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Article

Simulation of the fracture behavior of Zircaloy-4 cladding under reactivity-initiated accident conditions with a damage mechanics model combined with fuel performance codes FEMAXI-7 and RANNS

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Pages 208-219 | Received 09 Sep 2013, Accepted 21 Oct 2013, Published online: 15 Nov 2013

References

  • Fuketa T, Sugiyama T, Nagase F. Behavior of 60 to 78 MWd/kgU PWR fuels under reactivity-initiated accident conditions. J Nucl Sci Technol. 2006;43:1080–1088.
  • Fuketa T, Sasajima H, Sugiyama T. Behavior of high-burnup PWR fuels with low-tin Zircaloy-4 cladding under reactivity-initiated-accident conditions. Nucl Technol. 2001;133:50–62.
  • Tomiyasu K, Sugiyama T, Fuketa T. Influence of cladding-peripheral hydride on mechanical fuel failure under reactivity-initiated accident conditions. J Nucl Sci Technol. 2007;44:733–742.
  • Suzuki M, Saitou H, Fuketa T. Analysis of pellet-clad mechanical interaction process of high-burnup pwr fuel rods by RANNS code in reactivity-initiated accident conditions. Nucl Technol. 2006;155:282–292.
  • Dassault Systemes Simulia Corp. Abaqus 6.11 analysis user's manual. Providence (RI): Dassault Systemes Simulia Corp.; 2011. Chapter 23, Progressive Damage and Failure.
  • Nagase F, Fuketa T. Investigation of hydride rim effect on failure of Zircaloy-4 cladding with tube burst test. J Nucl Sci Technol. 2005;42:58–65.
  • Fukuda T. Development of pre-cracked cladding specimen for PCMI failure study. Paper presented at: Fuel Safety Research Meeting 2010; 2010 May 19–20; Tokai, Japan.
  • Fukuda T. The cladding fracture behavior under biaxial stress condition. IAEA-TECDOC-CD-1709. Mito (Japan): International Atomic Energy Agency; 2011.
  • Suzuki M, Saitou H, Udagawa Y. Light water reactor fuel analysis code FEMAXI-7; model and structure. JAEA-Data/Code 2010-035. Ibaraki (Japan): JAEA; 2010. Japanese.
  • Le Saux M, Besson Z, Carassou J, Poussard S, Averty X. A model to describe the anisotropic viscoplastic mechanical behavior of fresh and irradiated Zircaloy-4 fuel claddings under RIA loading conditions. J Nucl Mater. 2008;378:60–69.
  • Hillerborg A, Modder M, Petersson PE. Analysis of crack formation and crack growth in concrete by means of fracture mechanics and finite elements. Cement Concrete Res. 1976;6:773–781.
  • Grange M, Besson J, Andrieu E. An anisotropic Gurson type model to represent the ductile rupture of hydrided Zircaloy-4 sheets. Int J Fracture. 2000;105:273–293.
  • Desquines J, Koss DA, Motta AT, Cazalis B, Petit M. The issue of stress state during mechanical tests to assess cladding performance during a reactivity-initiated accident (RIA). J Nucl Mater. 2011;412:250–267.
  • Siefken LJ, Coryell EW, Harvego EA, Hohorst JK. SCDAP/RELAP5/MOD3.3 code manual: MATPRO – a library of material properties for light-water reactor accident analysis. NUREG/CR-6150. Washington (DC): US Nuclear Regulatory Commission; 2001.
  • Hagrman DL, Reymann GA. MATPRO-VERSION 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior. NUREG/CR-0497. Washington (DC): US Nuclear Regulatory Commission; 1981.
  • Ohira K, Itagaki N. Thermal conductivity measurements of high burnup UO2 pellet and a benchmark calculation of fuel center temperature. Paper presented at: Proceedings of the ANS Topical Meeting on Light Water Reactor Fuel Performance; 1997 Mar 2–6; Portland, OR. p. 541–549.
  • MacDonald PE, Thompson LB. MATPRO: version 09. A handbook of materials properties for use in the analysis of light water reactor fuel rod behavior. TREE-NUREG-1005. Washington (DC): US Nuclear Regulatory Commission; 1976.
  • Ross AM, Stoute RL. Heat transfer coefficient between UO2 and Zircaloy-2. CRFD-1075. Ontario (Canada): Atomic Energy Commission of Canada Limited; 1962.
  • Tachibana T, Furuya H, Koizumi M. Effect of temperature and deformation rate on fracture strength of sintered uranium dioxide. J Nucl Sci Technol. 1976;13:497–502.
  • Geelhood KJ, Luscher WG, Beyer CE, Cuta JM. FRAPTRAN 1.4: a computer code for the transient analysis of oxide fuel rods. NUREG/CR-7023. Washington (DC): US Nuclear Regulatory Commission; 2011.
  • Glendening A, Koss DA, Motta AT, Pierron ON, Daum RS. Failure of hydrided Zircaloy-4 under equal-biaxial and plane-strain tensile deformation. J ASTM Int. 2005;2:833–848.

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