Персона: Богданова, Екатерина Владимировна
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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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Екатерина Владимировна
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- ПубликацияТолько метаданныеAnalysis 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
- ПубликацияТолько метаданныеDevelopment of a test task for coupled neutronic and thermal-hydraulic analysis of LFR fuel cell(2025) Chubarov, M.; Zakharov, M.; Bogdanova, E.; Smirnov, A.; Ouen, D.; Tikhomirov, G.; Чубаров, Максим Алексеевич; Захаров, Михаил Юрьевич; Богданова, Екатерина Владимировна; Смирнов, Антон Дмитриевич; Оуэн, Дэниэл Юджин; Тихомиров, Георгий Валентинович
- 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 метаданныеTest Results of Variance Reduction Techniques Applied to the Deep Penetration Problem(2022) Bogdanova, E. V.; Tikhomirov, G. V.; Богданова, Екатерина Владимировна; Тихомиров, Георгий Валентинович
- ПубликацияОткрытый доступTesting 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%.
- ПубликацияОткрытый доступThe concerted calculation of the BN-600 reactor for the deterministic and stochastic codes(IOP Publishing Ltd, 2017) Bogdanova, E. V.; Kuznetsov, A. N.; Богданова, Екатерина ВладимировнаThe solution of the problem of increasing the safety of nuclear power plants implies the existence of complete and reliable information about the processes occurring in the core of a working reactor. Nowadays the Monte-Carlo method is the most general-purpose method used to calculate the neutron-physical characteristic of the reactor. But it is characterized by large time of calculation. Therefore, it may be useful to carry out coupled calculations with stochastic and deterministic codes. This article presents the results of research for possibility of combining stochastic and deterministic algorithms in calculation the reactor BN-600. This is only one part of the work, which was carried out in the framework of the graduation project at the NRC "Kurchatov Institute" in cooperation with S. S. Gorodkov and M. A. Kalugin. It is considering the 2-D layer of the BN-600 reactor core from the international benchmark test, published in the report IAEA-TECDOC-1623. Calculations of the reactor were performed with MCU code and then with a standard operative diffusion algorithm with constants taken from the Monte - Carlo computation. Macro cross-section, diffusion coefficients, the effective multiplication factor and the distribution of neutron flux and power were obtained in 15 energy groups. The reasonable agreement between stochastic and deterministic calculations of the BN-600 is observed.
- ПубликацияТолько метаданныеOpportunities for US-Russian collaboration on the safe disposal of nuclear waste(2021) Tracy, C. L.; Park, S.; Plevaka, M.; Bogdanova, E.; Богданова, Екатерина ВладимировнаThe United States and Russia both possess large quantities of nuclear waste, generated during the production of nuclear energy and nuclear weapons. To ensure that this radioactive material remains safely sequestered for tens of thousands of years or more, both countries plan to bury it in deep geologic repositories. However, US and Russian repository design strategies are highly distinct. For example, Russia plans to dilute waste in aluminophosphate glass, package waste in stainless steel containers, and bury waste in hard, crystalline granite gneiss rock. The US approach includes the use of borosilicate glass, multi-component superalloy containers, and porous volcanic tuff or highly-plastic bedded salt. The relative efficacies of these design choices remain uncertain. This represents a unique opportunity for applied, comparative study of various natural and engineered barriers to the release of radioactive materials. US-Russian collaboration and sharing of data on repository performance could provide a better technical basis for the long-term immobilization of nuclear waste.
- ПубликацияОткрытый доступVisualization 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 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 доступАлгоритмы неаналогового моделирования Монте-Карло в расчетах ядерных реакторов(НИЯУ МИФИ, 2023) Богданова, Е. В.; Богданова, Екатерина Владимировна; Тихомиров, Г. В.