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Article

  • Title

    PRINCIPLES OF REPRESENTATIVE THERMOHYDRAULICAL SCENARIOS CHOOSING USED FOR STRENGTH SIMULATION OF NPP UNITS

  • Authors

    Maslo Oleksandr
    Bulakh Pavlo
    Chyzhyk Ganna
    Shirokov Andrii
    Byalonovich Andrii

  • Subject

    ENERGETICS. HEAT ENGINEERING. ELECTRICAL ENGINEERING

  • Year 2019
    Issue 3(59)
    UDC 539.4
    DOI 10.15276/opu.3.59.2019.02
    Pages 14-18
  • Abstract

    Safe operation of main equipment of nuclear power plants requires periodic estimation of such factors as size of accumulated fatigue damages of elements constructions and predicted value of fatigue damage on rated period of future operation. One of the most important stages of nuclear power plant resource prolongation is analysis of thermomechanical influence of the operational regimes on the strength of the elements and its units. Herewith for one of the element operational regimes a couple of several variations are performed, each of which is characterized by its unique set of thermomechanical parameters, such as pressure, temperature, velocity of coolant consumption, etc. A huge amount of such variations can increase of calculation time and software requirements. The main goal of this research is decreasing of amount of thermohydraulic scenarios justification of possible operational regimes of main equipment of nuclear power plants that are subjected by full analysis during possible amount of fatigue damages determination. This research describes a possibility of decreasing amount of required simulations using reasonable exclusion of unrepresentative variants on the basis of mathematic based comparison. This paper described main principles of thermohydraulical scenarios while using this methodic for strength simulation of pressure compensator of nuclear power plants with WWER-1000. The numerical parameters for estimation on typical scenarios diagrams are performed, and it gives a possibility to determine reasonable scenarios from groups that are close by their character of simulated processes. The accuracy of obtained data during reasonable thermohydraulical scenarios is at satisfactory level according to its total analysis, and it gives possibility to recommend the pro-posed approach for usage.

  • Keywords resource, thermohydraulical regimes, principles of scenarios choosing
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  • References

    Література

    1. Фадеев И.Д., Дмитриева И.В., Рогожкин С.А., Шепелев С.Ф. Обобщение опыта пуска РУ БН-800 для обоснования теплогидравлических параметров в режимах нормальной эксплуатации. Безо-пасность, эффективность и экономика атомной енергетики: пленарные и секционные док. ХІ междунар. науч.-тех. конф., г.Москва, 23–24 мая 2018 г. Москва, 2018. С. 587–590.

    2. Численное моделирование теплогидравлических процессов в верхней камере быстрого реактора / С.А. Рогожкин, С.Л. Осипов, И.Д. Фадеев, С.Ф. Шепелев, А.А. Аксенов, С.В. Жлуктов, М.Л. Са-зонова, В.В. Шмелев. Атомная энергия. 2013. Т. 115, Вып. 5. С. 295–298.

    3. Расчетный комплекс для задач обоснования безопасности аэс при запроектных и тяжелых авари-ях / А.Е. Киселев, К.С. Долганов, Д.Ю. Томащик, Р.И. Бакин, А.А. Киселев, С.Н. Красноперов. Безопасность, эффективность и экономика атомной енергетики: пленарные и секционные док. ХІ междунар. науч.-тех. конф., г.Москва, 23–24 мая 2018 г. Москва, 2018. С. 446–449.

    4. Yuchuan Guo, Guanbo Wang, Dazhi Qian, Heng Yu, Bo Hu, Simao Guo, Xiangmiao Mi, Jimin Ma. Accident safety analysis of flow blockage in an assembly in the JRR-3M research reactor using system code RELAP5 and CFD code FLUENT. Annals of Nuclear Energy. 2018, Vol. 122, P. 125–136.

    5. Li Lei, Zhang Zhijian. Development of Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor 18th International Conference on Nuclear Engineering: Volume 2 Xi’an, China, May 17–21, 2010, P. 497–505.

    6. Daeseong Jo, Jonghark Park, Heetaek Chae. Development of thermal hydraulic and margin analysis code for steady state forced and natural convective cooling of plate type fuel research reactors. Progress in Nuclear Energy. 2014, Vol. 71, P. 39–51.

     

    References

    1. Fadeev, I.D., Dmitrieva, I.V., Rogozhkin, S.A., & Shepelev, S.F. (2018). Generalization of the start-up experience of the BN-800 RP to substantiate thermal-hydraulic parameters in normal operation. Safety, efficiency and economics of nuclear energy: plenary and sectional docs. XI Intern. scientific-tech Conf., (pp. 587–590). Moscow.

    2. Rogozhkin, S.A., Osipov, S.L., Fadeev, I.D., Shepelev, S.F., Aksenov, A.A., Zhluktov, S.V., Sazonova, M.L., & Shmelev, V.V. (2013). Numerical simulation of thermo-hydraulic processes in the upper chamber of the fast reactor. Atomic Energy, 115, 5, 295–298.

    3. Kiselev, A.E., Dolganov, K.S., Tomaschik, D.Yu., Bakin, R.I., Kiselev, A.A., & Krasnoperov, S.N. (2018). Calculation complex for the problems of justifying the safety of NPP in case of beyond design basis and severe accidents. Safety, efficiency and economics of nuclear power engineering: plenary and sectional docks. XI Intern. scientific-tech Conf. (pp. 446–449). Moscow.

    4. Yuchuan Guo, Guanbo Wang, Dazhi Qian, Heng Yu, Bo Hu, Simao Guo, Xiangmiao Mi, & Jimin Ma. (2018). Accident safety analysis of flow blockage in an assembly in the JRR-3M research reactor using system code RELAP5 and CFD code FLUENT. Annals of Nuclear Energy, 122, 125–136.

    5. Li Lei, & Zhang Zhijian. (2010). Development of Thermal-Hydraulic Analysis Code for Plate Type Fuel Reactor, 18th International Conference on Nuclear Engineering: Volume 2 Xi’an, (pp. 497–505). China.

    6. Daeseong Jo, Jonghark Park, & Heetaek Chae. (2014). Development of thermal hydraulic and margin analysis code for steady state forced and natural convective cooling of plate type fuel research reactors Progress in Nuclear Energy, 71, 39–51.

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