Nuclear Engineering and Technology

  • 제48권5호
  • /
  • Pages.1264-1272
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  • 2016
  • /
  • 1738-5733(pISSN)
  • /
  • 2234-358X(eISSN)
한국원자력학회 (Korean Nuclear Society)
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Estimation of Leak Rate Through Cracks in Bimaterial Pipes in Nuclear Power Plants

  • Park, Jai Hak (Department of Safety Engineering, Chungbuk National University) ;
  • Lee, Jin Ho (Department of Mechanical and Material Assessment, Korea Institute of Nuclear Safety) ;
  • Oh, Young-Jin (Department of Structural Integrity and Materials, KEPCO Engineering and Construction Company)
  • 투고 : 2015.09.30
  • 심사 : 2016.05.02
  • 발행 : 2016.10.25
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초록

The accurate estimation of leak rate through cracks is crucial in applying the leak before break (LBB) concept to pipeline design in nuclear power plants. Because of its importance, several programs were developed based on the several proposed flow models, and used in nuclear power industries. As the flow models were developed for a homogeneous pipe material, however, some difficulties were encountered in estimating leak rates for bimaterial pipes. In this paper, a flow model is proposed to estimate leak rate in bimaterial pipes based on the modified Henry-Fauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based on the proposed flow model, a program was developed to estimate leak rate for a crack with linearly varying cross-sectional area. Using the program, leak rates were calculated for through-thickness cracks with constant or linearly varying cross-sectional areas in a bimaterial pipe. The leak rate results were then compared and discussed in comparison with the results for a homogeneous pipe. The effects of the crack morphology parameters and the variation in cross-sectional area on the leak rate were examined and discussed.

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참고문헌

  1. D.M. Norris, A. Okamoto, B. Chexal, T. Griesbach, PICEP: Pipe Crack Evaluation Program, EPRI Report Number EPRI NP-3596-SR, Electric Power Research Institute, Palo Alto, CA, 1984.
  2. H.S. Mehta, N.T. Patel, S. Ranganath, Application of the Leak-before-break Approach to BWR Piping, EPRI Report Number NP-4991, Electric Power Research Institute, Palo Alto, CA, 1986.
  3. D.D. Paul, J. Ahmad, P.M. Scott, L.E. Flanigan, G.M. Wilkowski, Evaluation and Refinement of Leak-rate Estimation Models, NUREG/CR-5128, Rev. 1, Battelle, Columbus (OH), 1994.
  4. D.O. Harris, D.D. Dedhia, S.C. Lu, Theoretical and User's Manual for Pc-PRAISE, USNRC Report NUREG/CR-5864, U.S. Nuclear Regulatory Commission, Washington, DC, 1992.
  5. D.O. Harris, D. Dedhia, WinPRAISE 07; Expanded PRAISE Code in Windows, Structural Integrity Associates, Inc., San Jose, CA, 2007.
  6. R.E. Henry, The two-phase critical discharge of initially saturated or subcooled liquid, Nucl. Sci. Eng. 41 (1970) 336-342. https://doi.org/10.13182/NSE70-A19091
  7. R.E. Henry, H.K. Fauske, Two-phase critical flow at low qualities, part 1: experimental, Nucl. Sci. Eng. 41 (1970) 79-91. https://doi.org/10.13182/NSE70-A20366
  8. R.E. Henry, H.K. Fauske, Two-phase critical flow at low qualities, part II: analysis, Nucl. Sci. Eng. 41 (1970) 92-98. https://doi.org/10.13182/NSE70-A20367
  9. S. Rahman, N. Ghadiali, D. Paul, G. Wilkowski, Probabilistic Pipe Fracture Evaluations for Leak-rate-detection Applications, USNRC Report NUREG/CR-6004, U.S. Nuclear Regulatory Commission, Washington, DC, 1995.
  10. PRO-LOCA-GUI/PRO-LOCA, User's Guide (Version 3.5.32), Battelle, Columbus, OH, 2009.
  11. J.H. Park, Y.K. Cho, S.H. Kim, J.H. Lee, Estimation of leak rate through circumferential cracks in pipes in nuclear power plants, Nucl. Eng. Technol. 47 (2015) 332-339. https://doi.org/10.1016/j.net.2014.11.008