elements of the reactor - remains the most difficult task, the solution of which can take several decades. It is extremely important to exclude the occurrence of any emergency processes during the removal of corium. The purpose of this work was to solve a coordinated hydrodynamic and neutron-physical problem characterized by a large number of randomly oriented and irregularly located corium particles in water as part of the development of a benchmark for this class of problems. Monte Carlo- based precision codes were used to perform a neutronic analysis. The positions of particles with corium were obtained from the results of numerical simulation. The analysis results obtained using the codes involved showed good consistency for all the states considered. It was shown that modern neutronic codes based on the Monte Carlo method successfully cope with the geometric formation and solution of the problem with a nontrivial distribution of corium particles in water. The results of the study can be used to justify the safety of corium handling procedures, including its extraction from a damaged power unit.
- ПубликацияОткрытый доступA methodology for evaluating the transmutation efficiency of long-lived minor actinides(2021) Ashraf, O. ; Тихомиров, Георгий Валентинович; Tikhomirov, G. V.© 2021 Elsevier B.V.Up to now no definite internationally recognized quantitative criterion of minor actinides (MAs) transmutation efficiency was worked out, although this would be highly desirable. The absolute and relative total mass reduction of MAs are completely inadequate because they ignore the accumulation of higher radiotoxic long-lived MAs from the transmuted nuclide. In the current work, we introduce a new criterion for transmutation efficiency of MAs in nuclear reactors and demonstrate its efficiency by comparing two molten salt reactors; the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) and the Small Molten Salt Fast Reactor (SMSFR). Our proposed criterion takes into account the mass of all useful actinides, short-lived MAs, and short-lived fission products (FPs). In contrast, the mass parameters calculate the reduction in the MAs mass regardless of the produced nuclides. We introduce a new approach to load MAs into both reactors. The proposed approach merges the advantages of both homogeneous and heterogeneous approaches. The overall change in the actinides and FPs mass during the irradiation has been calculated using direct SERPENT-2 calculations. The results show that the transmutation efficiency of 241Am (the prime isotope for the transmutation) in the SD-TMSR is much higher than in the SMSFR. After 1500 days of radiation, the transmutation efficiency reaches 82.6% for SD-TMSR, however, for SMSFR it reaches 52.5%.
- ПубликацияТолько метаданныеSKETCH-N/ATHLET steady-state and dynamic coupling scheme verification on Kalinin-3 benchmark results(2020) Zimin, V. ; Nikonov, S. ; Perin, Ya. ; Henry, R. ; Velkov, K. ; Романенко, Владислав Игоревич; Тихомиров, Георгий Валентинович; Tikhomirov, G. V.; Romanenko,V. I.; Зимин, Вячеслав ГеннадьевичThe paper describes the multi-physics coupling scheme between the SKETCH-N nodal neutronics code and the best-estimate thermohydraulic code ATHLET v3.2. Some first results are discussed. Various possible options of coupling have been considered. A scheme is selected and applied for data exchange between the codes based on MPI library. The verification and validation were performed using the transient of the Kalinin-3 international Benchmark. The simulation results show good agreement with experimental data and calculations performed by the participants of the benchmark. Parallel to the coupling scheme development, a visualization system to process the results is being created. The steady-state analysis is carried out using both simple and complex thermohydraulic models. The calculations of the transient \Switch off of one MCP (Main Coolant Pump) at nominal power" is performed applying a more elaborate thermohydraulic model taking into account inter- channel mass transfer.
- ПубликацияОткрытый доступNeutronic analysis of VVER-1000 fuel assembly with different types of burnable absorbers using Monte-Carlo code Serpent(2019) Khrais, R. A.; Tikhomirov, G. V.; Saldikov, I. S.; Smirnov, A. D.; Тихомиров, Георгий Валентинович; Смирнов, Антон Дмитриевич© 2019 Published under licence by IOP Publishing Ltd. A neutronic study on the fuel assembly of a Russian type nuclear reactor VVER-1000 fuelled with low enriched Uranium (LEU) plus 12 UO 2 +4%Gd 2 O 3 rods was performed. This type of fuel requires validated computational methods and codes able to provide reliable predictions of the neutronics characteristics. Gadolinium self-shielding effect and isotopes accumulation in Rim region make it necessary to study the geometric modelling effect on the code calculations. The modelling of this fuel type was tested using Monte-Carlo and deterministic codes. In this study, Serpent results are verified using two nuclear data libraries ENDFb.6.8 and ENDFb.7. Also, this study investigates the effect UGd rods division into multiple radial layers on the reactivity, isotopic generation and burnup radial distribution. The same procedure is done on another type of neutron absorber Erbium (UEr) and the results are compared with UGd. The sensitivity of the results determines the validity of Monte-Carlo code in such a computational task comparing two types of neutron absorbers in addition to determining the geometric requirements.
