Nuclear Engineering and Technology

  • 제54권3호
  • /
  • Pages.1085-1097
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  • 2022
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  • 1738-5733(pISSN)
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  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Assessment of thermal fatigue induced by dryout front oscillation in printed circuit steam generator

  • Kwon, Jin Su (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology) ;
  • Kim, Doh Hyeon (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology) ;
  • Shin, Sung Gil (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology) ;
  • Lee, Jeong Ik (Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology) ;
  • Kim, Sang Ji (Korea Atomic Energy Research Institute)
  • 투고 : 2021.03.08
  • 심사 : 2021.09.02
  • 발행 : 2022.03.25
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초록

A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.

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과제정보

This work was supported by the National Research Foundation of Korea (NRF), South Korea grant funded by the Korean Government (MIST). (No. 2018M2A8A4081307) and supported by the KAI-NEET Institute, South Korea, KAIST, Korea. (N11210064). We also would like to acknowledge the technical support from ANSYS Korea, South Korea.

참고문헌

  1. Seong Jun Bae, et al., Condensation heat transfer and multi-phase pressure drop of CO2 near the critical point in a printed circuit heat exchanger, Int. J. Heat Mass Tran. 129 (2019) 1206-1221. https://doi.org/10.1016/j.ijheatmasstransfer.2018.10.055
  2. Di Ronco, Antonio Cammi Andrea, Stefano Lorenzi, Preliminary analysis and design of the heat exchangers for the molten salt fast reactor, Nuclear Engineering and Technology 52 (1) (2020) 51-58. https://doi.org/10.1016/j.net.2019.07.013
  3. Hong Xu, et al., The optimization for the straight-channel PCHE size for supercritical CO2 Brayton cycle, Nuclear Engineering and Technology (2020).
  4. Jin Su Kwon, et al., Compact heat exchangers for supercritical CO2 power cycle application, Energy Convers. Manag. 209 (2020) 112666. https://doi.org/10.1016/j.enconman.2020.112666
  5. Tony Johnston, William Levy, Rumbold Svend, Application of printed circuit heat exchanger technology within heterogeneous catalytic reactors, Amer Inst Chem Eng Annual Meeting 2001 (November 2001).
  6. Shin, Wook Chang, Hee Cheon No, Experimental study for pressure drop and flow instability of two-phase flow in the PCHE-type steam generator for SMRs, Nucl. Eng. Des. 318 (2017) 109-118. https://doi.org/10.1016/j.nucengdes.2017.04.004
  7. Xiaofei Yuan, Lixin Yang, Zemin Shang, Experimental and numerical investigation on flow boiling in a small semi-circular channel of plate once-through steam generator, Heat Tran. Eng. (2021) 1-22.
  8. Kim, Sang Ji, Taewoo Kim, Design methodology and computational fluid analysis for the printed circuit steam generator (PCSG), J. Mech. Sci. Technol. (2020) 1-12.
  9. Muhammad Ilyas, Fatih Aydogan, Steam generator performance improvements for integral small modular reactors, Nuclear Engineering and Technology 49 (8) (2017) 1669-1679. https://doi.org/10.1016/j.net.2017.08.011
  10. Koroush Shirvan, Pavel Hejzlar, Mujid S. Kazimi, The design of a compact integral medium size PWR, Nucl. Eng. Des. 243 (2012) 393-403. https://doi.org/10.1016/j.nucengdes.2011.11.023
  11. Han-Ok Kang, Hun Sik Han, Young-In Kim, Thermal-hydraulic design of a printed-circuit steam generator for integral reactor, The KSFM Journal of Fluid Machinery 17 (6) (2014) 77-83.
  12. Alexey Lokhov, Ron Cameron, Vladislav Sozoniuk, OECD/NEA study on the economics and market of small reactors, Nuclear Engineering and Technology 45 (6) (2013) 701-706. https://doi.org/10.5516/NET.02.2013.517
  13. H.T. Liu, S. Kakac, F. Mayinger, Characteristics of transition boiling and thermal oscillation in an upflow convective boiling system, Exp. Therm. Fluid Sci. 8 (3) (1994) 195-205. https://doi.org/10.1016/0894-1777(94)90048-5
