Персона: Смирнов, Антон Дмитриевич
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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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Антон Дмитриевич
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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 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 ).
- 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 доступAnalysis of the methods for group constants generation for calculation of a large SFR core using Serpent 2 and CriMR codes(2020) Gerasimov, A. S.; Akpuluma, D. A.; Smirnov, A. D.; Pugachev, P. A.; Tikhomirov, G. V.; Смирнов, Антон Дмитриевич; Пугачев, Павел Александрович; Тихомиров, Георгий Валентинович© Published under licence by IOP Publishing Ltd.This work aimed at generating homogenized group constants using the Serpent code and then using the CriMR diffusion code to model the large SFR OECD 3600 MWth MOX core. The results were compared with a full core reference Monte Carlo solution by Serpent. Reactivity feedback parameters were also considered. Generating the group constants from separate fuel assemblies allows for simultaneously carrying out calculations and then using the results as input in diffusion codes rather than waiting so long for a 3D full core Monte Carlo calculation to be completed. From the results of the integral parameters we see a close agreement in the calculation codes. The differences can be attributed to the errors that could arise from generating the constants from individual sub-assemblies. The differences in the underlying physics and approximations used in development of the codes could also be a factor. Another way the errors could be reduced is by checking to see that the sub-assembly configurations used in the non-multiplying zones are as close as possible to the real layout in a full 3D core.
- ПубликацияОткрытый доступCORIUMSITY program code for the consequences analysis of a severe core melt accident(2020) Saldikov, I. S.; Bogdanova, E. V.; Pugachev, P. A.; Ryzhov, S. N.; Smirnov, A. D.; Ternovykh, M. Y.; Tikhomirov, G. V.; Богданова, Екатерина Владимировна; Пугачев, Павел Александрович; Рыжов, Сергей Николаевич; Смирнов, Антон Дмитриевич; Терновых, Михаил Юрьевич; Тихомиров, Георгий Валентинович© Published under licence by IOP Publishing Ltd.As part of the tasks to improve the nuclear safety of nuclear power plants, a new program code was developed. The CORIUMSITY program code developed, considered in this work, is intended to analyze the scenario in which an accident at a nuclear power plant is simulated with the melting of the core and the formation of the so-called "corium"- a mixture of nuclear and structural materials of the nuclear reactor core, formed as a result of thermal and mechanical impact during an accident. The CORIUMSITY program code, is intended to analyze several scenarios of different accidents, include an accident with reactor core melting. The functions of this code can help in solving many urgent nuclear safety problems. One of the main methods of operation of the CORIUMSITY code algorithms is the matrix exponential method, which consists in using a matrix function of a square matrix, in which as values are used indicators corresponding to nuclides from the CORIUMSITY code database. The program implements an iterative Euler method for solving the system of levels of nuclear fuel burnup. The CORIUMSITY code was verified with benchmark data to assess the accuracy of the calculation.
- ПубликацияТолько метаданныеVisualization of neutron characteristics distribution of debris particles(2020) Takezawa, H.; Muramoto, T.; Nishiyama, J.; Obara, T.; Pugachev, P. A.; Bogdanova, E. V.; Saldikov, I. S.; Smirnov, A. D.; Ternovykh, M. Y.; Tikhomirov, G. V.; Пугачев, Павел Александрович; Богданова, Екатерина Владимировна; Смирнов, Антон Дмитриевич; Терновых, Михаил Юрьевич; Тихомиров, Георгий Валентинович© 2020 National Research Nuclear University. All rights reserved.Accident at Fukushima Daiichi nuclear power plant led to increase of importance of safe-ty justification for processes at post-accident facilities in nuclear industry. One of such pro-cesses is extraction of corium from reactors cavity. Recriticality of this process is defined by potential unacceptable accident. This paper introduces supporting code for neutron fluxes and reaction rates visualization in systems with complex geometry that can be used in model-ing of corium removing works. Visualization code is based on Unreal Engine 4 game engine. Code allows observing neutronic functionals distribution in three dimensions. The reseach and provided implementation details help to understand the physical processes that take place as the accidents occur during corium removing works.
- ПубликацияОткрытый доступNeutronic modeling of a subcritical system with corium particles and water (from international benchmark)(2020) Smirnov, A. D.; Bogdanova, E. V.; Pugachev, P. A.; Saldikov, I. S.; Ternovykh, M. Y.; Tikhomirov, G. V.; Смирнов, Антон Дмитриевич; Богданова, Екатерина Владимировна; Пугачев, Павел Александрович; Терновых, Михаил Юрьевич; Тихомиров, Георгий Валентинович; Obara, T.; Nishiyama, J.; Muramoto, T.; Takezawa, H.