Персона: Смирнов, Антон Дмитриевич
Загружается...
Email Address
Birth Date
Научные группы
Организационные подразделения
Организационная единица
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.
Статус
Фамилия
Смирнов
Имя
Антон Дмитриевич
Имя
13 results
Результаты поиска
Теперь показываю 1 - 10 из 13
- ПубликацияТолько метаданныеIn 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 ДмитриевичProfessor 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.
- ПубликацияТолько метаданныеТЕСТОВЫЕ ЗАДАЧИ ДЛЯ ВЕРИФИКАЦИИ ПРОГРАММ РАСЧЕТА НЕЙТРОННО-ФИЗИЧЕСКИХ ХАРАКТЕРИСТИК АКТИВНОЙ ЗОНЫ РЕАКТОРА БН-1200(2016) Смирнов, А. Д.; Смирнов, Антон Дмитриевич; Тихомиров Георгий ВалентиновичЦелью 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.
- 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.
- ПубликацияТолько метаданныеMonte 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 ВалентиновичFuel 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.
- ПубликацияТолько метаданныеCurrent 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
- 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.
- 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 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 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.