Nuclear Engineering and Technology

  • 제54권6호
  • /
  • Pages.2262-2275
  • /
  • 2022
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Conceptual design of a copper-bonded steam generator for SFR and the development of its thermal-hydraulic analyzing code

  • Im, Sunghyuk (Center for Thermometry and Fluid Flow Metrology, Korea Research Institute of Standards and Science) ;
  • Jung, Yohan (Versatile Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Hong, Jonggan (Versatile Reactor Technology Development, Korea Atomic Energy Research Institute) ;
  • Choi, Sun Rock (Versatile Reactor Technology Development, Korea Atomic Energy Research Institute)
  • 투고 : 2021.06.22
  • 심사 : 2021.12.04
  • 발행 : 2022.06.25
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초록

The Korea Atomic Energy Research Institute (KAERI) studied the sodium-water reaction (SWR) minimized steam generator for the safety of the sodium-cooled fast reactor (SFR), and selected the copper bonded steam generator (CBSG) as the optimal concept. This paper introduces the conceptual design of the CBSG and the development of the CBSG sizing analyzer (CBSGSA). The CBSG consists of multiple heat transfer modules with a crossflow heat transfer configuration where sodium flows horizontally and water flows vertically. The heat transfer modules are stacked along a vertical direction to achieve the targeted large heat transfer capacity. The CBSGSA code was developed for the thermal-hydraulic analysis of the CBSG in a multi-pass crossflow heat transfer configuration. Finally, we conducted a preliminary sizing and rating analysis of the CBSG for the trans-uranium (TRU) core system using the CBSGSA code proposed by KAERI.

키워드

과제정보

This work was supported by the National Research Foundation of Korea, Republic of Korea (NRF) grant and National Research Council of Science & Technology (NST) grant funded by the Korean government (MSIT) [grant numbers 2021M2E2A2081061, CAP-20-03-KAERI].

참고문헌

  1. John E. Kelly, Generation IV International Forum: a decade of progress through international cooperation, Prog. Nucl. Energy 77 (2014) 240-246. https://doi.org/10.1016/j.pnucene.2014.02.010
  2. K. Aoto, P. Dufour, Y. Hongyi, J.P. Glatz, Y.I. Kim, Y. Ashurko, R. Hill, N. Uto, A summary of sodium-cooled fast reactor development, Prog. Nucl. Energy 77 (2014) 247-265. https://doi.org/10.1016/j.pnucene.2014.05.008
  3. S.J. Ahn, K.S. Ha, W.P. Chang, S.H. Kang, K.L. Lee, C.W. Choi, S.W. Lee, J. Yoo, J.- H. Jeong, T. Jeong, Evaluation of a sodium-water reaction event caused by steam generator tubes break in the prototype generation IV sodium-cooled fast reactor, Nucl. Eng. Technol. 48 (4) (2016) 952-964. https://doi.org/10.1016/j.net.2016.02.016
  4. J. Yoo, J. Chang, J.Y. Lim, J.S. Cheon, T.H. Lee, S.K. Kim, K.L. Lee, H.K. Joo, Overall system description and safety characteristics of prototype Gen IV sodium cooled fast reactor in Korea, Nucl. Eng. Technol. 48 (5) (2016) 1059-1070. https://doi.org/10.1016/j.net.2016.08.004
  5. J. Hong, J.-W. Han, Technology status of sodium-water reaction minimized heat exchanger. KAERI Report, 2018. KAERI/AR-1198/2018.
  6. J. Hong, Development Report on the Sodium-Water Reaction Minimized Heat Exchanger, KAERI report, 2018. SFR-330-P2-486-001.
  7. S. Im, J. Hong, J.-W. Han, S.R. Choi, et al., Preliminary Development of a Copper Bonded Steam Generator for Minimizing a Sodium-Water Reaction, Transactions of the Korean Nuclear Society Spring Meeting, Jeju, Korea, May 23-24, 2019.
  8. E.R. Adam, C.V. Gregory, Brief history of the operation of the prototype fast reactor at Dounreay, Nucl. Eng. 35 (1994) 112-117.
  9. Fast Reactor Database 2006 Update, 2006. IAEA-TECDOC-1531.
  10. L.M. Finch, FOREIGN TRIP REPORT. No. BNWL-602. Battelle-Northwest, Pacific Northwest Lab., Richland, Wash, 1966.
  11. D.V. Sherwood, Y. Chikazawa, A reliable steam generator that allow the elimination of the secondary sodium circuit in an LMFBR, Nucl. Technol. 150 (2005) 111-119. https://doi.org/10.13182/nt05-a3609
  12. B. Lubarsky, S.J. Kaufman, Review of Experimental Investigations of Liquid-Metal Heat Transfer, 1956. NACA-TR-1270.
  13. ASME Boiler and Pressure Vessel Code, Section-II, Part D, Properties (Metric), ASME BPVC.II.D.M-2015.
  14. Y. Birol, Thermal fatigue testing of CuCrZr alloy for high temperature tooling application, J. Mater. Sci. 45 (2010) 4501-4506. https://doi.org/10.1007/s10853-010-4542-0
  15. E.K. Kim, S.-O. Kim, M.-H. Wi, J.H. Eoh, Development of the Thermal Hydraulic Analysis Code for a Copper Bonded Steam Generator in LMR, 2002. KAERI/TR-2300/2002.
  16. J.C. Chen, Correlation for boiling heat transfer to saturated fluids in convective flow, Ind. Eng. Chem. Process Des. Dev. 5 (3) (1966) 322-329. https://doi.org/10.1021/i260019a023
  17. J.B. Heineman, An Experimental Investigation of Heat Transfer to Superheated Steam in Round and Rectangular Channels, 1960. ANL-6213.
  18. R.K. Shah, D.P. Sekulic, Fundamentals of Heat Exchanger Design, John Wiley & Sons Inc., Hoboken, New Jersey, 2003.
  19. A. Triboix, Exact and approximate formulas for cross flow heat exchangers with unmixed fluids, Int. Commun. Heat Mass Tran. 36 (2009) 121-124. https://doi.org/10.1016/j.icheatmasstransfer.2008.10.012
  20. S.J. Yoon, P. Sabharwall, E.S. Kim, Numerical study on crossflow printed circuit heat exchanger for advanced small modular reactors, Int. J. Heat Mass Tran. 70 (2014) 250-263. https://doi.org/10.1016/j.ijheatmasstransfer.2013.10.079
  21. S.R. Choi, "Preliminary Report on the System Design Concept for the TRU Core", 2018. SFR-050-P2-486-001.