Персона:
Смирнов, Антон Дмитриевич

Научные группы

Научная группа

Лаборатория инженерного компьютерного моделирования (Кафедра №5 ИЯФиТ)

Организационные подразделения

Организационная единица

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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Смирнов

Имя

Антон Дмитриевич

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Только метаданные
Monte Carlo codes benchmarking on sub-critical fuel debris particles system for neutronic analysis
(2022) Smirnov, A.; Bogdanova, E.; Pugachev, P.; Ternovykh, M.; Saldikov, I.; Tikhomirov, G.; Смирнов, Антон Дмитриевич; Богданова, Екатерина Владимировна; Пугачев, Павел Александрович; Терновых, Михаил Юрьевич; Тихомиров, Георгий Валентинович
Fuel debris removal is the most challenging part of damaged nuclear power station decommissioning. It is important to carry out nuclear safety calculations accurately and quickly enough. Here, it was clarified that modern codes based on the Monte Carlo method were capable of performing neutronic analysis with the same accuracy and without significant differences in the results. The benchmark calculations were performed using three codes: MVP, Serpent, and MCU. In this study, the comparison focused on multiplication factor, neutron fluxes and reaction rates relative difference, and calculation time of many fuel debris particles system. Then the calculation results were used when codes comparing. It was shown that the calculation results showed good agreement between all codes. It was assumed that minor differences in the thermal range of neutron fluxes can be caused by different thermal neutrons scattering treatment for all codes. The study also showed that solving such problems requires significant computing power and time. It has been proven that the statistical geometry model in the MVP and the explicit stochastic geometry model in the Serpent have the possibility to provide solutions with the same accuracy, but much faster.
Только метаданные
In Memoriam: Professor Pavel Leonidovich Kirillov (Aug. 20, 1927-Oct. 10, 2021)
(2022) Pioro, I.; Duffey, R. B.; Murogov, V.; Tikhomirov, G.; Smirnov, A.; Тихомиров, Георгий Валентинович; Смирнов, Антон Дмитриевич
Professor Pavel L. Kirillov died on Oct. 8, 2021, on his 95th year after a life as a husband, father, and an internationally renowned scientist, researcher, and educator in the field of nuclear engineering, thermalhydraulics, heat transfer, and two-phase flow. He was passionate and dedicated in everything that he did and leaves an incredible legacy to the profession. He was born on Aug. 20, 1927 in Russia, and received his M.A.Sc. degree in thermal physics in 1950 (Moscow Power-Engineering Institute (MPEI) (Московский Энергетический Институт (МЭИ)), Faculty of Physics and Power Engineering (Физико-Энергетический Факультет), Ph.D. and Doctor of Technical Sciences degrees—in 1959 and 1969, respectively. Professor P. L. Kirillov was a Fellow of the International and National Engineering Academies; member of the Russian Nuclear Society and ASME; member of Scientific Councils of the Institute of Atomic Energy by the name of I. V. Kurchatov (ИАЭ им. И.В. Курчатова) (1985–1990) and A.I. Leypunsky Institute for Physics and Power Engineering (IPPE) (Физико-Энергетический Институт (ФЭИ)) (from 1975); member of the Journal Boards of the Atomic Energy (Атомная Энергия) (from 1977) and the ASME Journal of Nuclear Engineering and Radiation Science (from 2014). After graduating from the MPEI (МЭИ) in 1950, Pavel Kirillov has joined the IPPE (ФЭИ) (Obninsk, Russia), currently, State Scientific Centre of the Russian Federation—Leypunsky Institute for Physics and Power Engineering, Joint-Stock Company (IPPE JSC) (Акционерное общество «Государственный научный центр Российской Федерации – Физико-энергетический институт имени А.И. Лейпунского» (АО «ГНЦ РФ—ФЭИ»)) as a junior scientist in 1950 (he participated in construction and operation of the world's first nuclear power plant in Obninsk, AM-1 (“Atom Peaceful”–1 in Russian abbreviations (Атом Мирный)), which was commissioned at the IPPE on June 27, 1954), and worked there on various positions: Senior scientist (1953–1954); head of laboratory (1954–1969); head of branch (1969–1975); director of thermal-physics division (1975–1995); deputy director of thermal-physics division (1995–2010); advisor of the director of thermal-physics division (from 2010) and of the general director of IPPE JSC. He was an associate professor (1959–1965); professor (1965–1972); chair of the thermal-physics department (1972–1985); and chair of the nuclear-power-plant department at the Obninsk Branch of the Moscow Engineering Physics Institute (MEPhI) (Обнинский филиал Московского Инженерно-Физического Института (МИФИ)) (1985–1992). Professor Pavel Kirillov has prepared