Персона: Самохин, Дмитрий Сергеевич
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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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Дмитрий Сергеевич
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- ПубликацияОткрытый доступVerification the Reliability of Using (Non-Nuclear Grade) Electronic Devices in Nuclear Installations(2020) Samokhin, D.; Alslman, M.; Vostrilova, A.; Самохин, Дмитрий Сергеевич© 2020 The Authors. Published by Elsevier B.V.This paper describes the experimental approach to be performed to examine the reliability indicator of certain electrical devices manufactured by companies in the Russian Federation. The tests will be conducted according to technical specifications and test procedures which were prepared for these devices. In addition, a comparison referring to the reliability parameters between results and manufactured specifications will be provided or conducted.
- ПубликацияОткрытый доступComputer analysis of the system during generation ceramic covering on a cladding of a fuel element simulator with sublayer lead-magnesium-zirconium(2020) Orlova, E.; Samokhin, D.; Holkin, I.; Орлова, Екатерина Андреевна; Самохин, Дмитрий Сергеевич© 2020 The Authors. Published by Elsevier B.V.The method of computer analysis of system parameters at formation of protective coating on the shell of fuel element of fast reactors is presented, which can serve as a prerequisite for development of cognitive architecture of control of nuclear power engineering processes by control of composition of gas phase in reactor core, containment, other tanks and making decisions on results of process parameters deviation from stationary function. The work was carried out to justify the prospects of using a liquid metal sublayer (LMS) instead of a gas sublayer in fuel elements, which allows, due to the high thermal conductivity of the liquid metal, to increase the width of the gap between the fuel and the shell and, therefore, to increase the fuel element life limited by the contact of the fuel with the shell. At the same time, safety in case of UTOP (uncontrolled increase of power), ULOF (termination of coolant flow through the reactor) accidents is increased due to significant decrease of temperature in the fuel center. However, the possibility of using LMS is limited by the substantial mass transfer-corrosion dissolution in the lead of the components of the fuel element shell in the hot zone and precipitation in the cold zone. In this article, the basis for computer analysis was studies on the absorption of gas (nitrogen) when forming a protective ceramic coating of zirconium nitride on the surface of the shell of a pipe-in-pipe type fuel element simulator to suppress the corrosion process from eutectic alloy lead-magnesium-zirconium. Abnormal intense decrease of nitrogen pressure (p, kPa) in expansion tank with Pb-Mg-Zr alloy at increased temperature (about 950 K) according to linear kinetic law (t, c), connected with its interaction with system components, was revealed: P, kPa=-0.0002 t-77.867 R2=0.942. Method of diagnostics of system parameters at formation of anticorrosive coating by nitrogen absorption is developed, empirical dependencies of process flow in wide range of temperature are built.
- ПубликацияОткрытый доступReactor with metallic fuel and lead-208 coolant(2020) Samokhin, D.; Khorasanov, G.; Самохин, Дмитрий Сергеевич© 2020 The Authors. Published by Elsevier B.V.This time, several projects dedicated to fast reactors (FRs) with lead and lead-bismuth coolants, BREST-OD-300, SVBR-100, RBETS-M, BRUTs are proposed in Russia. They will have several valuable consuming properties: Chemical inertness, low neutron absorption, low activation and others. Usage of lead coolant leads also to the possibility of achieving a hard-enough neutron spectrum that allows increasing the incineration probability of 241Am, 237Np and other low fissile actinides. High power FRs have large-sized cores that limits the value of neutron energy by the value of 0.5 MeV, which is insufficient for incineration of above mentioned actinides. Small and medium power reactors have smaller cores and, respectively, have harder neutron spectra. Usage of lead and low moderating innovative fuel allow further increasing neutron energy to the value inquired for low fissile actinides incineration.In the paper a possibility of obtaining a neutron spectrum with the average value of neutron energy higher than 0.5 MeV is considered. It is performed in the frame of the project of BRUTs series reactors, i.e. small power LFRs proposed in the Obninsk Institute for Nuclear Power Engineering. A scope for achieving a hard neutron spectrum in the reactor BRUTs-25 core of small sizes, D×H = 0.5×0.4 m2, is shown. Findings are that in the core fueled with Pu-Am-Np-Zr alloy and cooled with enriched lead, 208Pb, the average value of neutron energy, , is high-enough, about 0.95 MeV, as well as the share of fast neutron, En>0.8 MeV, in the neutron spectrum is very high, about 40%. In such of conditions, 241Am and 237Np incineration probabilities in the core center are higher than 50% and values of their one-group fission cross-sections are higher than 0.7 barn. This circumstance allows burning 15-16wt% of low fissile isotopes for one campaign of BRUTs-25. The presence of 241Am in the fuel, in a quantity of 28.7 kg, allows transmuting about 8.6 kg of its mass for one reactor campaign that lasts about 3 years (1000 effective days). It means that to transmute the quantity of 241Am produced by the VVER-1000 for one year, equal to 25.8 kg, it will be needed about 3 BRUTs-25 type low power reactors operating for 1000 effective days.
