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Volumn 7, Issue , 2007, Pages 529-566

7.14 - The Performance of Large-scale Structures and Validation of Assessment Procedures

Author keywords

[No Author keywords available]

Indexed keywords

ASSESSMENT PROCEDURE; LARGE SCALE STRUCTURES;

EID: 33747188240     PISSN: None     EISSN: None     Source Type: Book    
DOI: 10.1016/B0-08-043749-4/07103-2     Document Type: Chapter
Times cited : (3)

References (68)
  • 1
    • 84903416723 scopus 로고    scopus 로고
    • ASME, Boiler and Pressure Vessel Code, Section III, Rules for construction of nuclear facility components, Division 1, Appendix G, "Protection against Non-Ductile Failure," American Society for Mechanical Engineers.
    • ASME, 2001a, Boiler and Pressure Vessel Code, Section III, Rules for construction of nuclear facility components, Division 1, Appendix G, "Protection against Non-Ductile Failure," American Society for Mechanical Engineers.
    • (2001)
  • 2
    • 84903416724 scopus 로고    scopus 로고
    • ASME, Boiler and Pressure Vessel Code, Section XI, Rules for inservice inspection of nuclear power plant components, non-mandatory Appendix A, Analysis of flaws, American Society for Mechanical Engineers.
    • ASME, 2001b, Boiler and Pressure Vessel Code, Section XI, Rules for inservice inspection of nuclear power plant components, non-mandatory Appendix A, Analysis of flaws, American Society for Mechanical Engineers.
    • (2001)
  • 3
    • 84903416725 scopus 로고
    • "CSNI Project from Fracture Analyses of Large-scale International Reference Experiments (Project FALSIRE)," Oak Ridge National Laboratory report no. NUREG/CR-5997 (ORNL/TM-12307).
    • B. R. Bass, et al., 1993, "CSNI Project from Fracture Analyses of Large-scale International Reference Experiments (Project FALSIRE)," Oak Ridge National Laboratory report no. NUREG/CR-5997 (ORNL/TM-12307).
    • (1993)
    • Bass, B.R.1
  • 5
    • 84903416714 scopus 로고    scopus 로고
    • "Effects of Cladding," European Commission DG-JRC/IAM, Petten, Report no. NESC-DOC ETF (99) 27 (EUR 19633 EN).
    • B. R. Bass, C. Faidy and R. Murgatroyd, 2000b, "Effects of Cladding," European Commission DG-JRC/IAM, Petten, Report no. NESC-DOC ETF (99) 27 (EUR 19633 EN).
    • (2000)
    • Bass, B.R.1    Faidy, C.2    Murgatroyd, R.3
  • 6
    • 84903416715 scopus 로고    scopus 로고
    • "NESC-IV Project: An Interim Report," European Commission, Directorate-General Joint Research Center, Petten, Report no. NESCDOC MAN (02) 04.
    • B. R. Bass, W. J. McAfee, P. T. Williams, D. J. Swan, N. Taylor, K. Nilsson and P. Minnebo, 2002, "NESC-IV Project: An Interim Report," European Commission, Directorate-General Joint Research Center, Petten, Report no. NESCDOC MAN (02) 04.
    • (2002)
    • Bass, B.R.1    McAfee, W.J.2    Williams, P.T.3    Swan, D.J.4    Taylor, N.5    Nilsson, K.6    Minnebo, P.7
  • 8
    • 84903416717 scopus 로고    scopus 로고
    • Final report, NESC-I project overview, European Commission, Directorate General Joint Research Centre, Petten.
    • B. R. Bass, J. B. Wintle, R. C. Hurst and N. Taylor, 2001, Final report, NESC-I project overview, European Commission, Directorate General Joint Research Centre, Petten.
    • (2001)
    • Bass, B.R.1    Wintle, J.B.2    Hurst, R.C.3    Taylor, N.4
  • 9
    • 84903416708 scopus 로고    scopus 로고
    • BEGL, R6: Assessment of the integrity of structures containing defects, Revision 4, British Energy Generation, Gloucester.
    • BEGL, 2001, R6: Assessment of the integrity of structures containing defects, Revision 4, British Energy Generation, Gloucester.