- ПубликацияОткрытый доступNeutronic modelling of nanofluids as a primary coolant in VVER-440 reactor using the Serpent 2 Monte Carlo code(2019) Abdullah, H.; Smirnov, A. D.; Tikhomirov, G. V.; Смирнов, Антон Дмитриевич; Тихомиров, Георгий Валентинович© 2019 Published under licence by IOP Publishing Ltd. The nanofluids as an engineered fluid offer a large enhancement in heat transfer, in particular for boiling heat transfer and critical heat flux. These features lead to increase the power density of the nuclear reactor. In this paper, we investigate the neutronic simulation of nanofluids as a primary coolant in VVER-440 reactor and study the availability of using it without changing the system characteristics. The analysis of nanofluid fuel assembly is performed by using Serpent code. As a result of changing effective multiplication factor of the six types of nanoparticles which have been studied extensively for their heat transfer prosperities and absorption cross sections including Al 2 O 3 , Si, Zr, TiO 2 , CuO, and Ti with different volume fractions, it can be concluded the optimum nanoparticles are alumina at concentration 0.01 volume fraction.
- ПубликацияОткрытый доступModeling and criticality calculation of the Molten Salt Fast Reactor using Serpent code(2019) Ashraf, O.; Smirnov, A. D.; Tikhomirov, G. V.; Смирнов, Антон Дмитриевич; Тихомиров, Георгий Валентинович© 2019 Published under licence by IOP Publishing Ltd. In Molten Salt Fast Reactors (MSFR), a liquid-fuel circulates through the cylinder core and transport the fission heat to the Intermediate external Heat Exchangers (IHX), therefore liquid salt allows carrying the fuel and transfer heat. The MSFR supposed to work in a closed Th-based fuel cycle with a full reprocessing of fission products and all actinides in the core. The aim of this paper is; modeling the primary circuit of the MSFR (based on the European model) in order to identify the composition of the start-up fuel required to the criticality. In conclusion, the compositions of the start-up liquid fuel required for criticality and long life cycle were determined precisely for three different types of fissile materials ( 233 UF 4 , PuF 3 and TRUF 3 ).
- ПубликацияОткрытый доступЭкспериментальные комплексы исследовательского ядерного реактора ИРТ МИФИ(НИЯУ МИФИ, 2013) Абов, Ю. Г.; Алферов, В. П.; Бушуев, А. В.; Горбунов, А. В.; Графутин, В. И.; Гулько, А. Д.; Ермаков, О. Н.; Ефременко, Ю. В.; Зайцев, К. Н.; Камнев, В. А.; Крамер-Агеев, Е. А.; Кожин, А. Ф.; Кузнецов, С. П.; Кулаков, В. Н.; Липенгольц, А. А.; Львов, Д. В.; Ляпунов, С. М.; Мищерина, О. В.; Петрова, Е. В.; Петрунин, В. Ф.; Попов, В. Д.; Портнов, А. А.; Сахаров, В. К.; Трошин, В. С.; Фунтиков, Ю. В.; Хохлов, В. Ф.; Шагурин, И. И.; Шейно, И. Н.; Щуровская, М. В.; Тихомиров, Георгий Валентинович; Алеева, Татьяна Борисовна; Джепаров, Фридрих Саламонович; Болоздыня, Александр Иванович; Акимов, Дмитрий Юрьевич; Зубарев, Виктор Николаевич; Стогов, Юрий Владимирович; Сосновцев, Валерий Витальевич; Щуровская, Мария ВладимировнаВ данной книге представлены описания экспериментальных комплексов для проведения нейтронно-физических исследований на исследовательском ядерном реакторе ИРТ МИФИ. В основу книги легли материалы научного семинара ИРТ-2013, проведенного 29 января 2013 г. физико-техническим факультетом и сотрудниками Атомного центра НИЯУ МИФИ. Основная задача издания – информировать об экспериментальных возможностях реактора ИРТ МИФИ и способствовать повышению качества подготовки специалистов в области ядерной физики и технологий, развитию у студентов, аспирантов и молодых специалистов практических навыков работы с современным ядернофизическим оборудованием, обеспечению профессорско-преподавательского состава НИЯУ МИФИ условиями для проведения исследовательских работ, а также привлечения научных центров и предприятий Росатома к участию в совместных исследованиях. Данное издание может быть использовано в качестве пособия для совершенствования системы подготовки и переподготовки специалистов для ведущих научных центров атомной отрасли России.