- ПубликацияТолько метаданныеTo estimate reliability of hydroaccumulator system of nuclear power plants and improve its design and reliability(2020) Abdul Awal Rana, M.; Zihad Ul Haque, M.; Shahabuddin, A. K. M.; Samokhin, D.; Sanam, S. A.; Самохин, Дмитрий Сергеевич© Published under licence by IOP Publishing Ltd.There are around twenty branches in a Nuclear Power Plant that operate to maintain the plant and one of the most important branches of them is 'Safety and Reliability'. Approximately forty percent of the investment is paid to ensure the safety and reliability of the plant. Goal of this research work is to estimate reliability of a mechanical system which is chosen as Hydroaccumulator System of VVER type reactor. Getting the reliability, we will be trying to improve the system and enhance its reliability. To do so, firstly, reliability is calculated from a 'prototype reactor's' hot loop which is located at Obninsk Institute for Nuclear Power Engineering (INPE). Thus, calculations were done and its applications being known, then finally these formulas and experiences are being used to calculate and improve the Hydroaccumulator System's design along with its reliability to increase the safety of the NPP.
- ПубликацияОткрытый доступClarification the reliability of electronic components used in nuclear industry(2019) Samokhin, D. S.; Leonova, T. N.; Alslman, M.; Vostrilova, A. D.; Самохин, Дмитрий Сергеевич; Леонова, Татьяна Николаевна; Факультет бизнес-информатики и управления комплексными системами© Published under licence by IOP Publishing Ltd. When calculating the reliability indicator of any device, under specified operating condition, usually reference data on the characteristics of the reliability of the elements. Data on the reliability of the groups of electrical-radiation equipment (ERE), used in the design, manufacture and operation of equipment, instruments, devices include information on the mathematical models to calculate (predict) values for the operational intensity of product groups, including information on the storage conditions and information on the reliability groups ERE and the coefficients of the models. This paper considers methods for the refinement of correction factors to estimate the failure rate of ERE under specified conditions. The results of the work allow us to clarify the existing coefficients for the prevention and reduction of the failure rate of ERE, as well as to prevent failures of elements and systems due to personnel errors.
- ПубликацияТолько метаданныеRecent advances in nuclear power technologies(2020) Samokhin, D.; Самохин, Дмитрий Сергеевич© 2020 Elsevier Inc.The chapter presents the design of nuclear reactors and their appearance, as well as the author’s opinion and fundamental issues arising from these reactors. The focus is on the arrangement of the reactor core. Questions that are examined include how to organize reliable cooling of the reactor core, what main structural measures are required and applied to the physical and technical characteristics of nuclear reactors that define these measures, and so on, without going into a detailed look at the individual elements of the cooling circuit. Methods for the direct conversion of heat into electrical energy are discussed in the chapter.
- ПубликацияТолько метаданныеNuclear reactor safety(2020) Samokhin, D. S.; Самохин, Дмитрий Сергеевич© 2020 Elsevier Ltd All rights reserved.This chapter outlines modern methods for assessing and ensuring the reliability and safety of nuclear power plants (NPP), describes in detail potential hazard factors that may occur during a nuclear accident at a NPP, and what measures to take to prevent such occurrences. Much attention is paid to probabilistic safety analysis (PSA) of NPPs. The principles of the ideology and practice of ensuring and regulating the safety of NPPs are systemically formulated. The types, indicators, and safety analyses currently used are considered, and their classifications are proposed. We will consider the history and principles of nuclear safety including deterministic and probabilistic approach (PSA, LPSA), active-passive safety systems, etc.
- ПубликацияОткрытый доступОпределение показателей надежности оборудования и персонала ядерных объектов по нечетко-вероятностным моделям, учитывающим опыт эксплуатации(ИАТЭ НИЯУ МИФИ, 2012) Самохин, Д. С.; Самохин, Дмитрий Сергеевич; Волков, Ю. В.