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Experimental investigation of two-phase flow and wall heat transfer during reflood of single rod heater

단일 가열봉의 재관수 시 2상유동 및 벽면 열전달에 관한 실험적 연구

  • Park, Youngjae (Department of Nuclear Engineering, Kyung Hee University) ;
  • Kim, Hyungdae (Department of Nuclear Engineering, Kyung Hee University)
  • Received : 2020.10.26
  • Accepted : 2020.11.17
  • Published : 2020.12.31

Abstract

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

Keywords

References

  1. Jacopo Buongiorno, "Can corrosion and CRUD actually improve safety margins in LWRs?," Annals of Nuclear Energy 63 (2014) 9-21 https://doi.org/10.1016/j.anucene.2013.07.019
  2. L.E. Hochreiter, "FLECHT-SEASET program final report," NUREG/CR4167, 1985
  3. Stephen M. Bajorek and Fan-Bill Cheung, "Rod Bundle Heat Transfer Thermal-Hydraulic Program," Nuclear Technology 205:1-2 307-327, 2019 https://doi.org/10.1080/00295450.2018.1510697
  4. 문상기, 박현식, 조석, 김복득, 박종국, 천세영, "대형냉각재상실사고 시 재관수 현상에 대한 실험 및 이론 연구 현황 분석," 한국원자력연구원, KAERI/AR-666, 2003
  5. 신고리 3호기 최종안전성분석보고서, 2015
  6. Juan J. Carbajo, "A study on the rewetting temperature," Nuclear Engineering and Design 84 21-52, 1985 https://doi.org/10.1016/0029-5493(85)90310-3
  7. R.B. Duffey and D.T.C. Porthouse, "The physics of rewetting in water reactor emergency core cooling," Nuclear Engineering and Design 25 379-394, 1973 https://doi.org/10.1016/0029-5493(73)90033-2
  8. Kevin D. Kimball and Ramendra P. Roy, "Quench front propagation during bottom reflooding of heated annular channel," Nuclear Engineering and Design 76 79-88 1983 https://doi.org/10.1016/0029-5493(83)90049-3