- ПубликацияТолько метаданныеCurrent status of SMRs and S&MRs development in the world(2023) Pioro, I. L. ; Duffey, R. B. ; Kirillov, P. L. ; Dort-Goltz, N. ; Тихомиров, Георгий Валентинович; Смирнов, Антон Дмитриевич; Smirnov, A. D.; Tikhomirov, G. V.This chapter examines Small Modular Reactors (SMRs), which are modular-type nuclear reactors with installed capacities ≤ 300 MWel with claimed features of “modularity” in design, production, and/or construction, and Small- and Medium-size Reactors (S&MRs), with installed capacities ≤ 300 MWel (Small) and > 300–700 MWel (Medium-size), many having claimed features of “modularity” in design, production, and/or construction. The requirements and objectives for any and all new nuclear reactors of any and all sizes are given as: safer than previous “generations”; having low financial risk exposure and capital cost; ease and speed of build; readily licensable; simple to operate and secure; assured fuel supply and sustainability; providing social value and acceptance; and still being competitive. Existing SMRs and S&MRs are tabulated by type, country, and status. Although many SMR designs and concepts have been proposed, Russia is the first country in the world to develop, design, and put into operation two SMRs, and Russian technology is examined in detail in this chapter, with numerous diagrams and photos of various systems provided.
- ПубликацияОткрытый доступStudy the neutronic feasibility of using Zr as an energy regulator instead of traditional methods(2021) Abdelghafar, Galahom A.; Elazaka, A. I.; Tikhomirov, G. V.; Тихомиров, Георгий Валентинович© 2021 John Wiley & Sons LtdThe control of the reactor energy increases the reactor fuel cycle time. In this work, a new method has been investigated to control the reactor energy instead of traditional methods. SERPENT2 version 2.1.30 was used to investigate the possibility of using Zirconium rods as an energy regulator in the VVER-1000. Zirconium rods were used as water displacers. The effect of different Zr rod diameter on the infinite multiplication factor (k∞) of the reactor and the fuel concentration at different fuel burnup steps has been analyzed. Larger fuel pitches have been investigated to increase the Zr rod diameter. The main safety parameters such as the void volume reactivity coefficients, the Doppler reactivity coefficients and the Moderator temperature coefficient have been studied at different fuel pitch and different Zr rod diameter. A comparison between the effect of Zr rods and boric acid on the reactor reactivity proved the efficiency of using Zr rods as reactivity regulation.
- ПубликацияОткрытый доступA new approach of spectral regulation in the pressurized water reactors using Zr rods(2020) Elazaka, A. I.; Tikhomirov, G. V.; Тихомиров, Георгий Валентинович© Published under licence by IOP Publishing Ltd.The elongation of the fuel cycle period and the reactor reactivity control are crucial for all reactor suppliers. The traditional methods of boric acid, burnable absorbers, and control rods to control the reactor reactivity have their defects. A new method of the neutron spectral shift is studied to perform the reactor reactivity control and maintain the length of the fuel cycle. The neutron spectrum regulation can be achieved with different methods, including the Zr rods insertion in the fuel assemblies as water displacers. In this work, SERPENT2 version 2.1.30 was used to investigate a new fuel assembly design in the basics of Sierpinski carpet geometry to utilize the Zr rods as water displacers. The study of different fuel pitches of the Sierpinski fuel assembly is investigated to choose the reference fuel of the new design. The effect of boric acid concentration increasing and defining the maximum boric acid concentration in the Sierpinski fuel assembly are studied. The primary safety parameters of the Doppler effect and moderator temperature reactivity coefficients calculations approved the safety of the new Sierpinski fuel assembly design. The infinite multiplication factor at different steps of burnup showed the efficiency of the Sierpinski fuel assembly model to control the reactor reactivity. The average breeding factor to the burnup limit of 65 GWd/T is 0.68 in the Sierpinski fuel assembly with the presence of Zr rods.