Nuclear Engineering and Technology

  • 제56권1호
  • /
  • Pages.257-264
  • /
  • 2024
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
권호 표지
DOI QR코드

DOI QR Code

Impingement wastage experiment with SUS 316 in a printed circuit steam generator

  • Siwon Seo (School of Mechanical and Control Engineering, Handong Global University Pohang) ;
  • Bowon Hwang (School of Mechanical and Control Engineering, Handong Global University Pohang) ;
  • Sangji Kim (Korea Atomic Energy Research Institute) ;
  • Jaeyoung Lee (School of Mechanical and Control Engineering, Handong Global University Pohang)
  • 투고 : 2022.11.09
  • 심사 : 2023.09.30
  • 발행 : 2024.01.25
PDF 다운로드
⟨ 이전 논문 다음 논문 ⟩

초록

The sodium cooled fast reactor (SFR) is one of the Gen-IV reactors with the most operating experience accumulated. Although the technology level is the most mature among the Gen-IV reactors, there is still a safety problem that has not been solved, which is the sodium-water reaction. Since sodium and water are separated only by a heat transfer tube with a thickness of only a few mm, there is inherently a risk of a sodium-water reaction (SWR) accident in the SFR. In this study, it is attempted to quantitatively evaluate the resistance of SWR accidents by replacing the shell and tube steam generator with printed circuit steam generator (PCSG) as a method to mitigate the SWR accident. To do this, a CATS-S (Compact Accident Tolerance Steam Generator-SWR) facility was designed and built. And for the quantitative evaluation of accident resistance, a methodology for measuring the impingement wastage rate was established. As a result of this research, the impingement wastage rate caused by SWR generated in a PCSG was measured first time. It was confirmed that the impingement wastage phenomenon was suppressed in the PCSG, and the accident resistance was higher than that of the SWR through comparison with the experimental results performed in the existing shell and tube steam generator. In conclusion, a PCSG is more resistant to impingement wastage as a result of the SWR accident than existing shell and tube steam generators, and it is estimated that a PCSG can mitigate SWR accidents, an inherent problem of SFR.

키워드

과제정보

This work was supported by the National Research Foundation of Korea (NRF) grant funded by the Korea government (MSIP). (No. 2017M2A8A4018624 and No. 2017M2A8A4018812).

참고문헌

  1. C. Grandy, US Department of Energy and Nuclear Regulatory Commission - Advanced Fuel Cycle Research and Development Seminar Series, Argonne National Laboratory, ANL-AFCI-238, 2008. August.
  2. I.L. Pioro, et al., Handbook of Generation IV Nuclear Reactors, ELSEVIER, 2016, p. 97. Ch.5.
  3. M. Hori, Sodium water reactions in steam generators of liquid metal fast breeder reactors, At. Energy Rev. 18 (1980) 3.
  4. Mari Marianne Uematsu, et al., Comparison of JSFR design with EDF requirements for future SFR, J. Nucl. Sci. Technol. 52 (3) (2015) 434-447. https://doi.org/10.1080/00223131.2014.953015
  5. D. Plancq, et al., Progress in the ASTRID Sodium Gas Heat Exchanger Development", 2017. IAEA-CN245-286.
  6. H.V. Chamberlain, Project Summary - Sodium-Water Reactions Related to LMFBR Steam Generators", APDA-257, Atomic Power Development Associates, 1970.
  7. N. Kanegae, et al., Wastage and Self-Wastage Phenomena Resulting from Small Leak Sodium-Water Reaction", PNC TN941 76-27, Power Reactor & Nuclear Fuel Development Corporation, 1976.
  8. M. Nisimura, et al., Sodium-Water Reaction Test to Confirm Thermal Influence on Heat Transfer Tubes", PNC TN9400 2003-014, Power Reactor & Nuclear Fuel Development Corporation, 2003.
  9. K. Shimoyama, Wastage-Resistant Characteristics of 12Cr Steel Tube Material", PNC TN9410 2004-009, Power Reactor & Nuclear Fuel Development Corporation, 2004.
  10. S. Kishore, et al., An experimental study on impingement wastage of mod 9Cr 1Mo steel due to sodium water reaction, Nucl. Eng. Des. 243 (2012) 49-55. https://doi.org/10.1016/j.nucengdes.2011.11.008
  11. S. Kishore, et al., Impingement wastage experiments with 9Cr 1Mo steel, Nucl. Eng. Des. 297 (2016) 104-110. https://doi.org/10.1016/j.nucengdes.2015.11.024
  12. R. Duan, Z. Wang, X. Yang, R. Luo, Numerical simulation of sodium-water reaction products transport in steam generator of liquid metal fast breeder reactor on small water/steam leak, J. Nucl. Sci. Technol. (Tokyo, Jpn.) 38 (7) (2001) 527-532. https://doi.org/10.1080/18811248.2001.9715063
  13. S. Kim, J. Eoh, S. Kim, Development of a numerical analysis methodology for the multi-dimensional and multi-phase phenomena of a sodium-water reaction in an SFR steam generator, Ann. Nucl. Energy 34 (2007) 839-848. https://doi.org/10.1016/j.anucene.2007.04.001
  14. S.H. Jang, T. Takata, A. Yamaguchi, A. Uchibori, A. Kurihara, H. Ohshima, Numerical approach for quantification of selfwastage phenomena in sodium-cooled fast reactor, Nucl. Eng. Technol. 47 6 (2015) 700-771. https://doi.org/10.1016/j.net.2015.06.008
  15. S.H. Jang, T. Takata, A. Yamaguchi, "numerical analysis of self-wastage phenomena caused by sodium-water reaction in sodium-cooled fast reactor through simulant experiment ", J. Energy Power Eng. 9 (2015) 539-547.
  16. T. Kim, S. Kim, An analysis of the pressure wave propagation by the sodium-water reaction in the printed circuit steam generator using the SWAAM-II code, Ann. Nucl. Energy 179 (15) (2022), 109423.
  17. C. Gerardi, et al., Fundamental Na-CO2 Interactions in Compact Heat Exchangers Experiments (SNAKE): FY12 Status Report, ANL-ARC- 199, 2012.
  18. J.H. Eoh, et al., Wastage and self-plugging by a potential CO2 ingress in a supercritical CO2 power conversion system of an SFR, J. Nucl. Sci. Technol. 47 (11) (2010) 1023-1036. https://doi.org/10.1080/18811248.2010.9711667
  19. S. Park, J. Min, T. Lee, M. Wi, Investigation of plugging and wastage of narrow sodium channels by sodium and carbon dioxide interaction, Korean Chem. Eng. Res. 54 (6) (2016) 863-870. https://doi.org/10.9713/kcer.2016.54.6.863
  20. Q. Li, et al., Compact heat exchanger: a review and future applications for a new generation of high temperature solar receivers, Renew. Sustain. Energy Rev. 15 (2011) 4855-4875. https://doi.org/10.1016/j.rser.2011.07.066
  21. J. Guidez, G. Prele, Superphenix: Technical and Scientific Achievements, Atlantis Press, 2017, p. 160. Ch.13.
  22. J.E. Hesselgreaves, Compact Heat Exchangers: Selection, Design and Operation (2nd), Pergamon Press, 2001.
  23. Baldev Raj, P. Chellapandi, P.R. Vasudeva Rao, Sodium Fast Reactors with Closed Fuel Cycle", CRC Press, 2017, p. 474.
  24. M. Hori, Sodium-Water Reaction Studies for MONJU Steam Generator, IAEA, International Working Group on Fast Reactor, 1975.