Article Sidebar

PDF
Published: 2020-06-15
Abstract Views: 100
Views PDF: 29
DOI: https://doi.org/10.53747/jnst.v10i2.32
Issue
Vol. 10 No. 2 (2020)
Section
Articles
How to Cite
Doan Manh, L., Nguyen Van, T., & Tran Chi, T. (2020). Analysis of in-vessel accident progression in VVER1000 NPP during SBO accident with external reactor vessel cooling method. Nuclear Science and Technology, 10(2), 1-14. https://doi.org/10.53747/jnst.v10i2.32

Analysis of in-vessel accident progression in VVER1000 NPP during SBO accident with external reactor vessel cooling method

Doan Manh Long1, Nguyen Van Thai2, Tran Chi Thanh3
1 Nuclear Training Center, 140 Nguyen Tuan, Thanh Xuan, Hanoi
2 Department of Nuclear Engineering and Environmental Physics, School of Engineering Physics, Hanoi University of Science and Technology, 01 Dai Co Viet, Hai Ba Trung, Hanoi
3 Viet Nam Atomic Energy Institute, 59 Ly Thuong Kiet, Hoan Kiem, Hanoi

Main Article Content

Abstract

In this study, the MELCOR v1.8.6 code was utilized to perform an analysis of the in-vessel accident progression in VVER1000 reactor during the Station Black-Out (SBO) accident with and without external reactor vessel cooling (ERVC) strategy. The analysis presented the predictions of the main phenomena during the accident such as failure of fuel cladding, collapse of lower core support plate, relocation of core debris to lower plenum and mass of debris components in lower plenum, and provided comparisons between two cases in term of main parameters such as integrity time of reactor and structure components of molten pool. These parameters are very important inputs for further research on the application of external vessel cooling strategy for VVER1000 reactor.

Article Details

Keywords

External Reactor Vessel Cooling, SBO, VVER1000, In-Vessel Melt Retention, Severe Accident

References

[1]. Tuomisto, H., and Theofanous, T.G. “A consistent approach to severe accident management”. Nuclear Engineering and Design 148, 171-183, 1994.
[2]. Kyma ̈la ̈inen, O., et al. “In-vessel retention of corium at the Loviisa plant”. Nuclear Engineering and Design, 169, 109-130, 1997.
[3]. Theofanous, T.G., et al. “In-vessel cool-ability and retention of a core melt”, DOE/ID-10460, Volume 1, Revised October, 1996.
[4]. Esmaili, H., and Khatib-Rahbar, M. “Analysis of in-vessel retention and ex-vessel fuel coolant interaction for AP1000”. Energy Research, Inc., ERI/NRC 04-21, NUREG/CR-6849, 2004.
[5]. Rempe, J.L., et al. “In Vessel Retention strategy for higher power reactors”. Final Report, INEEL/EXT-04-02561, 2005.
[6]. Ji Xing et al. “HPR1000: Advanced pressurized water reactor with Active and Passive safety”. Engineering 2 79-87, 2016.
[7]. Theofanous, T.G., et al. “In-vessel cool-ability and retention of a core melt”, DOE/ID-10460, Volume 1, Revised October, 1996.
[8]. Esmaili, H., et al. “An Assessment of Ex-Vessel Steam Explosions in the AP600 Advanced Pressurized Water Reactor”, Energy Research, Inc., ERI/NRC 95-211, 1996.
[9]. Bui and Dinh. “Modeling of heat transfer in heated-generating liquid pools by an effective diffusivity-convectivity approach”, In: Proceedings of 2nd European Thermal-Sciences Conference, Rome, Italy, pp. 1365–1372, 1992.
[10]. Tran, C.T., and Dinh, T.N. “The effective convectivity model for simulations of melt pool heat transfer in a light water reactor pressure vessel lower head”. Part I&II. Progress in Nuclear Energy 51, 849–871, 2009.
[11]. Dombrovskii, L.A., et al. “Numerical Simulation of the Stratified-Corium Temperature Field and Melting of the Reactor Vessel for a Severe Accident in a Nuclear Power Station”. Thermal Engineering, Vol. 45, No. 9, pp. 775-765, 1998.
[12]. Yue Jin et al. “In-and ex-vessel coupled analysis of IVR-ERVC phenomenon for large scale PWR”. Annals of Nuclear Energy 80 322-337, 2015.
[13]. SANGIORGI Marco et al. “In-Vessel Melt Retention Analysis of a VVER1000 NPP”. EUR 27951; doi 10.2790/62596, 2016.
[14]. Polina Tusheva et al. “Analysis of severe accidents in VVER1000 reactors using integral code ASTEC”. Proceeding of the 17th International Conference on Nuclear Engineering, Brussels, Belgium, 2009.
[15]. Sandia National Laboratories, “MELCOR computer code manuals”, Ver.1.8.6, Rev.3 vol. NUREG/CG 6119, 2005.
[16]. Belon, S., et al. “Insight of core degradation simulation in integral codes throughout ASTEC/MELCOR crosswalk comparisons and ASTEC sensitivities studies”. The 8th European Review Meeting on Severe Accident Research, Warsaw, Poland, 2017.