Journal of the Korea Academia-Industrial cooperation Society (한국산학기술학회논문지)

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Effects of Crud on reflood heat transfer in Nuclear Power Plant

핵연료 크러드가 원전 재관수 열전달에 미치는 영향

  • Yoo, Jin (School of Mechanical Engineering, Chungnam National University) ;
  • Kim, Byoung Jae (School of Mechanical Engineering, Chungnam National University)
  • 유진 (충남대학교 기계공학부) ;
  • 김병재 (충남대학교 기계공학부)
  • Received : 2021.03.23
  • Accepted : 2021.05.07
  • Published : 2021.05.31
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Abstract

CRUD (chalk river unidentified deposits) is a porous material deposited on the surface of nuclear fuel during nuclear power plant operation. The CRUD is composed of metal oxides, such as iron, nickel, and chromium. It is essential to investigate the effects of the CRUD layer on the wall heat transfer between the nuclear fuel surface and the coolant in the event of a nuclear accident. CRUD only negatively affects the temperature of the nuclear fuel due to heat resistance because the effects of the CRUD layer on two-phase boiling heat transfer are not considered. In this study, the physical property models for the porous CRUD layer were developed and implemented into the SPACE code. The effects of boiling heat transfer models on the peak cladding temperature and quenching were investigated by simulating a reflood experiment. The calculation results showed some positive effects of the CRUD layer.

크러드는 원자력 발전소 운전 시 핵연료 표면에 침적되는 철-니켈-크롬 등의 금속 산화물로 이루어진 다공성 물질이다. 그 두께는 수십 ㎛ 수준이다. 발전소의 냉각재상실사고 시 크러드 층은 핵연료-냉각수 열전달에 영향을 미치게 되어 원전 안전성 측면에서 그 영향을 살펴보는 것이 중요하다. 일반적으로 크러드는 열저항으로 인하여 핵연료 온도를 높이는 부정적 효과가 있는 것으로 알려져 있었다. 그 이유는 크러드에 의하여 핵비등, 최소막비등온도, 단상증기 열전달, 임계열유속, 막비등 열전달 등 2상유동 열전달 특성을 고려하지 않았기 때문이다. 본 연구에서는 다공성 크러드 물질의 물성치를 모델링하고 이를 국내 원전안전해석 코드인 SPACE에 탑재하였다. 크러드는 다공성 고체 물질이고 표면이 거칠기 때문에 최소막비등온도와 단상증기 열전달이 증가할 것으로 예상된다. 이에 최소막비등온도와 단상증기 열전달이 최대 피복재 온도 및 급냉에 미치는 영향을 평가하였다. 시험 계산은 기존 FLECHT-SEASET 재관수 실험 장치에 기반으로 수행되었다. 계산결과 최소막비등온도가 상승하여 급냉시간이 줄어들었다. 단상증기 열전달의 경우 약 20% 증가할 때까지는 최대 피복재 온도가 하강하였다. 크러드 층이 원전 안정성 측면에서 긍정적인 효과가 있음을 확인하였다.

Keywords

Acknowledgement

본 연구는 한국수력원자력(주) 연구과제(No. 2018-TECH-8)로 수행되었음.