    • (2001)
  • 13
    • 84903416709 scopus 로고
    • "Pressurized-thermal-shock Test of 6-in.-thick Pressure Vessels, PTSE-1: Investigation of Warm Prestressing and Upper-shelf Arrest," Oak Ridge National Laboratory report no. NUREG/CR-4106 (ORNL-6135).
    • R. H. Bryan, et al., 1985a, "Pressurized-thermal-shock Test of 6-in.-thick Pressure Vessels, PTSE-1: Investigation of Warm Prestressing and Upper-shelf Arrest," Oak Ridge National Laboratory report no. NUREG/CR-4106 (ORNL-6135).
    • (1985)
    • Bryan, R.H.1
  • 14
    • 0021516847 scopus 로고
    • Results and conclusions from the first pressurized-thermal-shock experiment
    • Bryan R.H., et al. Results and conclusions from the first pressurized-thermal-shock experiment. Int. J. Nucl. Eng. Des. 1985, 89:145-159.
    • (1985) Int. J. Nucl. Eng. Des. , vol.89 , pp. 145-159
    • Bryan, R.H.1
  • 15
    • 84903416710 scopus 로고
    • "Pressurized-thermal-shock Test of 6-in.-thick Pressure Vessel, PTSE-2: Investigation of Low Tearing Resistance and Warm Prestressing," Oak Ridge National Laboratory report no. NUREG/CR-4888 (ORNL-6377).
    • R. H. Bryan, et al., 1987, "Pressurized-thermal-shock Test of 6-in.-thick Pressure Vessel, PTSE-2: Investigation of Low Tearing Resistance and Warm Prestressing," Oak Ridge National Laboratory report no. NUREG/CR-4888 (ORNL-6377).
    • (1987)
    • Bryan, R.H.1
  • 17
    • 84903416711 scopus 로고    scopus 로고
    • BSI, BS 7910:1999: Guide on methods for assessing the acceptability of flaws in metallic structures, incorporating amendment 1, British Standards Institution, London.
    • BSI, 2000, BS 7910:1999: Guide on methods for assessing the acceptability of flaws in metallic structures, incorporating amendment 1, British Standards Institution, London.
    • (2000)
  • 18
    • 84903416712 scopus 로고
    • "Preliminary Development of an Integrated Approach to the Evaluation of Pressurized Thermal Shock Risk as Applied to the Oconee Unit 1 Nuclear Power Plant," Oak Ridge National Laboratory report no. NUREG/CR-3770 (ORNL/TM-9176).
    • T. J. Burns, et al., 1986, "Preliminary Development of an Integrated Approach to the Evaluation of Pressurized Thermal Shock Risk as Applied to the Oconee Unit 1 Nuclear Power Plant," Oak Ridge National Laboratory report no. NUREG/CR-3770 (ORNL/TM-9176).
    • (1986)
    • Burns, T.J.1
  • 19
    • 0020748976 scopus 로고
    • Fracture mechanics data deduced from thermal-shock and related experiments with LWR pressure vessel material
    • Cheverton R.D., et al. Fracture mechanics data deduced from thermal-shock and related experiments with LWR pressure vessel material. ASME J Pres. Ves. Technol. 1983, 105:102-110.
    • (1983) ASME J Pres. Ves. Technol. , vol.105 , pp. 102-110
    • Cheverton, R.D.1
  • 20
    • 84903416713 scopus 로고
    • "Pressure Vessel Fracture Studies Pertaining to the PWR Thermal-shock Issue: Experiments TSE-5, TSE-5A, and TSE-6," Oak Ridge National Laboratory report no. NUREG/CR-4249 (ORNL-6163).
    • R. D. Cheverton, et al., 1985, "Pressure Vessel Fracture Studies Pertaining to the PWR Thermal-shock Issue: Experiments TSE-5, TSE-5A, and TSE-6," Oak Ridge National Laboratory report no. NUREG/CR-4249 (ORNL-6163).
    • (1985)
    • Cheverton, R.D.1
  • 23
    • 84903416706 scopus 로고    scopus 로고
    • "Pressurized Thermal Shock Methodologies and Validation," International Atomic Energy Agency Technical Report Series (IAEA/TRS-XXX), to be published.
