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Published: 2014-06-30
Abstract Views: 104
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DOI: https://doi.org/10.53747/jnst.v4i2.221
Issue
Vol. 4 No. 2 (2014)
Section
Articles
How to Cite
Trinh , H. T., Tran , K. T., Tran , T. D., & Cao , D. L. (2014). Preliminary test results of anuclear pressurized thermal hydraulictesting facility . Nuclear Science and Technology, 4(2), 7-14. https://doi.org/10.53747/jnst.v4i2.221

Preliminary test results of anuclear pressurized thermal hydraulictesting facility.

Trinh Huu Toan1, Tran Kim Tuan1, Tran Thuy Duong1, Cao Duc Luong2
1 Nuclear Power Engineering Group, Hanoi University of Science and Technology, 01 Dai Co Viet Road, Hanoi ,Vietnam
2 Nuclear Power Engineering Group, Hanoi University of Science and Technology, 01 Dai Co Viet Road, Hanoi ,Vietnam.

Main Article Content

Abstract

In this work, the main features of a nuclear pressurizedthermal hydraulic testing facility named BKTF (BK testing facility) are presented. The facility consists of major systems and components similar to the standard PWR prototype, and it is a volume and pressure scaled model with the total mass of primary coolant in order of hundreds of kilograms. The model is designed to work at normal temperature and pressure less than 160 0C and 10 bars respectively. In this part of research, preliminary results of cold test and hot test are presented. These results show that the system is qualified enough to work under designed pressure and temperature. It is seen that the similarity between temperature and pressure trend lines with these from a real power system during starting up process. In addition, natural circulation was appreciated and compared with forced circulation during the cooling down process and this may help us to have a better vision on passive safety and completely recognize about safety system of nuclear power plant.

Article Details

Keywords

nuclear thermal hydraulic, pressurized water reactor, testing facility, LOCA, cold test, hot test, start up, cooling down natural circulation

References

1. Hsu Y.Y , Z.Y. Wang, Di Marzo, K. Almenas. Nuclear Engineering and Design Vol. 122, p.175-177 (1990).
2. Neil Todreas. Nuclear System I, Fundamental of Nuclear Thermal Hydraulics. p. 192. Hemi-sphere Publishing Corp. (1990).
3. Rohsenow, W.M “A method of correlating Heat Transfer for Surface Boiling Liquids” Trans. ASME, vol. 74, p.696 (1952).
4. Holman J.P. Heat Transfer, p 522. McGraw-Hill Publishing. 10th Edition, (2010).
5. Robert W. Fox, Alan T. McDonald. Introduc-tion to Fluid Mechanics, p. 12. John Willey and Sons. Fourth Edition (1994).
6. Takahashi. Lecturer for graduate students on Nuclear Thermal Hydraulics. Laboratory of Nuclear Research Reactor. Tokyo Institute of Technology 2012.
7. Nahavandi et al, Scaling laws for modeling nuclear reactor System. Nuclear Science and Engineering. Vol.79 p. 75-83 (1979).
8. D’Auria F, Galassi G. M. Scaling in Nuclear Reactor System Thermal Hydraulics. Nuclear Engineering and Design Vol. 240 pa.3268-3269 (2010).
9. Sadik Kakac. Nuclear Steam Generators and Waste Heat Boilers, p. 475-478. John Wiley and Sons. (1991).
10. Toan T.H et al. Design and construction o PWR testing facility, p.265-275. Proceedings of Vietnam-Japan Research/HRD Forum on Nu-clear Technology (2013).
11. Sumio Fuji. Operation of Nuclear Power Plant. Nuclear Steam Supply System and Electric Power Generation. GEN-HSW-10096 (Mitsubishi) p.18-23 (2010).