Персона:
Терновых, Михаил Юрьевич

Научные группы

Научная группа

Лаборатория инженерного компьютерного моделирования (Кафедра №5)

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.

Фамилия

Терновых

Имя

Михаил Юрьевич

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Теперь показываю 1 - 10 из 19

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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.
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Neutron-physical characteristics modeling of light water reactors critical corium slurry in water
(2020) Ternovykh, M. Y.; Abdumalikov, H. M.; Ryzhov, S. N.; Терновых, Михаил Юрьевич; Рыжов, Сергей Николаевич
© Published under licence by IOP Publishing Ltd.In this paper, the idea of modifying the benchmark by increasing effective multiplication factor value in the considered system without increasing the number of particles under consideration and fuel enrichment is discussed. The technology of data transmission, processing and comparison of the results of neutron-physical calculations using three modern codes developed in different countries and implementing the Monte Carlo method has been worked out. For the corium and water mixture parameters search, the method conventionally called the gradient descent method is used. The search method for areas of critical state corium slurry in water is described. This method can be used to justify nuclear safety in the corium extraction and transportation processes. The methodology is based on the combined use of 1-D and 3-D criticality calculations capabilities of the SCALE 6.2 program pack. Fall of corium particles in water simulation benchmark version is formulated. The possibility of using the algorithm to find the parameters of a corium and water mixture is demonstrated. This benchmark includes the critical state of corium slurry in water and assumes the use of regular structures in the formation of geometric models. The proposed version of the critical benchmark for the corium particles in water state contains 55% fuel.
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Features of training specialists for responsibility centers of the project direction "proryv"
(2020) Kutsenko, K. V.; Lavrukhin, A. A.; Pershukov, V. A.; Ternovykh, M. Yu.; Tikhomirov, G. V.; Куценко, Кирилл Владленович; Лаврухин, Алексей Анатольевич; Першуков, Вячеслав Александрович; Терновых, Михаил Юрьевич; Тихомиров, Георгий Валентинович
© Published under licence by IOP Publishing Ltd.National Research Nuclear University "MEPhI"(NRNU MEPhI) is the leading university of the consortium of reference universities of the State Atomic Energy Corporation Rosatom (ROSATOM), a regular supplier of young specialists of the highest level. Since 2014 NRNU MEPhI has been training graduates for responsibility centers of the project direction "Proryv". University responds quickly to staffing challenges, and provides the project "Proryv"the personnel with relevant unique competencies. These competencies include the ability to apply the knowledge gained to develop a new technological platform for nuclear power with the involvement of uranium-238 and spent nuclear fuel reprocessing products in the fuel cycle. Obtaining such unique knowledge is impossible without the use of external resources, namely, the involvement of leading scientists and practitioners of ROSATOM for conducting classes with students, organizing practices in scientific centers and at nuclear fuel cycle enterprises. The article discusses the features of the organization of the educational process in this area.
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Investigation of radiation levels in the region of fuel-transfer cask for transportation of spent fuel of light-water reactors
(2020) Gerasimov, A. S.; Shorov, V. Y.; Ternovykh, M. Y.; Терновых, Михаил Юрьевич
© Published under licence by IOP Publishing Ltd.The paper formulates the main problems associated with research on transmutation, which should be paid attention to by today's young researchers. The processes of production of hazardous nuclides during transmutation in reactor facilities are considered. The goals of transmutation and the choice of nuclides to be transmuted are discussed. The concept of radiotoxicity is explained as a measure of the radiological hazard of radioactive nuclides, based on the maximum permissible concentration of nuclides according to the IAEA standards. The problem of the formation of secondary radioactive nuclides in nuclear fuel during generation of neutrons for transmutation is discussed. The advantages and disadvantages of various methods of transmutation in nuclear installations are considered: inclusion of transmutable nuclides in nuclear fuel in fast reactors, transmutation in specialized thermal and fast transmutation reactor installations and ADS systems. The problem of accumulation of highly radioactive actinides in a transmutation facility during long-term transmutation and the problem of a potential hazard of the transmutation facility itself are discussed. The unacceptability of application of common-type power reactors for the transmutation of long-lived fission products is demonstrated.
Только метаданные
Evaluation of technological uncertainties using the sensitivity to nuclear data
(2024) Ryzhkov, A. A.; Tikhomirov, G. V.; Ternovykh, M. Y.; Gerasimov, A. S.; Рыжков, Александр Александрович; Тихомиров, Георгий Валентинович; Терновых, Михаил Юрьевич
Только метаданные
Angular distribution uncertainty influence in a large sodium-cooled fast reactor with mixed-oxide fuel
(2024) Ryzhkov, A. A.; Tikhomirov, G. V.; Ternovykh, M. Y.; Рыжков, Александр Александрович; Тихомиров, Георгий Валентинович; Терновых, Михаил Юрьевич
Только метаданные
Independent testing of new generation codes of the "Proryv" project
(2021) Suslov, I. R.; Tikhomirov, G. V.; Ternovykh, M. Y.; Khomyakov, Y. S.; Тихомиров, Георгий Валентинович; Терновых, Михаил Юрьевич; Хомяков, Юрий Сергеевич
Только метаданные
A review of the current nuclear data performance assessments in advanced nuclear reactor systems
(2024) Ryzhkov, A. A.; Tikhomirov, G. V.; Ternovykh, M. Y.; Рыжков, Александр Александрович; Тихомиров, Георгий Валентинович; Терновых, Михаил Юрьевич
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Testing the multigroup, group and subgroup options of the CONSYST / ABBN-RF system on criticality calculations of fast reactor models with MNUP fuel
(2020) Ternovykh, M. Y.; Bogdanova, E. V.; Терновых, Михаил Юрьевич; Богданова, Екатерина Владимировна
© Published under licence by IOP Publishing Ltd.The unified code CONSYST-2020 for neutron constants preparation and the ABBN-RF-2020 neutron data library are being testing as part of the new generation codes development for neutronic calculations of fast reactors with mixed nitride uranium-plutonium fuel (MNUP). Testing was carried out on prototype models of fast reactors with MNUP fuel. The errors of the multigroup, group and subgroup approximation are analyzed in comparison with the calculations performed using point-wise cross-sections libraries. The results obtained with point-wise cross-sections libraries are accepted in the final cross-verification as reference results. An assessment was made of the influence of approximations associated with averaging and preparing group cross-sections, and methodological errors determined by the selected spatial and angular computational grids. It was shown that the transition to the direct use of 299-group blocked constants reduces the error to 0.1 - 0.2%. The assessment of the efficiency of using the subgroup approximation shows the possibility of reducing the constant component of the error below 0.05%.
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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.