Evaluation of weight of coolant in the reactor plant during full loss of feeding coolant

Authors

  • Юрий Константинович Тодорцев Odessa National Polytechnic University Prospect Shevchenko, 1, Odessa, Ukraine, 65044, Ukraine
  • Евгения Александровна Кокол Odessa National Polytechnic University Prospect Shevchenko, 1, Odessa, Ukraine, 65044, Ukraine
  • Марк Витальевич Никольский Odessa National Polytechnic University Prospect Shevchenko, 1, Odessa, Ukraine, 65044, Ukraine

DOI:

https://doi.org/10.15587/2312-8372.2013.19542

Keywords:

pressurized water reactor 1000, the mass of a water-steam mixture, emergency mode

Abstract

Despite the complete assurance of specialists in the field of nuclear energy that the safety of nuclear power plants is enough high, this is still a concern for many people. Therefore, this article first presents a method of determining the mass of steam-water mixture in a pressurized water reactor in emergency conditions that will result more rapid reaction of the staff on the situation. The article presents the idea of the method, which consists in the fact that knowing the full weight of the reactor with all the internals and coolant, due to the strain gauge measuring transducers will be known how the weight of the reactor changes and it will be easy to say the proportion of vapor and liquid in it at the time of an emergency. Also the results of the pilot project are described, shown in Figure 2, which can form the basis of realization and implementation of the proposed method for improving nuclear power plants` safety.

Author Biographies

Юрий Константинович Тодорцев, Odessa National Polytechnic University Prospect Shevchenko, 1, Odessa, Ukraine, 65044

Doctor of Technical Sciences, Professor

Department of automation of heat power processes

Евгения Александровна Кокол, Odessa National Polytechnic University Prospect Shevchenko, 1, Odessa, Ukraine, 65044

Master

Department of automation of heat power processes

Марк Витальевич Никольский, Odessa National Polytechnic University Prospect Shevchenko, 1, Odessa, Ukraine, 65044

Postgraduate of Department of automation of heat power processes

References

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  12. Pelykh, S. N., Maksimov, M. V. (2011). Cladding rupture life control methods for a power-cycling WWER-1000 nuclear unit. Nuclear Engineering and Design, Vol. 241, 2956-2963.
  13. Pelykh, S. N., Maksimov, M. V., Parks, G. T. (2013). A method for VVER-1000 fuel rearrangement optimization taking into account both fuel cladding durability and burnup. Nuclear Engineering and Design, Vol. 257, 53-60.
  14. Pelykh, S. N., Maksimov, M. V., Baskakov, V. E. (2013). Grounds of VVER-1000 fuel cladding life control. Annals of Nuclear Energy. Iss, 58, 188-197.
  15. Kim, S. , Jun, H.-Y. (2013). Analysis of an ATLAS 6-in. cold-leg break simulation with MARS code. Annals of Nuclear Energy, 63, 268-275.
  16. Yang, J. , Choi, S.-W., Lim, J., Lee, D.-Y., Rassame, S., Hibiki, T., Ishii, M. (2012). Assessment of performance of BWR passive safety systems in a small break LOCA with integral testing and code simulation. Nuclear Engineering and Design, 247, 128-135.
  17. Leyer, S. , Wich, M. (2012). The Integral Test Facility Karlstein. Science and Technology of Nuclear Installations, Article number 439374.
  18. Lakehal, D., Labois, M. (2011). A New modelling strategy for phase-change heat transfer in turbulent interfacial two-phase flow. International Journal of Multiphase Flow, 37, 627-639.
  19. Kim, H. , Seung-hyun, B. (2010). A study on accuracy improvement in measuring liquid level inside pressurized vessels. Transactions of the Korean Institute of Electrical Engineers, 59, 1889-1893.
  20. Wang, X.-L., Cai, Q., Chen, Y.-Q. (2013). Study on pressurizer water level signal reconstruction based on support vector regression. Atomic Energy Science and Technology, 47, 1003-1007.
  21. Jang, G.-S., Suh, S.-M., Kim, S.-K., Suh, Y.-S., Park, J.-Y.(2013). A proactive alarm reduction method and its human factors validation test for a main control room for SMART. Annals of Nuclear Energy, 51, 125-134.
  22. Shin, Y. C. (2011). APR1400 man-machine interface system. Nuclear Engineering International, 56, 12-16.

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Published

2013-11-28

How to Cite

Тодорцев, Ю. К., Кокол, Е. А., & Никольский, М. В. (2013). Evaluation of weight of coolant in the reactor plant during full loss of feeding coolant. Technology Audit and Production Reserves, 6(1(14), 26–29. https://doi.org/10.15587/2312-8372.2013.19542

Issue

Vol. 6 No. 1(14) (2013): Technology audit

Section

Technology audit

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Copyright (c) 2016 Юрий Константинович Тодорцев, Евгения Александровна Кокол, Марк Витальевич Никольский

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