A study on macroscopic fuel cladding ductile-to-brittle transition at 300°C induced by radial hydrides

References

• Corrosion of zirconium alloys in nuclear power plants. Austria: IAEA; 1993. IAEA-TECDOC-684.
   Google Scholar
• Hermann A, Martin M, Poerschke P, et al. Ductility degradation of irradiated fuel cladding. Vol. IV. Switzerland: Paul Scherrer Institute; 2001, Paul Scherrer Institute Scientific Report; 2000.
   Google Scholar
• Uchida M, Ichikawa M. Effect of pre-hydriding on post-irradiation ping-tensile properties of Zirclaoy-2 cladding tube. J Nucl Sci Technol. 1976;13:585–590.
   Web of Science ®Google Scholar
• Garde AM Effects of irradiation and hydriding on the mechanical properties of Zircaloy-4 at high fluences. Proc. Zirconium in the nuclear industry: Eith International Symposium, ASTM STP 1023; 1988 Jun; San Diego (United States).
   Google Scholar
• Bai JB, Prioul C, François D. Hydride embrittlement in Zircaloy-4 plate: part I. Influence of microstructure on the hydride embrittlement in Zircaloy-4 at 20 °C and 350 °C. Metall Mater Trans A. 1994;25:1185–1197.
   Web of Science ®Google Scholar
• Arsene S, Bai JB, Bompard P. Hydride embrittlement and irradiation effects on the hoop mechanical properties of PWR and BWR Zircaloy cladding tubes part 1. Metall Mater Trans A. 2003;34A:553–566.
   Web of Science ®Google Scholar
• Yagnik SK, Kuo R-C, Rashid YR, et al. Effect of hydrides on the mechanical properties of Zircaloy-4. Proc.: 2004 International Meeting on LWR Fuel Performance; 2004 Sep 19–22; Orlando (United States).
   Google Scholar
• Daum RS, Majumdar S, Liu Y, et al. Radial-hydride embrittlement of high-burnup Zircaloy-4 fuel cladding. J Nucl Sci Technol. 2006;43:1054–1067.
   Web of Science ®Google Scholar
• Billone MC, Burtseva TA, Einziger R. Ductile-to-brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions. J Nucl Mater. 2013;433:431–448.
   Web of Science ®Google Scholar
• Hsu H-H, Chiang M-F, Chen Y-C. The influence of hydride on fracture toughness of recrystallized Zircaloy-4 cladding. J Nucl Mater. 2014;447:56–62.
   Web of Science ®Google Scholar
• Kim J-S, Kim T-H, Kook D-H, et al. Effects of hydride morphology on the embrittlement of Zircaloy-4 cladding. J Nucl Mater. 2015;456:235–245.
   Web of Science ®Google Scholar
• Ogata K, Baba T, Kamimura K, et al. Effect of increased hydrogen content on the mechanical performance of irradiated cladding tubes. Proc. Top Fuel 2012; 2012 Sep 2–6; Manchester (United Kingdom).
   Google Scholar
• Aomi M, Baba T, Miyashita T, et al. Evaluation of hydride reorientation behavior and mechanical properties for high-burnup fuel-cladding tubes in interim dry storage. Proc. 15th International Symposium on Zirconium in the Nuclear Industry, ASTM STP 1505; 2007 Jun 24–28; Sunriver (United States).
   Google Scholar
• Louthan MR, Marshall RP. Control of hydride orientation in Zircaloy. J Nucl Mater. 1963;9:170–184.
   Web of Science ®Google Scholar
• Marshall RP. Influence of fabrication history on stress-oriented hydrides in Zircaloy tubing. J Nucl Mater. 1967;24:34–48.
   Web of Science ®Google Scholar
• Barraclough KG, Beevers CJ. Some observations on the deformation characteristics of bulk polycrystalline zirconium hydride part 1. The deformation and fracture of hydrides based on the delta-phase. J Mater Sci. 1969;4:518–525.
   Web of Science ®Google Scholar
• Stehle H, Kaden W, Manzel R. External corrosion of cladding in PWRs. Nucl Eng Des. 1975;33:155–169.
   Web of Science ®Google Scholar
• Howe LM, Thomas WR. The effect of neutron irradiation on the tensile properties of zircaloy-2. J Nucl Mater. 1960;2:248–260.
