ISSN 0253-2778

CN 34-1054/N

open
Free to Publish. Free to Read.
Open AccessOpen Access JUSTC Original Paper

Coupled thermo-mechanical and fracture analysis of a nuclear reactor pressure vessel

  • Key Laboratory of Special Purpose Equipment and Advanced Manufacturing Technology, Ministry of Education & Zhejiang Province, Zhejiang University of Technology, Hangzhou 310014, China

Cite this:
https://doi.org/10.3969/j.issn.0253-2778.2017.06.007
  • Received Date: 27 July 2016
  • Rev Recd Date: 08 December 2016
  • Publish Date: 30 June 2017
    Abstract

  • Abstract

    To study the influence of pressurized thermal shock on the bearing capacity of the nuclear reactor pressure vessel (RPV) with a surface crack, the 3D finite element model was established for the beltline region around the crack by ABAQUS software. The transient temperature field and stress field were obtained. The XFEM was used to simulate the crack propagation in the thermo-mechanical coupling field. The ultimate bearing capacity of the vessel at different nil-ductility transition temperatures was tested employing elastic-plastic fracture analysis, and the results were compared with those obtained from elastic analysis to sum up the fracture characteristics of the material. Through mesh refinement, it was found that the crack propagation paths and the damage degrees for different element sizes are equivalent, which avoids pathological mesh dependence.

    Abstract

    To study the influence of pressurized thermal shock on the bearing capacity of the nuclear reactor pressure vessel (RPV) with a surface crack, the 3D finite element model was established for the beltline region around the crack by ABAQUS software. The transient temperature field and stress field were obtained. The XFEM was used to simulate the crack propagation in the thermo-mechanical coupling field. The ultimate bearing capacity of the vessel at different nil-ductility transition temperatures was tested employing elastic-plastic fracture analysis, and the results were compared with those obtained from elastic analysis to sum up the fracture characteristics of the material. Through mesh refinement, it was found that the crack propagation paths and the damage degrees for different element sizes are equivalent, which avoids pathological mesh dependence.
    • FullText(HTML)
  • loading
    • References(20)
  • [1]
    MARINI B, AVERTY X, WIDENT P. Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel [J]. Journal of Nuclear Materials, 2015, 465: 20-27.
    [2]
    CHEN M, LU F, WANG R, Et al. The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading [J]. Nuclear Engineering and Design, 2015, 288: 130-140.
    [3]
    JHUNG M J, CHOI Y H. Critical crack depth diagram of reactor vessel for pressurized thermal shock [J]. Nuclear Engineering and Design, 2009, 239(3): 425-433.
    [4]
    PUGH C E, RICHARD BASS B, DICKSON T L. Role of probabilistic analysis in integrity assessment of reactor pressure vessels exposed to pressurized thermal-shock conditions [J]. Engineering Failure Analysis, 2007, 14(3): 501-517.
    [5]
    QIAN G, NIFFENEGGER M. Procedure, method and computer codes for probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks [J]. Nuclear Engineering and Design, 2013, 258: 35-50.
    [6]
    ASME boiler and pressure vessel code, Section XI, Appendix A 4000: Material properties [S]. New York: ASME, 2013.
    [7]
    RCC-M code 2000 edition (Plus 2002 Addendum) [S].Paris: French Association for Design Construction and In-Service Inspection Rules for Nuclear Island Components, 2002.
    [8]
    Fracture analysis of vessels-Oak Ridge FAVOR, v04.1. Computer code: Theory and implementation of algorithms, methods, and correlations: NUREG/CR-6854 [S]. Washington, DC: Nuclear Regulatory Commission, 2004.
    [9]
    KANTO Y, ONIZAWA K, MACHIDA H, et al. Recent Japanese research activities on probabilistic fracture mechanics for pressure vessel and piping of nuclear power plant [J]. International Journal of Pressure Vessels and Piping, 2010, 87(1): 11-16.
    [10]
    MOS N, BELYTSCHKO T. Extended finite element method for cohesive crack growth [J]. Engineering Fracture Mechanics, 2002, 69(7): 813-833.
    [11]
    MOS N, DOLBOW J, BELYTSCHKO T. A finite element method for crack growth without remeshing [J]. International Journal for Numerical Methods in Engineering, 1999, 46(1): 131-150.
    [12]
    Code of Federal Regulations, Title 10, Section 50.61. Fracture toughness requirements for protection against pressurized thermal shock events [S]. Washington, DC: Nuclear Regulatory Commission, 1984.
    [13]
    Regulatory Guide, No. 1.99, Revision 2. Radiation embrittlement of reactor vessel materials[S]. Washington, DC: Nuclear Regulatory Commission, 1988.
    [14]
    BENZEGGAGH M L, KENANE M. Measurement of mixed-mode delamination fracture toughness of unidirectional glass/epoxy composites with mixed-mode bending apparatus [J]. Composites Science and Technology, 1996, 56(4): 439-449.
    [15]
    范天佑. 断裂动力学原理与应用[M]. 北京：北京理工大学出版社，2006.
    [16]
    PASQUALINO I P, ESTEFEN S F. A nonlinear analysis of the buckle propagation problem in deepwater pipelines [J]. International Journal of Solids and Structures, 2001, 38(46-47): 8481-8502.
    [17]
    International Atomic Energy Agency. Pressurized thermal shock in nuclear power plants: Good practices for assessments[R]. IAEA, VIENNA: 2010: IAEA-TECDOC-1627.
    [18]
    TANGUY B, BESSON J, PIQUES R, et al. Ductile to brittle transition of an A508 steel characterized by Charpy impact test [J]. Engineering Fracture Mechanics, 2005, 72(1):49-72.
    [19]
    KARLSEN W, DIEGO G, DEVRIENT B. Localized deformation as a key precursor to initiation of intergranular stress corrosion cracking of austenitic stainless steels employed in nuclear power plants [J]. Journal of Nuclear Materials, 2010, 406(1):138-151.
    [20]
    Guide to methods of assessing the acceptability of flaws in fusion welded structures: British Standards BS 7910[S]. London: British Standards Institution, 1999.
    • Supplements(0)
    • Related Articles
    • Track Citations
  • 加载中