    • L. M. Davies, et al., 2002, "Pressurized Thermal Shock Methodologies and Validation," International Atomic Energy Agency Technical Report Series (IAEA/TRS-XXX), to be published.
    • (2002)
    • Davies, L.M.1
  • 24
    • 84903416707 scopus 로고
    • "Test of 6-in-thick Pressure Vessels, Series 1: Intermediate Test Vessels V-1 and V-2," Oak Ridge National Laboratory report no. ORNL-4895.
    • R.W. Derby, et al., 1974, "Test of 6-in-thick Pressure Vessels, Series 1: Intermediate Test Vessels V-1 and V-2," Oak Ridge National Laboratory report no. ORNL-4895.
    • (1974)
    • Derby, R.W.1
  • 27
    • 84903416700 scopus 로고    scopus 로고
    • "A16: Guide for Defect Assessment and Leak Before Break Analysis," Commissariat a l'Energie Atomique (CEA) report no. SEMT/LISN/RT (Draft 10-Dec-1999), Saclay.
    • S. Drubay, S. Chapuliot and M.-H. Lacire, 1999, "A16: Guide for Defect Assessment and Leak Before Break Analysis," Commissariat a l'Energie Atomique (CEA) report no. SEMT/LISN/RT (Draft 10-Dec-1999), Saclay.
    • (1999)
    • Drubay, S.1    Chapuliot, S.2    Lacire, M.-H.3
  • 29
    • 84903416702 scopus 로고    scopus 로고
    • "Structural Integrity of Bi-metallic Components (BIMET)," Final summary report, Euratom Research Framework Programme 1994--1998, Nuclear Fission Safety.
    • C. Faidy, 2002, "Structural Integrity of Bi-metallic Components (BIMET)," Final summary report, Euratom Research Framework Programme 1994--1998, Nuclear Fission Safety.
    • (2002)
    • Faidy, C.1
  • 30
    • 84903416703 scopus 로고
    • "Survey of National Regulatory Requirements, Aging Materials Evaluations and Studies (AMES)," Report no. 4, European Commission DG XII.
    • R. Gerard, 1995, "Survey of National Regulatory Requirements, Aging Materials Evaluations and Studies (AMES)," Report no. 4, European Commission DG XII.
    • (1995)
    • Gerard, R.1
  • 31
    • 84903416704 scopus 로고    scopus 로고
    • "BIMET---Structural Integrity of Bi-metallic Components: Three-dimensional Finite Element Analysis of Test BIMET01," TWI report no. 8294/3A/99.
    • M. R. Goldthorpe and C. S. Wiesner, 1999, "BIMET---Structural Integrity of Bi-metallic Components: Three-dimensional Finite Element Analysis of Test BIMET01," TWI report no. 8294/3A/99.
    • (1999)
    • Goldthorpe, M.R.1    Wiesner, C.S.2
  • 32
    • 84903416694 scopus 로고
    • An assessment of the integrity of PWR pressure vessels, addendum to the second report of the study group, since 1982 under the chairmanship of Sir P. B. Hirsch.
    • P. B. Hirsch, 1987, An assessment of the integrity of PWR pressure vessels, addendum to the second report of the study group, since 1982 under the chairmanship of Sir P. B. Hirsch.
    • (1987)
    • Hirsch, P.B.1
  • 33
    • 84903416695 scopus 로고
    • "J--R Curve Characterization of Irradiated Low Upper Shelf Welds," Materials Engineering Associates, Report no. NUREG/CR-3506 (MEA-2038).
    • A. L. Hiser, et al., 1984, " J-- R Curve Characterization of Irradiated Low Upper Shelf Welds," Materials Engineering Associates, Report no. NUREG/CR-3506 (MEA-2038).
    • (1984)
    • Hiser, A.L.1
  • 34
    • 84903416696 scopus 로고    scopus 로고
    • IAEA, "Guidelines on Pressurised Thermal Shock Analysis for WWER Nuclear Power Plants," International Atomic Energy Agency, IAEA report no. IAEA-EBP-WWER-08.