   Web of Science ®Google Scholar
• Higgy HR, Hammad FH. Effect of neutron irradiation on the tensile properties of Zircaloy-2 and Zircaloy-4. J Nucl Mater. 1972;44:215–227.
   Web of Science ®Google Scholar
• Torimaru T, Yasuda T, Nakatsuka M. Changes in mechanical properties of irradiated Zircaloy-2 fuel cladding due to short term annealing. J Nucl Mater. 1996;238:169–174.
   Web of Science ®Google Scholar
• Ito K, Kamimura K, Tsukuda Y. Evaluation of irradiation effect on spent fuel cladding creep properties. Proc. 2004 International Meeting on LWR Fuel Performance; 2004 Sep 19–22; Orlando; United States.
   Google Scholar
• Hirose T, Ozawa M, Miura H, et al. 2013: Research on integrity of high burnup spent fuel under long term dry storage and transport. France: OECD/NEA/CSNI, NEA: CSNI/R; 2013. 10.
   Google Scholar
• Raynaud PA, Koss DA, Motta AT. Crack growth in the through-thickness direction of hydrided thin-wall zircaloy sheet. J Nucl Mater. 2012 Jan;420:69–82.
   Web of Science ®Google Scholar
• Wallace AC, Shek GK, Lepik OE. Effects of hydride morphology on Zr-2.5Nb fracture toughness. Proc.: Zirconium in the nuclear industry: Eith International Symposium, ASTM STP 1023; 1988 Jun; San Diego (United States).
   Google Scholar
• Ogata K, Aomi M, Baba T, et al. Progress in the research programs to elucidate axial cracking fuel failure at high burnup. Proc.: 2007 International LWR Fuel Performance Meeting; 2007 Sep; San Francisco (United States).
   Google Scholar
• ASTM B811-02. Standard Specification for Wrought Zirconium Alloy Seamless Tubes for Nuclear Reactor Fuel Cladding, American Society for Testing and Materials International; 2007.
   Google Scholar
• JIS H 4751:2016(E). Zirconium alloy tubes. Japanese Standards Association.
   Google Scholar
• Kim JH, Lee MH, Choi BK, et al. Effect of the hydrogen contents on the circumferential mechanical properties of zirconium alloy claddings. J Alloys Compd. 2007;431:155–161.
   Web of Science ®Google Scholar
• Yamanaka S, Setoyama D, Muta H, et al. Characteristics of zirconium hydrogen solid solution. J Alloy Compd. 2004 Jun;372:129–135.
   Web of Science ®Google Scholar
• Une K, Ishimoto S. Dissolution and precipitation behavior of hydrides in Zircaloy-2 and high Fe Zircaloy. J Nucl Mater. 2003;322:66–72.
   Web of Science ®Google Scholar
• Une K, Ishimoto S, Etoh Y, et al. The terminal solid solubility of hydrogen in irradiated Zircaloy-2 and microscopic modeling of hydride behavior. J Nucl Mater. 2009;389:127–136.
   Web of Science ®Google Scholar
• Mishima Y, Okubo T, Ohishi M, et al. Proving test on reliability for BWR fuel assembies. J Atom Energ Soc Jpn. 1987 Feb;29:90–115.
   Web of Science ®Google Scholar
• Miyashita T, Nakae N, Ogata K, et al. Verification program of BWR 9 × 9 fuel. Trans Atomic Energy Soc Jpn. 2008 Dec;7:380–395.
   Google Scholar
• Mishima Y. Cooperative study for standardization of testing procedures for Zircaloy cladding tubes for BWR, using domestically manufactured ones. J Atom Energ Soc Jpn. 1974 Jun;16:306–311.
   Google Scholar
• Lee D, Adamson RB Modeling of localized deformation in neutron irradiated zircaloy-2. Proc.: 3rd International Conference on Zirconium in the Nuclear Industry, ASTM STP 633; 1976 Aug 10–12; Quebec (Canada).
   Google Scholar
• Yasuda T, Nakatsuka M, Yamahsita K Deformation and fracture properties of neutron-irradiated recrystallized zircaloy-2 cladding under uniaxial tension. Proc.: 7th International Symposium on Zirconium in the Nuclear Industry, ASTM STP 939; 1985 Jun 24–27; Strasbourg (France).
   Google Scholar