Transactions of the Korean Society of Mechanical Engineers B (대한기계학회논문집B)

The Korean Society of Mechanical Engineers (대한기계학회)
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Thermal-Hydraulic Research Review and Cooperation Outcome for Light Water Reactor Fuel

경수로핵연료 열수력 연구개발 분석 및 연산학 협력 성과

  • In, Wang Kee (Dept. of LWR Fuel Technology Development, Korea Atomic Energy Research Institute) ;
  • Shin, Chang Hwan (Dept. of LWR Fuel Technology Development, Korea Atomic Energy Research Institute) ;
  • Lee, Chi Young (Dept. of Fire Protection Engineering, Pukyong Nat'l Univ.) ;
  • Lee, Chan (Dept. of LWR Fuel Technology Development, Korea Atomic Energy Research Institute) ;
  • Chun, Tae Hyun (Dept. of LWR Fuel Technology Development, Korea Atomic Energy Research Institute) ;
  • Oh, Dong Seok (Dept. of LWR Fuel Technology Development, Korea Atomic Energy Research Institute)
  • 인왕기 (한국원자력연구원 경수로핵연료기술개발부) ;
  • 신창환 (한국원자력연구원 경수로핵연료기술개발부) ;
  • 이치영 (부경대학교 소방공학과) ;
  • 이찬 (한국원자력연구원 경수로핵연료기술개발부) ;
  • 전태현 (한국원자력연구원 경수로핵연료기술개발부) ;
  • 오동석 (한국원자력연구원 경수로핵연료기술개발부)
  • Received : 2016.07.19
  • Accepted : 2016.10.05
  • Published : 2016.12.01
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Abstract

The fuel assembly for pressurized water reactor (PWR) consists of fuel rod bundle, spacer grid and bottom/top end fittings. The cooling water in high pressure and temperature is introduced in lower plenum of reactor core and directed to upper plenum through the subchannel which is formed between the fuel rods. The main thermal-hydraulic performance parameters for the PWR fuel are pressure drop and critical heat flux in normal operating condition, and quenching time in accident condition. The Korea Atomic Energy Research Institute (KAERI) has been developing an advanced PWR fuel, dual-cooled annular fuel and accident tolerant fuel for the enhancement of fuel performance and the localization. For the key thermal-hydraulic technology development of PWR fuel, the KAERI LWR fuel team has conducted the experiments for pressure drop, turbulent flow mixing and heat transfer, critical heat flux(CHF) and quenching. The computational fluid dynamics (CFD) analysis was also performed to predict flow and heat transfer in fuel assembly including the spent fuel assembly in dry cask for interim repository. In addition, the research cooperation with university and nuclear fuel company was also carried out to develop a basic thermal-hydraulic technology and the commercialization.

가압경수로에 장전되는 핵연료집합체는 연료 봉 다발과 지지격자 및 상하단 고정체로 구성되어 있다. 고온 고압의 냉각수는 원자로 하부로 유입되어 연료 봉 사이로 형성된 부수로를 따라 노심 상부로 흐른다. 경수로핵연료의 주요 열수력 성능인자는 정상운전시 압력강하 및 임계열속이며 사고시에는 급랭 시간이다. 한국원자력연구원에서는 경수로핵연료의 성능을 향상시키고 국산화를 위해 고성능 경수로핵연료, 이중냉각 핵연료 및 사고저항성 핵연료를 개발하였다. 경수로핵연료의 열수력 핵심기술을 개발하기 위해 압력강하 실험, 난류 유동혼합/열전달 실험, 임계열속 및 급랭 시험을 수행하였으며 전산유체역학 방법도 활용하였다. 더불어 사용후핵연료의 임시저장을 위한 건식저장 용기의 열유동에 대한 전산유체해석을 수행하였다. 한편, 경수로핵연료의 열수력 기반기술을 개발하고 실용화를 위해 대학 및 산업체와 협력연구도 진행하였다.

Keywords

References

  1. Shin, C. H., Chun, T. H., Oh, D. S. and In, W. K., 2012, "Evaluation of Loss Coefficient for an End Plug with Side Holes in Dual-cooled Annular Nuclear Fuel," J. of Mechanical Science and Technology, Vol. 26, No. 10, pp. 1-6. https://doi.org/10.1007/s12206-011-1008-5
  2. Lee, C. Y., Shin, C. H. and In, W. K., 2012, "Pressure Drop in Dual-cooled Annular and Cylindrical Solid Fuel Assemblies for Pressurized Water Reactor," Nuclear Engineering and Design, Vol. 50, pp. 287-293.
  3. In, W. K., Hwang, Shin, C. H., Kwack, Y. K. and Lee, C. Y., 2015, " Measurement and CFD Calculation of Spacer Loss Coefficient for a Tight-Lattice Fuel Bundle," Nuclear Engineering and Design, Vol. 284, pp. 153-161. https://doi.org/10.1016/j.nucengdes.2014.12.021
  4. Lee, C. Y., Shin, C. H. and In, W. K., 2013, "Effect of Gap Width on Turbulent Mixing of Parallel Flow in a Square Channel with a Cylindrical Rod," Experimental Thermal and Fluid Science, Vol. 47, pp. 98-107. https://doi.org/10.1016/j.expthermflusci.2013.01.005
  5. In, W. K., Shin, C. H. and Lee, C. Y., 2014, "Experimental Observation of Forced Flow Mixing in a Tight-Lattice Rod Bundle," American Nuclear Soc. Annual Meeting, June 16, Reno NV, USA.
  6. Lee, C. Y., Shin, C. H. and In, W. K., 2015, "Convective Heat Transfer Experiment of Rod Bundle Flow with Twist-vane Spacer Grid," Nuclear Engineering and Design, Vol. 295, pp. 173-181. https://doi.org/10.1016/j.nucengdes.2015.10.004
  7. Lee, C. and In, W. K., 2014, "Prediction of Water Droplet Evaporation in Zircaloy Surface," J. of Nuclear Science and Technology, Vol. 51, No. 4, pp. 348-356.
  8. Lee, C., Chun, T. H. and In, W. K., 2014, "Effect of Change in Surface Condition induced by Oxidation on Transient Pool Boiling Heat Transfer of Vertical Stainless Steel and Copper Rodlets," Int. J. Heat and Mass Transfer, Vol. 79, pp. 397-407. https://doi.org/10.1016/j.ijheatmasstransfer.2014.08.030
  9. In, W. K., Hwang, D. H. and Jeong, J. J., 2013, " A Subchannel and CFD Analysis of Void Distribution for the BWR Fuel Bundle Test Benchmark," Nuclear Engineering and Design, Vol. 258, pp. 211-225. https://doi.org/10.1016/j.nucengdes.2013.02.006
  10. In, W. K., Kwack, Y. K., Kook, D. H. and Koo, Y. H., 2014, "CFD Analysis of Fluid Flow and Heat Transfer in Spent Fuel for a Dry Storage," NTHAS9: The Korean-Japan Symposium on Nuclear Thermal Hydraulics and Safety, November 16-19, Buyeo, Korea.