    • IAEA, 1997, "Guidelines on Pressurised Thermal Shock Analysis for WWER Nuclear Power Plants," International Atomic Energy Agency, IAEA report no. IAEA-EBP-WWER-08.
    • (1997)
  • 35
    • 84903416697 scopus 로고
    • JWC, "Structural Integrity of Very Thick Steel Plate for Nuclear Reactor Pressure Vessels," Japan Welding Council report no. JWES-AE-7806, Tokyo.
    • JWC, 1977, "Structural Integrity of Very Thick Steel Plate for Nuclear Reactor Pressure Vessels," Japan Welding Council report no. JWES-AE-7806, Tokyo.
    • (1977)
  • 36
    • 0035908024 scopus 로고    scopus 로고
    • Potential roles for the Master curve in regulatory applications
    • Kirk M., Mitchell M. Potential roles for the Master curve in regulatory applications. Int. J. Pres. Ves. Piping 2001, 78:111-123.
    • (2001) Int. J. Pres. Ves. Piping , vol.78 , pp. 111-123
    • Kirk, M.1    Mitchell, M.2
  • 38
    • 0035908010 scopus 로고    scopus 로고
    • Key features arising from structural analysis in the NESC-I benchmark experiment
    • Lidbury D.P.G., et al. Key features arising from structural analysis in the NESC-I benchmark experiment. Int. J. Pres. Ves. Piping 2001, 78:225-236.
    • (2001) Int. J. Pres. Ves. Piping , vol.78 , pp. 225-236
    • Lidbury, D.P.G.1
  • 40
    • 0008528294 scopus 로고
    • Significance of warm prestress to crack initiation during thermal shock
    • Loss F.J., Gray R.A., Hawthorne J.R. Significance of warm prestress to crack initiation during thermal shock. Int. J. Nucl. Eng. Des. 1978, 46:395-408.
    • (1978) Int. J. Nucl. Eng. Des. , vol.46 , pp. 395-408
    • Loss, F.J.1    Gray, R.A.2    Hawthorne, J.R.3
  • 41
    • 0018542084 scopus 로고
    • Investigation of warm prestress for the case of small delta T during a reactor loss-of-coolant accident
    • Loss F.J., Gray R.A., Hawthorne J.R. Investigation of warm prestress for the case of small delta T during a reactor loss-of-coolant accident. ASME, J. Pres. Ves. Technol. 1979, 101:298-304.
    • (1979) ASME, J. Pres. Ves. Technol. , vol.101 , pp. 298-304
    • Loss, F.J.1    Gray, R.A.2    Hawthorne, J.R.3
  • 42
    • 3342918533 scopus 로고    scopus 로고
    • Introduction of fatigue cracks into the large thick-walled hollow cylinder for the NESC-I project
    • K. May, et al., 1998, Introduction of fatigue cracks into the large thick-walled hollow cylinder for the NESC-I project. In: "Proceedings of the ASME Pressure Vessels and Piping Conf.," vol. 362, p. 305.
    • (1998) "Proceedings of the ASME Pressure Vessels and Piping Conf.," , vol.362 , pp. 305
    • May, K.1
  • 43
    • 84903416688 scopus 로고
    • "Test of 6-in-thick Pressure Vessels, Series 4: Intermediate Test Vessels V-5, and V-9," Oak Ridge National Laboratory report no. ORNL/NUREG-7.
    • J. G. Merkle, et al., 1977, "Test of 6-in-thick Pressure Vessels, Series 4: Intermediate Test Vessels V-5, and V-9," Oak Ridge National Laboratory report no. ORNL/NUREG-7.
    • (1977)
    • Merkle, J.G.1
  • 44
    • 84903416689 scopus 로고
    • "An Examination of the Size Effects and Data Scatter Observed in Small-specimen Cleavage Fracture Toughness Testing," Oak Ridge National Laboratory report no. ORNL-6377.
    • J. G. Merkle, et al., 1984, "An Examination of the Size Effects and Data Scatter Observed in Small-specimen Cleavage Fracture Toughness Testing," Oak Ridge National Laboratory report no. ORNL-6377.
