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

Полная страница элемента

Результаты поиска

Теперь показываю 1 - 10 из 19

Открытый доступ
Nuclear 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.
Только метаданные
Analysis of Methods and Technologies for Assessing the Composition of the Corium Formed as a Result of the Accident at the Fukushima Daiichi NPP
(2022) Ryzhov, S. N.; Bogdanova, E. V.; Ryzhkov, A. A.; Pugachev, P. A.; Tikhomirov, G. V.; Ternovykh, M. Y.; Aleeva, T. B.; Рыжов, Сергей Николаевич; Богданова, Екатерина Владимировна; Рыжков, Александр Александрович; Пугачев, Павел Александрович; Тихомиров, Георгий Валентинович; Терновых, Михаил Юрьевич; Алеева, Татьяна Борисовна
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
Estimation of the Duration of Cooling Time of the SFA of the VVER-1200 Reactor Depending on the Type of Transport Container
(2020) Demin, V. M.; Savander, V. I.; Ternovykh, M. Y.; Abu, Sondos, M. A.; Демин, Виктор Максимович; Савандер, Владимир Игоревич; Терновых, Михаил Юрьевич
© 2020, Pleiades Publishing, Ltd.Abstract: As the burnup increases, the requirements of nuclear and radiation safety for spent nuclear fuel (SNF) at the subsequent stages of operation grow. The analysis and estimation of the required cooling time of the spent fuel assemblies (SFAs) of the VVER-1200 reactor for transportation in various types of transport containers (TUK-13 and TUK-141O) are performed. The estimates are based on the analysis of residual energy release and gamma-radiation intensity of SNF depending on the cooling time for different burnups. The data on the absorbed dose rate of neutron and gamma radiation from SNF after 4 and 5 yr of cooling are presented.
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
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
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
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
Nuclear 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 Валентинович
Nuclear 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.