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Article

  • Title

    CALCULATION OF VVER-1000 CORE BAFFLE TEMPERATURE DISTRIBUTION FOR IT'S SWELLING ASSESSMENT

  • Authors

    Filonov V.
    Filonova Y.
    Dubyk Y.
    Bohdan A.

  • Subject

    ENERGETICS. HEAT ENGINEERING. ELECTRICAL ENGINEERING

  • Year 2020
    Issue 1(60)
    UDC 532.5
    DOI 10.15276/opu.1.60.2020.04
    Pages 35-46
  • Abstract

    This paper presents simplified CFD-model of VVER-1000 core baffle cooling to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal, thus a more detailed analysis of the core baf- fle metal thermohydraulic cooling characteristics is required. In this paper, an approach with computational fluid dynamics (CFD) using is proposed. It allows us to consider local hydrodynamic coolant flow characteristics, as well as azimuthal distributions of characteristic parameters. An analytical model was developed to obtain characteristic parameters of simplified CFD-model, and consequently reasonably narrow its limits. Computational model is limited by the height of the baffle and performed using 60-degree symmetry, which included: core, baffle, core barrel, simplified geometry of connecting studs and coolant domains. Core is simplified as a homogeneous body with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model takes into account cooling flow of the coolant through the nuts grooves, which allows obtaining a more realistic temperature field in studs. Calculated convection coefficient and temperature are in good agreement with analytical model, and give a more convenient result comparing to RELAP5. Obtained temperature distributions were used to estimate baffle swelling process. Due to the less conservative results in temperature distribution swelling and creep deformations significantly decreased. Developed model was further improved and used to calculate baf- fle temperature field changes during the representative transitional process of normal operation conditions violations. Results of transient process simulation are used in assessing of progressive form-change calculating need.

  • Keywords CFD-simulation, VVER-1000, core baffle, temperature distribution, material swelling
  • Viewed: 275 Dowloaded: 2
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  • References

    Література

    1. Bцttcher M., KrьЯmann R. Primary loop study of a VVER-1000 reactor with special focus on coolant mix- ing. Nuclear Engineering and Design. 2010. 240, 9. P. 2244–2253. DOI: 10.1016/j.nucengdes.2010.02.044.

    2. Simulation of mixing effects in a VVER-1000 reactor / U. Bieder, G. Fauchet, S. Betin, N. Kolev, D. Popov. Nucl. Eng. Des. 2007. 237. P. 1718–1728. DOI 10.1016/j.nucengdes.2007.02.015.

    3. Best-estimate simulation of a VVER MSLB core transient using the NURESIM platform codes / I. Spasov, S. Mitkov, N. Kolev, et al. Nuclear Engineering and Design. 2017. 321. P. 26–37. DOI: 10.1016/j.nucengdes.2017.03.032.

    4. Three dimensional thermal hydraulic characteristic analysis of reactor core based on porous media method / R. Chen, M. Tian, S. Chen, et al. Annals of Nuclear Energy. 2017. 104. P. 178–190. DOI: 10.1016/j.anucene.2017.02.020.

    5. Thermal analysis of a PWR core internal baffle structure / C. Pйniguel, I. Rupp, N. Lingneau, et al. Pro- ceedings of PVP2006-ICPVT-11-93299. Vancoucer, BC, Canada, July 23–27, 2006.

    6. Filonova Y.S., Filonov V.V., Dubyk Y.R. Reactor baffle cooling CFD framework for swelling assess- ment. Proceedings of the 2018 26th International Conference on Nuclear Engineering – ICONE26- 82365. London, Great Britain, July 22-26, 2018.

    7. IAEA VERLIFE: Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation, 2013.

    8. Ключников А.А., Шараевский И.Г., Фиалко Н.М. Теплофизика безопасности атомных электро- станций: монография. Чернобыль : Ин-т проблем безопасности АЭС, 2010. 484 с.

    9. ANSYS CFX-Solver Theory Guide.

    10. Menter F., CFD Best Practice Guidelines for CFD Code Validation for Reactor-Safety Applications, EVOL-ECORA-D01, 2002.

    11. Мирзов И. В. Особенности расчета напряженно-деформированного состояния выгородки ВВЭР- 1000. UJV Reza.s. Чехия. 2017. №147. С. 1–8.

    12. Марголин Б.З., Мурашова А.И., Неустроев В.С. Влияние напряжений на радиационное распуха- ние аустенитных сталей. Научно – технический журнал «Вопросы материаловедения». 2011. № 4(68). С. 124-139.

    13. IAEA VERLIFE: Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation, 2013.

    14. Нормы расчета на прочность оборудования и трубопроводов для АЭС. Москва : Энергоатомиз- дат, 1989. 525 с.

    References

    1. Bцttcher, M., & KrьЯmann, R. (2010). Primary loop study of a VVER-1000 reactor with special focus on coolant mixing. Nuclear Engineering and Design, 240, 9, 2244–2253. DOI: 10.1016/j.nucengdes.2010.02.044.

    2. Bieder, U., Fauchet, G., Betin, S., Kolev, & N., Popov, D. (2007). Simulation of mixing effects in a VVER-1000 reactor. Nucl. Eng. Des., 237, 1718–1728. DOI: 10.1016/j.nucengdes.2007.02.015.

    3. Spasov, I., Mitkov, S., & Kolev, N., et al. (2017). Best-estimate simulation of a VVER MSLB core transient using the NURESIM platform codes. Nuclear Engineering and Design, 321, 26–37. DOI: 10.1016/j.nucengdes.2017.03.032.

    4. Chen, R., Tian, M., & Chen, S., et al. (2017). Three dimensional thermal hydraulic characteristic analysis of reactor core based on porous media method. Annals of Nuclear Energy, 104, 178-190. DOI: 10.1016/j.anucene.2017.02.020.

    5. Pйnigue,l C., Rupp, I., & Lingneau, N., et al. (2006). Thermal analysis of a PWR core internal baffle structure. Proceedings of PVP2006-ICPVT-11-93299. Vancoucer, BC, Canada, July 23-27.

    6. Filonova, Y.S., Filonov, V.V., & Dubyk, Y.R. (2018). Reactor baffle cooling CFD framework for swelling assessment. Proceedings of the 2018 26th International Conference on Nuclear Engineering – ICONE26-82365. London, GreatBritain, July 22-26.

    7. IAEA VERLIFE: (2013). Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation.

    8. Klyuchnikov, A.A., Sharaevskii, I.G., & Fialko, N.M. (2010). Thermophysics of nuclear power plants safety: monograph. Chernobyl: Institute for NPP Safety Problems.

    9. ANSYS CFX-Solver Theory Guide.

    10. Menter, F. (2002). CFD Best Practice Guidelines for CFD Code Validation for Reactor-Safety Applica- tions, EVOL-ECORA-D01.

    11. Mirzov, I.V. (2017). Features of the VVER-1000 core baffle stress-strain state calculation. UJV Reza.s. Czech. 147, 1–8.

    12. Margolin, B.Z., Murashova, A.I., & Neustroev, V.S. (2011). Effect of stress on radiation swelling of austenitic steels. Voprosy Materialovedeniya, 4(68), 124–139.

    13. IAEA VERLIFE: (2013). Unified Procedure for Lifetime Assessment of Components and Piping in WWER NPPs during Operation.

    14. Strength calculation standards for NPP equipment and pipelines. (1989). Moscow: Energoatomizdat, 525 p.

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