    • (1984)
    • Merkle, J.G.1
  • 45
    • 84903416690 scopus 로고    scopus 로고
    • "Technical Basis for an ASTM Standard on Determining the Reference Temperature, T0, for Ferritic Steels in the Transition Range," Oak Ridge National Laboratory report no. NUREG/CR-5504 (ORNL/TM-13631), November 1998.
    • J. G. Merkle, K. Wallin and D. E. McCabe, 1998, "Technical Basis for an ASTM Standard on Determining the Reference Temperature, T0, for Ferritic Steels in the Transition Range," Oak Ridge National Laboratory report no. NUREG/CR-5504 (ORNL/TM-13631), November 1998.
    • (1998)
    • Merkle, J.G.1    Wallin, K.2    McCabe, D.E.3
  • 46
    • 0023855906 scopus 로고
    • Review of limit loads of structures containing defects
    • Miller A.G. Review of limit loads of structures containing defects. Int. J. Pres. Ves. Piping 1988, 32:197-327.
    • (1988) Int. J. Pres. Ves. Piping , vol.32 , pp. 197-327
    • Miller, A.G.1
  • 47
    • 84903416691 scopus 로고
    • "Crack Arrest Behavior in SEN Wide Plates of Quenched and Tempered A533B Steel Tested under Nonisothermal Conditions," Oak Ridge National Laboratory report no. NUREG/CR-4930 (ORNL-6388).
    • D. J. Naus, et al., 1987, "Crack Arrest Behavior in SEN Wide Plates of Quenched and Tempered A533B Steel Tested under Nonisothermal Conditions," Oak Ridge National Laboratory report no. NUREG/CR-4930 (ORNL-6388).
    • (1987)
    • Naus, D.J.1
  • 49
    • 84903416693 scopus 로고
    • NRC, "Pressurized Thermal Shock (PTS)," U.S. Nuclear Regulatory Commission report no. SECY-82-465.
    • NRC, 1982, "Pressurized Thermal Shock (PTS)," U.S. Nuclear Regulatory Commission report no. SECY-82-465.
    • (1982)
  • 51
    • 84903416646 scopus 로고
    • "Pressure Vessel Thermal Shock at U.S. Pressurized Water Reactors: Events and Precursors, 1963 to mid-1981," Oak Ridge National Laboratory report no. ORNL/NSIC-112.
    • D. L. Phung, 1982, "Pressure Vessel Thermal Shock at U.S. Pressurized Water Reactors: Events and Precursors, 1963 to mid-1981," Oak Ridge National Laboratory report no. ORNL/NSIC-112.
    • (1982)
    • Phung, D.L.1
  • 52
    • 84903416647 scopus 로고    scopus 로고
    • "A Review of Large-scale Experiments Relevant to Pressure Vessel Integrity under PTS Conditions," Oak Ridge National Laboratory report no. NUREG/CR-6699 (ORNL/TM-2000/360).
    • C. E. Pugh and B. R. Bass, 2001, "A Review of Large-scale Experiments Relevant to Pressure Vessel Integrity under PTS Conditions," Oak Ridge National Laboratory report no. NUREG/CR-6699 (ORNL/TM-2000/360).
    • (2001)
    • Pugh, C.E.1    Bass, B.R.2
  • 54
    • 0035908118 scopus 로고    scopus 로고
    • Consistency in fracture assessment criteria
    • Rintamaa R., et al. Consistency in fracture assessment criteria. Int. J. Pres. Ves. Piping 2001, 78:125-135.
    • (2001) Int. J. Pres. Ves. Piping , vol.78 , pp. 125-135
    • Rintamaa, R.1
  • 57
    • 84903416635 scopus 로고
    • "Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant," Oak Ridge National Laboratory report no. NUREG/CR-4022 (ORNL/TM-9408).
    • D. L. Selby, et al., 1985a, "Pressurized Thermal Shock Evaluation of the Calvert Cliffs Unit 1 Nuclear Power Plant," Oak Ridge National Laboratory report no. NUREG/CR-4022 (ORNL/TM-9408).
    • (1985)
    • Selby, D.L.1
  • 58
    • 84903416636 scopus 로고
    • "Pressurized Thermal Shock Evaluation of the H. B. Robinson Unit 2 Nuclear Power Plant," Oak Ridge National Laboratory report no. NUREG/CR-4183 (ORNL/TM-9567).
