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Published: 2011-06-30
Abstract Views: 62
DOI: https://doi.org/10.53747/nst.v1i2.299
Issue
Vol. 1 No. 2 (2011)
Section
Articles
How to Cite
DUONG , N. H., & NGUYEN , T. T. (2011). THERMAL HYDRAULIC SYSTEM AND NUMERICAL MODELLING OF VVER-1000 NUCLEAR REACTOR IN VIEW OF THE 1ST NINH THUAN NUCLEAR POWER PLANT. Nuclear Science and Technology, 1(2), 15-26. https://doi.org/10.53747/nst.v1i2.299

THERMAL HYDRAULIC SYSTEM AND NUMERICAL MODELLING OF VVER-1000 NUCLEAR REACTOR IN VIEW OF THE 1ST NINH THUAN NUCLEAR POWER PLANT

DUONG NGOC HAI1,2, NGUYEN TAT THANG3,4
1 Institute of Mechanics, Vietnam Academy of Science and Technology (VAST), 264 Doi Can, Ba Dinh, Hanoi ,Vietnam.
2 University of Engineering and Technology, Vietnam National University of Hanoi (VNU), 144 Xuan Thuy, Hanoi, Vietnam,
3 Institute of Mechanics, Vietnam Academy of Science and Technology (VAST), 264 Doi Can, Ba Dinh, Hanoi ,Vietnam. 2)
4 Tokyo Institute of Technology (TITECH),2-12-1 Ookayama, Meguro-ku, Tokyo 152-8550, Japan.

Main Article Content

Abstract

This paper presents an overview of the thermal-hydraulic system of Russian VVER-1000 pressurized water reactors (PWRs) and some results of our study on the numerical simulation using VISA_RELAP5 integrated code. This research has been carried out in view of the 1st nuclear power plant of Vietnam that will use VVER type reactors. It should be noted that knowledge of the VVER reactor system is still lacking in Vietnam now. The data that we used in our research is real data of the Kalinin nuclear power plant (NPP) in Russia. Therefore this research has important practical implications especially for the preparation for the safe operation and proper management of incidents (accidents) for the 1st NPP of Vietnam.

Article Details

Keywords

VVER-1000, Thermal hydraulic, VISA, RELAP5, PWR, Ninh Thuan NPP

References

[1] Vietnam Atomic Energy Institute (2011), Report on an overview of nuclear power technology, MOST.
[2] Vietnam Atomic Energy Institute (2011), Report on the research on the sites of the planned
nuclear power plants in Vietnam, MOST.
[3] C. Allison et al, RELAP5/Mod3 Code Manual, Volumes I-V, NUREG/CR-5535, INEL- 95/0174.
[4] Kim, K.D. et. al. (2001), A visual environment for system analysis codes, Progress in Nuclear Energy, Vol.39, 335-344.
[5] K.D. Kim et. al. (2007), Development of a Visual System Analyzer based on reactor system analysis codes, Progress in Nuclear Energy, Vol.49, 452-462.
[6] W. P. Baek et. al. (2009), Safety assessment of Korean nuclear facilities: current status and
future, Nucl. Eng. Technol., Vol.41, No.4, 391-402.
[7] D. N. Hai (2001), Atomic Energy - Nuclear Safety and Some Problems of Thermal Hydraulics, National Conference on Engineering Mechanics, 68-74.
[8] D. T. Ba (2005), Preliminary study and application of RELAP5 to the calculation and safety analysis of nuclear reactors, Institute of Mechanics.
[9] D. N. Hai et. al. (2005), A model for steady-state thermal hydraulic analysis of nuclear research reactors - COOLOD-N2, The 2005 Annual National Scientific Conference on Fluid Mechanics, 137-146.
[10] N. T. Thang (2006), Application of some thermal hydraulic simulation softwares to the modelling, calculation and simulation of the thermal hydraulic system of the Da Lat nuclear research reactor, Institute of Mechanics.
[11] D. N. Hai and N. T. Thang (2006), A study on the thermal hydraulic calculation and simulation of a simple model of a nuclear power plant using RELAP5, The 2006 Annual National Scientific Conference on Fluid Mechanics, 193-202.
[12] D. N. Hai, D. T. Ba and N. T. Thang (2006), Some results of study of nuclear reactor thermal hydraulics, The National Scientific Conference on Engineering Mechanics and Automation, 85-96.
[13] Hai N. D. and Thang T. N. (2007), Thermal hydraulic calculation of DRR and NPP simple
model using RELAP5, The 4th Asian Specialist Meeting on Future Small-Sized LWR Development, 20-1 to 20-9.
[14] D. N. Hai and N. T. Thang (article in press), Application of VISA and RELAP5 to the modellings, simulations and calculations of thermal hydraulic transients of PWR nuclear power plants, Annual National Scientific Conference on Fluid Mechanics.
[15] D. N. Hai and N. T. Thang (2011), Application of Visa and Relap5 softwares to themodellings, simulations and calculations of the thermal hydraulic system of nuclear power plants with VVER-type reactors, IX National Conference on Nuclear Science and Technology, Ninh Thuan - Vietnam, 18-20 August, 142-147.
[16] Antonia Wenisch (2007), AES-92 for Belene: The Mystery Reactor, Austrian Institute of Ecology, Vienna.
[17] The International Topical Meeting on VVER Safety (1995).
[18] S. Pylev (1999), Assessment Study of RELAP5/MOD3.2 Based on the Kalinin NPP Unit-1
Stop of Feedwater Supply to the Steam Generator No. 4, NUREG/IA-0167.
[19] Yu. G. Dragunov et. al. (2007), Prospects for Development of VVER-type Pressurized Light-Water Reactor Installations, Thermal Engineering, Vol.54, No.5, 343-347.
[20] S.K. Agrawal et. al. (2006), The VVERs at KudanKulam, Nuclear Engineering and Design, Vol.236, 812–835.
[21] I. Spasov et. al. (2009), CATHARE Multi-1D Modeling of Coolant Mixing in VVER-1000 for RIA Analysis, Hindawi Publishing Corporation Science and Technology of Nuclear Installations, Volume 2010, Article ID 457094, doi:10.1155/2010/457094.
[22] Lucia Bonavigo and Mario De Salve, Issues for Nuclear Power Plants Steam Generators, Politecnico di Torino, Italy.
[23] N. B. Trunov et. al. (2008), Steam generators-Horizontal or Vertical (which type should be used in nuclear power plants with VVER?), Atomic Energy, 105, 165-174.
[24] Nikolay Fil (2011), Design, Safety Technology and Operability Features of Advanced VVERs, IAEA Headquarters, Vienna, Austria.
[25] Miles Pomper (2009), The Russian Nuclear Industry: Status and Prospects, Nuclear Energy Futures Paper, No. 3, January.