Catalog

    |
    Get Citation
    PDF
    XML
    [1]
    MARINI B, AVERTY X, WIDENT P. Effect of the bainitic and martensitic microstructures on the hardening and embrittlement under neutron irradiation of a reactor pressure vessel steel [J]. Journal of Nuclear Materials, 2015, 465: 20-27.
    [2]
    CHEN M, LU F, WANG R, Et al. The deterministic structural integrity assessment of reactor pressure vessels under pressurized thermal shock loading [J]. Nuclear Engineering and Design, 2015, 288: 130-140.
    [3]
    JHUNG M J, CHOI Y H. Critical crack depth diagram of reactor vessel for pressurized thermal shock [J]. Nuclear Engineering and Design, 2009, 239(3): 425-433.
    [4]
    PUGH C E, RICHARD BASS B, DICKSON T L. Role of probabilistic analysis in integrity assessment of reactor pressure vessels exposed to pressurized thermal-shock conditions [J]. Engineering Failure Analysis, 2007, 14(3): 501-517.
    [5]
    QIAN G, NIFFENEGGER M. Procedure, method and computer codes for probabilistic assessment of reactor pressure vessels subjected to pressurized thermal shocks [J]. Nuclear Engineering and Design, 2013, 258: 35-50.
    [6]
    ASME boiler and pressure vessel code, Section XI, Appendix A 4000: Material properties [S]. New York: ASME, 2013.
    [7]
    RCC-M code 2000 edition (Plus 2002 Addendum) [S].Paris: French Association for Design Construction and In-Service Inspection Rules for Nuclear Island Components, 2002.
    [8]
    Fracture analysis of vessels-Oak Ridge FAVOR, v04.1. Computer code: Theory and implementation of algorithms, methods, and correlations: NUREG/CR-6854 [S]. Washington, DC: Nuclear Regulatory Commission, 2004.
    [9]
    KANTO Y, ONIZAWA K, MACHIDA H, et al. Recent Japanese research activities on probabilistic fracture mechanics for pressure vessel and piping of nuclear power plant [J]. International Journal of Pressure Vessels and Piping, 2010, 87(1): 11-16.
    [10]
    MOS N, BELYTSCHKO T. Extended finite element method for cohesive crack growth [J]. Engineering Fracture Mechanics, 2002, 69(7): 813-833.
    [11]
    MOS N, DOLBOW J, BELYTSCHKO T. A finite element method for crack growth without remeshing [J]. International Journal for Numerical Methods in Engineering, 1999, 46(1): 131-150.
    [12]
    Code of Federal Regulations, Title 10, Section 50.61. Fracture toughness requirements for protection against pressurized thermal shock events [S]. Washington, DC: Nuclear Regulatory Commission, 1984.
    [13]
    Regulatory Guide, No. 1.99, Revision 2. Radiation embrittlement of reactor vessel materials[S]. Washington, DC: Nuclear Regulatory Commission, 1988.
    [14]
    BENZEGGAGH M L, KENANE M. Measurement of mixed-mode delamination fracture toughness of unidirectional glass/epoxy composites with mixed-mode bending apparatus [J]. Composites Science and Technology, 1996, 56(4): 439-449.
    [15]
    范天佑. 断裂动力学原理与应用[M]. 北京：北京理工大学出版社，2006.
    [16]
    PASQUALINO I P, ESTEFEN S F. A nonlinear analysis of the buckle propagation problem in deepwater pipelines [J]. International Journal of Solids and Structures, 2001, 38(46-47): 8481-8502.
    [17]
    International Atomic Energy Agency. Pressurized thermal shock in nuclear power plants: Good practices for assessments[R]. IAEA, VIENNA: 2010: IAEA-TECDOC-1627.
    [18]
    TANGUY B, BESSON J, PIQUES R, et al. Ductile to brittle transition of an A508 steel characterized by Charpy impact test [J]. Engineering Fracture Mechanics, 2005, 72(1):49-72.
    [19]
    KARLSEN W, DIEGO G, DEVRIENT B. Localized deformation as a key precursor to initiation of intergranular stress corrosion cracking of austenitic stainless steels employed in nuclear power plants [J]. Journal of Nuclear Materials, 2010, 406(1):138-151.
    [20]
    Guide to methods of assessing the acceptability of flaws in fusion welded structures: British Standards BS 7910[S]. London: British Standards Institution, 1999.
    Recommended articles
    TrendMD
    Volume 47 Issue 6 page: 491-497
    Cover

    Article Metrics

    Article views (409) PDF downloads(204)
    Proportional views

    Copyright © Editorial Office of JUSTC, All Rights Reserved. 皖ICP备05002528号

    Supported by: Beijing Renhe Information Technology Co. Ltd    

    /

    DownLoad:  Full-Size Img  PowerPoint
    Return
    Return