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      KCI등재 SCIE SCOPUS

      Estimation of Leak Rate Through Cracks in Bimaterial Pipes in Nuclear Power Plants

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      https://www.riss.kr/link?id=A103669904

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      다국어 초록 (Multilingual Abstract)

      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 pipematerial, 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 HenryeFauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based onthe 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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      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 mo...

      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 pipematerial, 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 HenryeFauske flow model. In the new flow model, different crack morphology parameters can be considered in two parts of a flow path. In addition, based onthe 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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      다국어 초록 (Multilingual Abstract)

      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 pipematerial, 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 onthe 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.
      번역하기

      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 mo...

      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 pipematerial, 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 onthe 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 (Reference)

      1 D.O. Harris, "WinPRAISE 07; Expanded PRAISE Code in Windows"

      2 PRO-LOCA-GUI/PRO-LOCA, "User's Guide (Version 3.5.32)"

      3 R.E. Henry, "Two-phase critical flow at low qualities, part II: analysis" 41 : 92-98, 1970

      4 R.E. Henry, "Two-phase critical flow at low qualities, part 1: experimental" 41 : 79-91, 1970

      5 D.O. Harris, "Theoretical and User's Manual for Pc-PRAISE" U.S. Nuclear Regulatory Commission 1992

      6 R.E. Henry, "The two-phase critical discharge of initially saturated or subcooled liquid" 41 : 336-342, 1970

      7 S. Rahman, "Probabilistic Pipe Fracture Evaluations for Leak-rate-detection Applications" U.S. Nuclear Regulatory Commission 1995

      8 D.M. Norris, "PICEP: Pipe Crack Evaluation Program" Electric Power Research Institute 1984

      9 D.D. Paul, "Evaluation and Refinement of Leak-rate Estimation Models" NUREG 1994

      10 박재학, "ESTIMATION OF LEAK RATE THROUGH CIRCUMFERENTIAL CRACKS IN PIPES IN NUCLEAR POWER PLANTS" 한국원자력학회 47 (47): 332-339, 2015

      1 D.O. Harris, "WinPRAISE 07; Expanded PRAISE Code in Windows"

      2 PRO-LOCA-GUI/PRO-LOCA, "User's Guide (Version 3.5.32)"

      3 R.E. Henry, "Two-phase critical flow at low qualities, part II: analysis" 41 : 92-98, 1970

      4 R.E. Henry, "Two-phase critical flow at low qualities, part 1: experimental" 41 : 79-91, 1970

      5 D.O. Harris, "Theoretical and User's Manual for Pc-PRAISE" U.S. Nuclear Regulatory Commission 1992

      6 R.E. Henry, "The two-phase critical discharge of initially saturated or subcooled liquid" 41 : 336-342, 1970

      7 S. Rahman, "Probabilistic Pipe Fracture Evaluations for Leak-rate-detection Applications" U.S. Nuclear Regulatory Commission 1995

      8 D.M. Norris, "PICEP: Pipe Crack Evaluation Program" Electric Power Research Institute 1984

      9 D.D. Paul, "Evaluation and Refinement of Leak-rate Estimation Models" NUREG 1994

      10 박재학, "ESTIMATION OF LEAK RATE THROUGH CIRCUMFERENTIAL CRACKS IN PIPES IN NUCLEAR POWER PLANTS" 한국원자력학회 47 (47): 332-339, 2015

      11 H.S. Mehta, "Application of the Leakbefore-break Approach to BWR Piping" Electric Power Research Institute 1986

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      학술지 이력

      학술지 이력
      연월일 이력구분 이력상세 등재구분
      2023 평가예정 해외DB학술지평가 신청대상 (해외등재 학술지 평가)
      2020-01-01 평가 등재학술지 유지 (해외등재 학술지 평가) KCI등재
      2014-01-01 평가 SCIE 등재 (등재유지) KCI등재
      2014-01-01 평가 SCOPUS 등재 (등재유지) KCI등재
      2011-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2009-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2007-01-01 평가 등재학술지 유지 (등재유지) KCI등재
      2006-07-31 학술지명변경 한글명 : Jorunal of the Korean Nuclear Society -> Nuclear Engineering and Technology
      외국어명 : 미등록 -> Nuclear Engineering and Technology
      KCI등재후보
      2004-01-01 평가 등재후보학술지 선정 (신규평가) KCI등재후보
      2003-01-01 평가 등재후보 1차 PASS (등재후보1차) KCI등재후보
      2002-01-01 평가 등재후보학술지 유지 (등재후보1차) KCI등재후보
      1999-01-01 평가 등재후보학술지 선정 (신규평가) KCI등재후보
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      학술지 인용정보

      학술지 인용정보
      기준연도 WOS-KCI 통합IF(2년) KCIF(2년) KCIF(3년)
      2016 1.04 0.17 0.77
      KCIF(4년) KCIF(5년) 중심성지수(3년) 즉시성지수
      0.63 0.56 0.343 0.11
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