a large number of undergraduate and master-degree students; and 15 Ph.D. candidates. Knowledgeable, friendly and technically informed, Dr. P.L. Kirillov was a role model and mentor to numerous generations of researchers/scientists in nuclear engineering, thermalhydraulics, heat transfer, and two-phase-flow fields. He is definitely one of the most admired and ingenious researchers in these fields. His many researches and achievements include contributions in such special areas as molten-metals nuclear-reactor coolants; supercritical water; research (BR-10 (Fast Reactor (sodium-cooled)) and BOR-60 (fast experimental reactor (sodium-cooled)); power (BN-350 and BN-600 (fast sodium power reactors)), transportation (lead-bismuth-cooled), and spacecraft (BUK and TOPAZ) nuclear reactors. Professor Kirillov is well respected among his colleagues in the nuclear-engineering community all over the world despite being heavily focused on Russian developments to which he made major contributions. His 2009 text on “Hydrodynamic Calculations” (in Russian) covered and demonstrated his encyclopedic knowledge of fluid flow and heat transfer. The earlier 2007 major text “Thermophysical Properties of Materials for Nuclear Engineering” (in English) sets the standard for excellence, breadth and depth with not only essential basic data and tabulations, but includes fundamental design information for all types of reactors. For his outstanding work, Professor P. L. Kirillov was awarded with the following honored titles: Honored Scientist of Science and Engineering of the Russian Federation (1988) (Заслуженный деятель науки и техники РСФСР) and Honored Worker of the Atomic Industry of the Russian Federation (2016) (Заслуженный работник атомной промышленности Российской Федерации); and with three state orders and a number of state and jubilee medals. During his work at the IPPE (ФЭИ) and Obninsk Branch of MEPhI (МИФИ), Professor Kirillov has published over 350 technical publications including handbooks, reference books, textbooks, papers, inventions, and reports (see selected publications listed below). His professional contributions and critical thinking continued unabated and he was fully involved in the series of articles summarizing the status of nuclear energy in the world and its future prospects. Professor Pavel Leonidovich Kirillov was a respected technical leader, mentor, and friend to innumerable students, researchers, scientists, and engineers, and he will be sadly missed by all who had the privilege to know him. He was an outstanding contributor in every aspect of his prolific work and career in the true traditions of technical excellence and critical thinking, and his irreplaceable loss is deeply felt worldwide.
Только метаданные
Current status of SMRs and S&MRs development in the world
(2023) Pioro, I. L.; Duffey, R. B.; Kirillov, P. L.; Tikhomirov, G. V.; Smirnov, A. D.; Тихомиров, Георгий Валентинович; Смирнов, Антон Дмитриевич
Только метаданные
ТЕСТОВЫЕ ЗАДАЧИ ДЛЯ ВЕРИФИКАЦИИ ПРОГРАММ РАСЧЕТА НЕЙТРОННО-ФИЗИЧЕСКИХ ХАРАКТЕРИСТИК АКТИВНОЙ ЗОНЫ РЕАКТОРА БН-1200
(2016) Смирнов, А. Д.; Смирнов, Антон Дмитриевич; Тихомиров Георгий Валентинович
Целью данной дипломной работы является изучение специфики нейтронно-физического расчета быстрых реакторов, обзор существующих бенчмарков для реакторов этого типа и разработка системы тестовых задач для верификации программ нейтронно-физических расчета характеристик реактора БН-1200. В первой главе приведено подробное описание направлений деятельности в области нейтронного моделирования, обозначены основные особенности нейтронно-физического расчета быстрых реакторов, рассмотрены различные типы тестовых задач и дается обзор основных бенчмарков по реакторам на быстрых нейтронах. Во второй главе описывается классификация кодов нейтронной физики и рассмотрены программы, использующиеся в расчетах в рамках дипломной работы. В третьей главе представлены тестовые задачи, разработанные в ходе выполнения дипломной работы. Приведены решения этих тестов и анализ полученных результатов.
Только метаданные
Development of a test task for coupled neutronic and thermal-hydraulic analysis of LFR fuel cell
(2025) Chubarov, M.; Zakharov, M.; Bogdanova, E.; Smirnov, A.; Ouen, D.; Tikhomirov, G.; Чубаров, Максим Алексеевич; Захаров, Михаил Юрьевич; Богданова, Екатерина Владимировна; Смирнов, Антон Дмитриевич; Оуэн, Дэниэл Юджин; Тихомиров, Георгий Валентинович
Только метаданные
Current 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.
Открытый доступ
CORIUMSITY 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.
Открытый доступ
Visualization 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.
Открытый доступ
Neutronic 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.
Открытый доступ
Neutronic 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.