Nuclear Engineering and Technology

Korean Nuclear Society (한국원자력학회)
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Method of estimating break size in piping loop systems

  • Sheng-Dih Hwang (Department of Nuclear Systems Engineering, National Atomic Research Institute)
  • Received : 2024.02.28
  • Accepted : 2024.06.30
  • Published : 2024.11.25
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Abstract

The approach for determining the break size of recirculation loops in a multiple-loop power plant in the event of a loss of coolant accident (LOCA) is presented in this study. In this study, the MAAP5 simulation program was used. An approach to measuring the size of a crack or break in the cooling system is the temperature difference between the recirculation loops. This method does not require any additional facilities; it compares the temperatures of the cooling loops to determine which one has a rupture. The best data source was the loop monitoring system, which sends temperature data for analysis to the main control room. A real operating power reactor training simulator and the FSAR are applied to evaluate MAAP5, the methodology's engine. The results of the MAAP5 simulation code were consistent with those of the power plant simulator. Therefore, MAAP5 could produce enough analytical data to create the relationship diagram between temperature difference and break size. The study hypothesized that there exists a maximum value of temperature difference corresponding to each break size and suggested that applying the absolute maximum temperature difference can aid in identifying the break size. This approach proposes an assistive method for determining the size of a fracture or break in the recirculation system by leveraging the temperature difference between each loop. This approach eliminates the need for additional facilities, as temperature data from the recirculation loops can be transmitted to the main control room. After the reactor scram, operators can monitor the maximum temperature differences at the inlet to estimate the break size. Although the fitting curve used to preliminary estimate the Large Break Loss of Coolant Accident break size may overestimate the break size, it still provides valuable insights. This novel tool offers a rapid and comprehensive method for detecting LOCA events in the recirculation loops.

Keywords

Acknowledgement

I would like to thank my colleagues for helping me complete this study: Mrs. Lin, Chin-Tsu, for her technical assistance with data programming, and Dr. Kuo, Weng-sheng, for his writing technique advice.

References

  1. Hyong Chol Kim, Young Jin Lee, Sam Hee Han, Evaluation of fuel burnup effects on LOCA PCT using MARS/FRAPTRAN coupled code, Trans. Kor. Nucl. Soc. Spring Meeting JejuSpring Meeting Jeju (2017). Korea, May 18-19. 
  2. Zhongchun Li, et al., Preliminary study of uncertainty qualification methods based on the simplified LB LOCA model for PCT estimation, Prog. Nucl. Energy 128 (2020). 
  3. I. Catton, et al., Quantifying reactor safety margins Part 6: a physically based method of estimating pwr large break loss of coolant accident pct, Nucl. Eng. Des. 119 (1990) 109-117.  https://doi.org/10.1016/0029-5493(90)90076-A
  4. Jongseuk Park, et al., PCT margin quantification in nuclear power plants. Proceedings of the KNS Spring Meeting, 2007. 
  5. K. Nikitin, P. Mueller, J. Martin, W. van Doesburg, D. Hiltbrand, BWR loss of coolant accident simulation by means of RELAP5, Nucl. Eng. Des. 309 (2016) 113-121.  https://doi.org/10.1016/j.nucengdes.2016.09.008
  6. Kanji Tasaka, Yasuo Koizumi, Mitsuhiro Suzuki, Yoshinari Anoda, Yutaka Kukita, Hiroshige Kumamaru, Hideo Nakamura Taisuke Yonomoto, Masahiro KAWAJI and Hideo MURATA ROSA-III Experimental Program for BWR LOCA/ECCS Integral Simulation Tests, 1987. JAERI-1307. 
  7. Yu-seng Wang, Analysis of Ultimate Response Guideline of Lungmen Nuclear Power Plant, dissertation of NTHU, 2017. 
  8. Xiuqun Hou, Danmei Xie, Peng Zhang, Cong Wang, The simulation and safety analysis of CNP 600 primary loop under different broken areas of cold leg. Nuclear Science and Technology, 2014. 
  9. F. D'Auria, M. Leonardi, R. Pochard, Methodology for the evaluation of thermal hydraulic codes accuracy, in: Proceedings of International Conference on New Trends in Nuclear System Thermal Hydraulics, May-June 1994. Pisa, Italy. 
  10. Yasuharu Kawabe, Tamio Kohriyama, Masanori Ohtani, Comparison of MAAP4. 03 with RELAP5/M0D2, JAERI-Conf 99-005, 1998. 
  11. Cesare Frepoli, An overview of westinghouse realistic large break LOCA evaluation model, Sci. Technol. Nucl. Install. (2008). 
  12. Seihiro Itoya, Hideo Nagasaka, Kanji Tasaka, Assessment of SAFER03 code using ROSA-III break area spectrum tests on boiling water reactor loss-of-coolant accidents, J. Nucl. Sci. Technol. 24 (1987) 639-652.  https://doi.org/10.1080/18811248.1987.9735860
  13. Yasuo Koizumi, Kanji TA. Saka, Investigation of break location effects on thermal hydraulics during intermediate break loss-of coolant accident experiments at ROSA-III, J. Nucl. Sci. Technol. (Tokyo, Jpn.) 23 (11) (1986) 1008~1019.  https://doi.org/10.1080/18811248.1986.9735089
  14. Hideo Nakamura, Yutaka Kukita, Kanji Tasaka, BWR loss-of-coolant accident tests at ROSA-III with high-temperature emergency core coolant injection, J. Nucl. Sci. Technol. (Tokyo, Jpn.) 25 (1988) 169-179.  https://doi.org/10.1080/18811248.1988.9733573
  15. L.I.U. Peiqi, Z.H.A.O. Pengcheng, Y.U. Tao, X.I.E. Jinsen, Zhenping Chen, X.I. E. Chao, L.I.U. Zijing, Z.E.N.G. Wenjie, Analysis of LOCA with different break sizes in PWR, Nucl. Tech. 42 (2019). 
  16. KUOSHENG NUCLEAR POWER STATION SAFETY ANALYSIS REPORT FOR STRETCH POWER UPRATE（SPU)REV. 0, TAIWAN POWER COMPANY, 2012 (in Chinese). 
  17. Ho Wei-ming, Taipower Nuclear Monthly Report, P2, 2005/09/03, pp. 8-50 (in Chinese). 
  18. https://getdata-graph-digitizer.software.informer.com/.