  14. Gyeong-Hoi Koo, Jae-Han Lee, Creep-fatigue design studies for a sodium-cooled fast reactor with tube sheet-to-shell structure subjected to elevated temperature service, J. Mech. Sci. Technol. 24 (3) (2010) 711-719. https://doi.org/10.1007/s12206-010-0121-1
  15. Kaushik Chatterjee, Mohammad Modarres, A probabilistic physics-of-failure approach to prediction of steam generator tube rupture frequency, Nucl. Sci. Eng. 170 (2) (2012) 136-150. https://doi.org/10.13182/NSE11-27
  16. T. Chiang, D.M. France, T.R. Bump, Calculation of tube degradation induced by dryout instability in sodium-heated steam generators, Nucl. Eng. Des. 41 (2) (1977) 181-191. https://doi.org/10.1016/0029-5493(77)90108-X
  17. Han-Ok Kang, et al., Structural integrity confirmation of a once-through steam generator from the viewpoint of flow instability, J. Nucl. Sci. Technol. 44 (1) (2007) 64-72. https://doi.org/10.3327/jnst.44.64
  18. Yu Yang, et al., Optimizing the size of a printed circuit heat exchanger by multi-objective genetic algorithm, Appl. Therm. Eng. 167 (2020) 114811. https://doi.org/10.1016/j.applthermaleng.2019.114811
  19. Kim, In Hun, et al., Design study and cost assessment of straight, zigzag, S-shape, and OSF PCHEs for a FLiNaK-SCO2 Secondary Heat Exchanger in FHRs, Ann. Nucl. Energy 94 (2016) 129-137. https://doi.org/10.1016/j.anucene.2016.02.031
  20. E.W. Lemmon, et al., NIST Standard Reference Database 23: Reference Fluid Thermodynamic and Transport Properties-REFPROP, Version 10.0, National Institute of Standards and Technology, Standard Reference Data Program, Gaithersburg, 2018.
  21. U.S. NRC, TRACE V5. 0 theory manual, field equations, solution methods, and physical models, United States Nucl. Regul. Comm (2010).
  22. D. Mo France, et al., Characteristics of transition boiling in sodium-heated steam generator tubes, J. Heat Tran. 101 (1979) 270-275. https://doi.org/10.1115/1.3450959
  23. S.S. Samra, V.K. Dhir, Study of thermal oscillations at the dryout front in half heated tubes, J. Sol. Energy Eng. 107 (1985) 343-351. https://doi.org/10.1115/1.3267703
  24. A.M. Osman, J.V. Beck, Investigation of transient heat transfer coefficients in quenching experiments, J. Heat Tran. 112 (1990) 843-848. https://doi.org/10.1115/1.2910490
  25. Hong Hu, et al., Boiling and quenching heat transfer advancement by nano-scale surface modification, Sci. Rep. 7 (1) (2017) 1-16. https://doi.org/10.1038/s41598-016-0028-x
  26. Youho Lee, Jeong Ik Lee, Structural assessment of intermediate printed circuit heat exchanger for sodium-cooled fast reactor with supercritical CO2 cycle, Ann. Nucl. Energy 73 (2014) 84-95. https://doi.org/10.1016/j.anucene.2014.06.022
  27. Raciel de la Torre, , Juan-Luis Francois, Cheng-Xian Lin, Assessment of the design effects on the structural performance of the Printed Circuit Heat Exchanger under very high temperature condition, Nucl. Eng. Des. 365 (2020) 110713. https://doi.org/10.1016/j.nucengdes.2020.110713
  28. Kim, Eok Dong, et al., Numerical investigation on thermalehydraulic performance of new printed circuit heat exchanger model, Nucl. Eng. Des. 238 (12) (2008) 3269-3276. https://doi.org/10.1016/j.nucengdes.2008.08.002
  29. ASME Boiler and Pressure Vessel Code. Section III. Division 1. Subsection NB, ASME, 2015.
  30. Yaqiong Hou, Guihua Tang, Thermal-hydraulic-structural analysis and design optimization for micron-sized printed circuit heat exchanger, J. Therm. Sci. 28 (2) (2019) 252-261. https://doi.org/10.1007/s11630-018-1062-8
  31. Jian Wang, et al., Stress intensity simulation of printed circuit heat exchanger for S-CO 2 brayton cycle, in: 2019 5th International Conference on Transportation Information and Safety (ICTIS), IEEE, 2019, pp. 406-410.
  32. Hirokazu Tsuji, Kenzo Miya, Generating material strength standards of aluminum alloys for research reactors II. Design fatigue curve under non-effective creep condition, Nucl. Eng. Des. 155 (3) (1995) 547-557. https://doi.org/10.1016/0029-5493(94)00928-R
  33. Y. Zhang, T. Lu, Study of the quantitative assessment method for high-cycle thermal fatigue of a T-pipe under turbulent fluid mixing based on the coupled CFD-FEM method and the rainflow counting method, Nucl. Eng. Des. 309 (2016) 175-196. https://doi.org/10.1016/j.nucengdes.2016.09.021