References

  1. Cinosi et al, "The effective thermal conductivity of crud and heat transfer from crud-coated PWR fuel", Nuclear Engineering and Design, Vol. 241, pp.792-798, 2011. DOI: https://doi.org/10.1016/j.nucengdes.2010.12.015
  2. Huh et al., "Preliminary study on effect of the crud deposits during LBLOCA condition", Transactions of the Korean Nuclear Society Autumn Meeting, Korea Nuclear Society, PyeongChang, Korea, October 30-31, 2008.
  3. Lee and Kim, "Crud and oxide layer modeling for safety analysis of a PWR", Transactions of the Korean Nuclear Society Spring Meeting, Korea Nuclear Society, Jeju, Korea, May 12-13, 2016.
  4. Cinosi, N. and Walker, S.P., "CFD analysis of localized crud effects on the flow of coolant in nuclear rod bundle", Nuclear Engineering and Design, Vol.305, pp.28-38, 2016. https://doi.org/10.1016/j.nucengdes.2015.12.003
  5. Petrov et al., "Prediction of CRUD deposition on PWR fuel using a state-of-the-art CFD-based multi-physics computational tool", Nuclear Engineering and Design, Vol.229, pp.95-104, 2016. https://doi.org/10.1016/j.nucengdes.2015.10.010
  6. Isnaini et al., "Prediction of fuel temperature of AP1000 due to the formation of crud and oxide layer," J. Tek. Reaktor. Nukl, Vol.19 No.2, pp.1152-1166, 2018. DOI: http://dx.doi.org/10.17146/tdm.2017.19.2.3521
  7. Walter et al., "Proof-of-principle of high-fidelity coupled crud deposition and cycle deposition simulation", Annals of Nuclear Energy, Vol.85, pp.1152-1166, 2015. DOI: https://doi.org/10.1016/j.anucene.2015.07.034
  8. Hu et al., "Considering the thermal resistance of crud in LOCA analysis", Transactions of the American Nuclear Society, American Nuclear Society, America, Vol.101, pp.590-592, 2009.
  9. Sang-Jun Ha, Chan-Eok Park, Kyung-Doo Kim, and Chang-Hwan Ban, "Development of the space code for nuclear power plants," Nuclear Engineering and Technology, Vol.43, pp.45-62, 2011. DOI: https://doi.org/10.5516/NET.2011.43.1.045
  10. Korea Hydro & Nuclear Power Co. Ltd., SPACE 3.0 manual Volume 1 theory manual, S06NX08-K-1-TR-36, Rev.0
  11. Henshaw, J., et al., 2006. "A model of chemistry and thermal hydraulics in PWR fuel crud deposits", Journal of Nuclear Materials, Vol.353, pp.1-11. DOI: https://doi.org/10.1016/j.jnucmat.2005.01.028
  12. Pan et al, "Wick Boiling Performance in Porous Deposits with Chimneys", AICHE/ANS National Heat Transfer Conference Symposium on Multiphase and Heat Transfer, American Society of Mechanical Engineers, DENVER, COLORADO, United States, August 1985.
  13. Lee et al., "Thermal resistance effects of crud and oxide layers to the safety analysis", Top Fuel, Prague Czech Republic, September 30 ~ October 4, 2018.
  14. W. D. Kingery, J. Francl, R. L. Coble, T. Vasilos, J. "Thermal Conductivity: X, Data for Several Pure Oxide Materials Corrected to Zero Porosity", Journal of the American Ceramic Society, Vol.37, pp.107-111, 1954. https://doi.org/10.1111/j.1551-2916.1954.tb20109.x
  15. "Thermal Expansion, Heat Capacity and Thermal Conductivity of Nickel Ferrite (NiFe2O4)", Journal of the American Ceramic Society, No.5 pp.1559-1565, 2014.
  16. Gareth S. Parkinson, "Iron oxide surfaces", Surface Science Reports 71, 272-365, 2016.
  17. W.G. Luscher et. al., "Material Property Correlations: Comparison between FRAPCON-3.4, FRAPTRAN-1.4, and MATPRO", NUREG/ CR-7024, PNNL-19417, 2011.
  18. B. S. HEMINGWAY, "Thermodynamic properties for bunsenite, NiO, magnetite, Fe3O4, and hematite, Fe2O3, with comments on selected oxygen buffer reactions", American Mineralogist, Vol.75, pp.781-790, 1990.
  19. A.T. Nelson et. al., "Thermal Expansion, Heat Capacity and Thermal Conductivity of Nickel Ferrite (NiFe2O4)", Journal of the American Ceramic Society, MIT open access article, 2013.
  20. SCDAP/RELAP5/MOD3.3 Code Manual: MATPRO - A Library of Materials Properties for Light-Water-Reactor Accident Analysis.
  21. KL.E. Hochreiter, 1986, FLECHT SEASET Program: Final Report, NUREG/CR-4167, EPRI NP-4112, WCAP-10926.
  22. Carbajo, J., "A Study on the Rewetting Temperature", Nuclear Engineering and Design, Vol.84, pp.21-52, 1985. https://doi.org/10.1016/0029-5493(85)90310-3
  23. Inayatov, A. Y., "Correlation of Data on Heat Transfer Flow Parallel to Tube Bundles at Relative Tube Pitches of 1.1 < s/d < 1.6." Heat Transfer-Soviet Research. 7. 3. May-June 1975.
  24. Bhattacharyya, A. ,"Heat transfer and pressure drop with rough surfaces: A literature survey," AE-141, 1964.
  25. Ki-Yong Choi et al., "Development of a wall-to-fluid heat transfer package for the space code", Nuclear Engineering and Technology, Vol.41, pp.1143-1156, 2009. DOI: https://doi.org/doi.org/10.5516/NET.2009.41.9.1143