    • D. L. Selby, et al., 1985b, "Pressurized Thermal Shock Evaluation of the H. B. Robinson Unit 2 Nuclear Power Plant," Oak Ridge National Laboratory report no. NUREG/CR-4183 (ORNL/TM-9567).
    • (1985)
    • Selby, D.L.1
  • 60
    • 84903416638 scopus 로고
    • "CSNI Project for Fracture Analyses of Large-Scale International Reference Experiments (Phase I)," Gesellschaft für Anlagen-und Reaktorsicherheit report no. GRS-108, NEA/CSNI/R(94)12.
    • J. Sievers, et al., 1994, "CSNI Project for Fracture Analyses of Large-Scale International Reference Experiments (Phase I)," Gesellschaft für Anlagen-und Reaktorsicherheit report no. GRS-108, NEA/CSNI/R(94)12.
    • (1994)
    • Sievers, J.1
  • 61
    • 84903416639 scopus 로고    scopus 로고
    • SINTAP, Structural integrity assessment procedures for European industry, final procedure, EU Brite-Euram Project BE95-1426.
    • SINTAP, 1999, Structural integrity assessment procedures for European industry, final procedure, EU Brite-Euram Project BE95-1426.
    • (1999)
  • 62
    • 0031624294 scopus 로고    scopus 로고
    • A comparison of different failure assessment methodologies applied to the NESC-I test.
    • M. Smith, et al., 1998, A comparison of different failure assessment methodologies applied to the NESC-I test. In: "Proceedings of the ASME Pressure Vessels and Piping Conf.," vol. 362, pp. 281--287.
    • (1998) "Proceedings of the ASME Pressure Vessels and Piping Conf.," , vol.362 , pp. 281-287
    • Smith, M.1
  • 63
    • 84903416625 scopus 로고
    • USCFR, Title 10, Part 50, Article 61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Code of Federal Regulation, Office of the Federal Register, National Archives and Records Administration, Washington, DC, Revision.
    • USCFR, 1987, Title 10, Part 50, Article 61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, U.S. Code of Federal Regulation, Office of the Federal Register, National Archives and Records Administration, Washington, DC, Revision.
    • (1987)
  • 64
    • 84903416618 scopus 로고
    • USNRC, Regulatory Guide 1.154, Format and content of plant specific pressurized thermal shock safety analysis report for pressurized water reactors, U.S. Nuclear Regulatory Commission, Washington, DC.
    • USNRC, 1987, Regulatory Guide 1.154, Format and content of plant specific pressurized thermal shock safety analysis report for pressurized water reactors, U.S. Nuclear Regulatory Commission, Washington, DC.
    • (1987)
  • 65
    • 84903416620 scopus 로고
    • USNRC, Regulatory Guide 1.99, Revision 2, Radiation embrittlement of reactor vessel materials, U.S. Nuclear Regulatory Commission.
    • USNRC, 1988, Regulatory Guide 1.99, Revision 2, Radiation embrittlement of reactor vessel materials, U.S. Nuclear Regulatory Commission.
    • (1988)
  • 66
    • 84903416621 scopus 로고    scopus 로고
    • WCAP, Integrity evaluation for future operation of Virgil C Summer nuclear plant reactor vessel nozzle to pipe weld regions, WCAP-15617.
    • WCAP, 2000, Integrity evaluation for future operation of Virgil C Summer nuclear plant reactor vessel nozzle to pipe weld regions, WCAP-15617.
    • (2000)
  • 67
    • 84903416622 scopus 로고
    • "Historical Summary of the Heavy-section Steel Technology Program and some Related Activities in Light Water Reactor Pressure Vessel Safety Research," Oak Ridge National Laboratory report no. NUREG/CR-4489 (ORNL-6259).
    • G. D. Whitman, 1986, "Historical Summary of the Heavy-section Steel Technology Program and some Related Activities in Light Water Reactor Pressure Vessel Safety Research," Oak Ridge National Laboratory report no. NUREG/CR-4489 (ORNL-6259).
    • (1986)
    • Whitman, G.D.1


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