Mechanical Property Degradation of Unirradiated Zircaloy-4 Cladding After Creep Deformation

Abstract One of the limiting mechanisms of pressurized water reactor spent fuel cladding is creep owing to high temperature and rod internal pressure. Based on extensive studies, many countries have tentatively concluded that creep rupture is hard to occur under dry storage conditions and cannot severely degrade the integrity of the cladding if it meets the 400°C limitation owing to a self-limiting property. However, the changes in mechanical properties after creep deformation are not well understood due to the limited amount of relevant tests and analyses. In this regard, mechanical property degradation of unirradiated Zircaloy-4 cladding by creep deformation was investigated using a ring compression test and microscopic observation. In addition, the implication regarding spent fuel cladding integrity based on the test results is described.

[1]  J. Bair,et al.  A review on hydride precipitation in zirconium alloys , 2015 .

[2]  Ju-Seong Kim,et al.  A study on hydride reorientation of Zircaloy-4 cladding tube under stress , 2015 .

[3]  R. Kuo,et al.  Effect of Hydride Distribution on the Mechanical Properties of Zirconium-Alloy Fuel Cladding and Guide Tubes , 2015 .

[4]  Korukonda L. Murty,et al.  Creep Behavior of Hydrogenated Zirconium Alloys , 2014, Journal of Materials Engineering and Performance.

[5]  M. Daymond,et al.  Effect of thermo-mechanical cycling on zirconium hydride reorientation studied in situ with synchrotron X-ray diffraction , 2013 .

[6]  Michael C. Billone,et al.  Ductile-to-Brittle transition temperature for high-burnup cladding alloys exposed to simulated drying-storage conditions , 2013 .

[7]  C. Cappelaere,et al.  Thermal Creep Model for CWSR Zircaloy-4 Cladding Taking into Account the Annealing of the Irradiation Hardening , 2012 .

[8]  Y. Jung,et al.  Thermal creep of Zircaloy-4 tubes containing corrosion-induced hydrogen , 2011 .

[9]  L. Herranz,et al.  Creep assessment of Zry-4 cladded high burnup fuel under dry storage , 2011 .

[10]  Hyun-Gil Kim,et al.  THE EFFECTS OF CREEP AND HYDRIDE ON SPENT FUEL INTEGRITY DURING INTERIM DRY STORAGE , 2010 .

[11]  R. Kuo,et al.  Hydride reorientation in Zircaloy-4 cladding , 2008 .

[12]  Toyoshi Fuketa,et al.  Investigation of Hydride Rim Effect on Failure of Zircaloy-4 Cladding with Tube Burst Test , 2005 .

[13]  M. Billone,et al.  Thermal Creep of Irradiated Zircaloy Cladding , 2005 .

[14]  S. Lehmann,et al.  A creep rupture criterion for Zircaloy-4 fuel cladding under internal pressure , 2004 .

[15]  S. Yamanaka,et al.  Indentation creep study of zirconium hydrogen solid solution , 2004 .

[16]  P. Bouffioux,et al.  Irradiation Creep Behavior of Zr-Base Alloys , 2002 .

[17]  C. Domain,et al.  About the Mechanisms Governing the Hydrogen Effect on Viscoplasticity of Unirradiated Fully Annealed Zircaloy-4 Sheet , 2002 .

[18]  P. Bouffioux,et al.  Impact of Hydrogen on Plasticity and Creep of Unirradiated Zircaloy-4 Cladding Tubes , 2000 .

[19]  BF Kammenzind,et al.  The long range migration of hydrogen through Zircaloy in response to tensile and compressive stress gradients , 1998 .

[20]  D. D. Lanning,et al.  FRAPCON-3: A computer code for the calculation of steady-state, thermal-mechanical behavior of oxide fuel rods for high burnup , 1997 .

[21]  Harry Dipl Ing Spilker,et al.  Spent LWR fuel dry storage in large transport and storage casks after extended burnup , 1997 .

[22]  M. Mayuzumi,et al.  Creep deformation of an unirradiated zircaloy nuclear fuel cladding tube under dry storage conditions , 1990 .

[23]  B. Chin,et al.  Prediction of Maximum Allowable Temperatures for Dry Storage of Zircaloy-Clad Spent Fuel in Inert Atmosphere , 1989 .

[24]  Yutaka Matsuo,et al.  Thermal Creep of Zircaloy-4 Cladding under Internal Pressure , 1987 .

[25]  Y. Mishima,et al.  A resistometric study of the solution and precipitation of hydrides in unalloyed zirconium , 1968 .