Персона: Тихомиров, Георгий Валентинович
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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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Руководитель научной группы "Лаборатория виртуальной реальности в области ядерных технологий"
Руководитель научной группы "Лаборатория инженерного компьютерного моделирования"
Руководитель научной группы "Лаборатория инженерного компьютерного моделирования"
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Тихомиров
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Георгий Валентинович
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Теперь показываю 1 - 10 из 12
- 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.
- ПубликацияТолько метаданныеEvaluation 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
- ПубликацияТолько метаданныеAngular 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
- ПубликацияТолько метаданныеIndependent 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
- ПубликацияТолько метаданныеA 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
- 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 доступNuclear data uncertainty influence on the breeding ratio in sodium-cooled fast reactor systems(2023) Рыжков, Александр Александрович; Ryzhkov, A. A.; Ternovykh, M. Y.; Tikhomirov, G. V.; Терновых, Михаил Юрьевич; Тихомиров, Георгий ВалентиновичNuclear data are a main uncertainty source in neutron transport simulations making their consideration in reactor safety necessary. This arises anew with the state-of-the-art reactors known as Generation IV. Some of the reactors suggests providing the reactivity margin below the effective delayed neutron fraction excluding prompt criticality accidents, and the breeding ratio is the key factor in this. Consequently, assessing a degree of the breeding ratio accuracy is of interest. Therefore, in this work, the breeding ratio uncertainties are analyzed by performing a sensitivity and uncertainty analysis of the MET1000 and MOX3600 models with respect to nuclear data using SCALE. As a result, the breeding ratio uncertainties are obtained approximately equal to 2% as the main contributors are 239Pu(n, γ), 238U(n, γ), and 238U(n, n’ ). The uncertainty sources between the models are compared, and 16O preponderantly increases the total uncertainty not directly by its uncertainty but by its impact on the spectrum.
- 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.
- ПубликацияОткрытый доступNuclear data uncertainty on generation IV fast reactors criticality calculations analysis comparison(2023) Chereshkov, D. G.; Тихомиров, Георгий Валентинович; Терновых, Михаил Юрьевич; Рыжков, Александр Александрович; Tikhomirov, G. V.; Ternovykh, M. Y.; Ryzhkov, A. A.The new calculation code capabilities are applied in the current work as well as important fast reactor criticality parameters uncertainty assessment articles’ results based on different nuclear data libraries and covariance matrices. A comparative analysis of uncertainty estimations related to neutron reactions is presented for lead-cooled reactor models and sodium-cooled reactor models. For the models of advanced BN and BR fast reactors with three fuel types (UO2, MOX, MNUP), the multiplication factor uncertainty calculations are performed using 252-group covariance matrices based on ENDF/B-VII.1 library via the SCALE 6.2.4 code system. The main nuclear data uncertainty contributors in the multiplication factor are determined. Recommendations are formulated for improving the cross sections accuracy for several nuclides in order to provide more reliable results of fast reactor criticality calculations. Lead-cooled reactors have no operational history compared to light-water and sodium-cooled reactors. The experimental data insufficiency calls in the question about reliability of the simulation results and requires a comprehensive initial data uncertainty analysis for the neutron transport simulation. The obtained results support the idea that lead- and sodium-cooled reactors have close nuclear data sensitivity using one and the same computation tools, nuclear data libraries and fuel compositions. This makes it possible to use the accumulated data of benchmarks for sodium-cooled reactors in the safety determination of lead-cooled reactors.