Персона: Пугачев, Павел Александрович
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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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Павел Александрович
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- ПубликацияТолько метаданныеNumerical simulation of plutonium measurements using an active well coincidence counter (AWCC)(2024) Vladimirov,D.A.; Rogozhkin, V. Yu.; Gorbunova, A. Yu.; Aleeva, T. B.; Pugachev, P. A.; Алеева, Татьяна Борисовна; Пугачев, Павел Александрович
- ПубликацияТолько метаданныеПрограммная реализация и проверка эффективности метода переменных направлений решения трехмерного уравнения диффузии(2021) Пугачев, П. А.; Пугачев, Павел Александрович; Сироткин Алексей Михайлович
- ПубликацияТолько метаданныеVirtual analog of uranium-water subcritical assembly(2022) Kiryukhin, P. K.; Romanenko, V. I.; Khomyakov, D. A.; Shcherbakov, A. A.; Pugachev, P. A.; Yushin, I. M.; Ashraf, O.; Tikhomirov, G. V.; Кирюхин, Павел Константинович; Романенко, Владислав Игоревич; Хомяков, Дмитрий Андреевич; Щербаков, Александр Антонович; Пугачев, Павел Александрович; Тихомиров, Георгий Валентинович© 2022 Elsevier LtdVirtual reality (VR) technology is now being adopted in many industries, including entertainment, medicine, science, and engineering. In the nuclear field, the primary purposes of VR are: reducing radiation dose rates, security of nuclear facilities, visualization of physical processes, and training of personnel. Additionally, VR is a much cheaper alternative to expensive and license-requiring experimental nuclear facilities. This work focuses on reconstructing the workroom with the Uranium-Water Subcritical Assembly (UWSA) located at the National Research Nuclear University MEPhI to determine the optimal uranium–water ratio associated with this assembly in virtual reality. The creation of the virtual analog using Unreal Engine 4 was introduced to integrate the physical model into the virtual environment. The neutronic model of the UWSA was obtained by the MCU code. A similar model was generated by the Serpent code for verification purposes. Additional functions such as neutron flux visualization, radiation dose rate distribution visualization, and dose accumulation mechanics were introduced into the project to improve the quality of education. Visualization of both neutron flux in the assembly and gamma radiation distribution in the workroom was performed using particle systems and volumetric fog based on calculated and experimental data. Operating experience feedback was introduced to prevent or minimize difficulties that may occur in the future by learning from events that have already occurred.
- 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 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
- ПубликацияОткрытый доступ"ИНТЕРАКТИВНАЯ ЛАБОРАТОРНАЯ РАБОТА В ВИРТУАЛЬНОЙ РЕАЛЬНОСТИ "ПУСК РЕАКТОРА ИРТ МИФИ"(НИЯУ МИФИ, 2023) Пугачев, П. А.; Тихомиров, Г. В.; Кирюхин, П. К.; Григорьев, Е. В.; Щербаков, А. А.; Романенко, В. И.; Хомяков, Д. А.; Минаев, Е. В. ; Чернов, Е. В.; Романенко, Владислав Игоревич; Чернов, Евгений Владимирович; Тихомиров, Георгий Валентинович; Кирюхин, Павел Константинович; Хомяков, Дмитрий Андреевич; Щербаков, Александр Антонович; Пугачев, Павел АлександровичПрограмма предназначена для обучения студентов основам обращения с экспериментальными реакторами на примере операции пуска. Лабораторная работа выполнена в виде интерактивного приложения в виртуальной реальности, воссоздающего опыт работы на установке-прототипе - реакторе ИРТ МИФИ. Лабораторная работа включает окружение ИРТ МИФИ, полностью функциональный пульт управления реактором и математические модели, нейтронно-физические и теплофизические, обеспечивающие моделирование процессов инженерной точности. Тип ЭВМ: IBM PC-совмест. ПК; ОС: Windows 10 и выше.
- ПубликацияОткрытый доступDevelopment of virtual analogues of nuclear facilities in virtual reality(2020) Dashanova, E. A.; Zadeba, E. A.; Kiryukhin, P. K.; Pugachev, P. A.; Romanenko, V. I.; Tikhomirov, G. V.; Khomyakov, D. A.; Shcherbakov, A. A.; Yushin, I. M.; Дашанова, Екатерина Александровна; Задеба, Егор Александрович; Кирюхин, Павел Константинович; Пугачев, Павел Александрович; Романенко, Владислав Игоревич; Тихомиров, Георгий Валентинович; Хомяков, Дмитрий Андреевич; Щербаков, Александр Антонович© Published under licence by IOP Publishing Ltd.Using virtual reality technology - a modern trend. The nuclear industry is no exception. This article provides an overview of mathematical models used to create virtual analogue of critical assembly Godiva in virtual reality. Godiva - there is a simple example that allows to hone techniques for creating more complex virtual analogues of nuclear reactors and nuclear facilities. Mathematical models include stationary and dynamic ones. The stationary model is based on data from calculations carried out using Monte Carlo programs such as MCU, Serpent and Geant4. An approach is also described that makes it possible to calculate the reverse multiplication from the values of the effective multiplication factor for various states of the subcritical assembly. The dynamic model allows one to calculate the neutron-physical characteristics of the supercritical assembly during fast processes such as a neutron burst. In conclusion, there are other examples of virtual analogs created using similar approaches.
- 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 доступAnalysis of the methods for group constants generation for calculation of a large SFR core using Serpent 2 and CriMR codes(2020) Gerasimov, A. S.; Akpuluma, D. A.; Smirnov, A. D.; Pugachev, P. A.; Tikhomirov, G. V.; Смирнов, Антон Дмитриевич; Пугачев, Павел Александрович; Тихомиров, Георгий Валентинович© Published under licence by IOP Publishing Ltd.This work aimed at generating homogenized group constants using the Serpent code and then using the CriMR diffusion code to model the large SFR OECD 3600 MWth MOX core. The results were compared with a full core reference Monte Carlo solution by Serpent. Reactivity feedback parameters were also considered. Generating the group constants from separate fuel assemblies allows for simultaneously carrying out calculations and then using the results as input in diffusion codes rather than waiting so long for a 3D full core Monte Carlo calculation to be completed. From the results of the integral parameters we see a close agreement in the calculation codes. The differences can be attributed to the errors that could arise from generating the constants from individual sub-assemblies. The differences in the underlying physics and approximations used in development of the codes could also be a factor. Another way the errors could be reduced is by checking to see that the sub-assembly configurations used in the non-multiplying zones are as close as possible to the real layout in a full 3D core.
- 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.