The hydride stress reorientation behavior and the mechanical properties of irradiated cladding tubes were investigated to evaluate the high-burnup fuel-cladding tube properties in interim dry storage. As for the boiling water reactor ͑BWR͒ Zircaloy-2 ͑Zry-2͒ cladding, the hydride reorientation to the radial direction occurred at relatively low hoop stresses during the hydride reorientation treatment ͑HRT͒, such as less than 70 MPa. The increase of reorientation with hoop stress was not monotonic for the specimens in which a part of the hydrides remained precipitated at the HRT temperature, such as the case for 50GWd/t type cladding at a 300°C HRT. The degree of reorientation depended on the HRT solution temperature rather than on the estimated temperature at which the hydride precipitation occurred under the relatively moderate HRT conditions. In the relatively low cooling rate HRT, the hydride preferential precipitation in the Zr liner increased for Zr lined cladding compared to that in a relatively high cooling rate. The ductility of the specimens after the 300°C HRT showed relatively good correlation to the Polymax index which reflects the length or continuity of the hydrides regardless of their orientation. The ductility of the specimens after the 400°C, 0 MPa, 30°C/h HRT increased in ring compression testing at room temperature compared to no HRT ͑as-irradiated͒ specimens, and it indicated recovery of irradiation damage occurred at the 400°C annealing temperature and affected the ductility of the irradiated Zry-2 cladding. As for the pressurized water reactor Zircaloy-4 cladding, little increase in the radial hydride ratio occurred in a 100 MPa, 340°C or less HRT. On the other hand, the amount and the length of the hydride in the midwall area of the cladding depended on the temperature and the cooling rate from the HRT due to hydrogen migration from the hydride rim area. It is deduced that the ductility in ring compression deformation was affected by the orientation, amount, and length of hydride in the midwall area.
The hydride stress reorientation behavior and the mechanical properties of irradiated cladding tubes were investigated to evaluate the high-burnup fuel-cladding tube properties in interim dry storage. As for the boiling water reactor (BWR) Zircaloy-2 (Zry-2) cladding, the hydride reorientation to the radial direction occurred at relatively low hoop stresses during the hydride reorientation treatment (HRT), such as less than 70 MPa. The increase of reorientation with hoop stress was not monotonic for the specimens in which a part of the hydrides remained precipitated at the HRT temperature, such as the case for 50GWd/t type cladding at a 300°C HRT. The degree of reorientation depended on the HRT solution temperature rather than on the estimated temperature at which the hydride precipitation occurred under the relatively moderate HRT conditions. In the relatively low cooling rate HRT, the hydride preferential precipitation in the Zr liner increased for Zr lined cladding compared to that in a relatively high cooling rate. The ductility of the specimens after the 300°C HRT showed relatively good correlation to the Polymax index which reflects the length or continuity of the hydrides regardless of their orientation. The ductility of the specimens after the 400°C, 0 MPa, 30°C/h HRT increased in ring compression testing at room temperature compared to no HRT (as-irradiated) specimens, and it indicated recovery of irradiation damage occurred at the 400°C annealing temperature and affected the ductility of the irradiated Zry-2 cladding. As for the pressurized water reactor Zircaloy-4 cladding, little increase in the radial hydride ratio occurred in a 100 MPa, 340°C or less HRT. On the other hand, the amount and the length of the hydride in the midwall area of the cladding depended on the temperature and the cooling rate from the HRT due to hydrogen migration from the hydride rim area. It is deduced that the ductility in ring compression deformation was affected by the orientation, amount, and length of hydride in the midwall area.
The high-burnup BWR 9×9 lead use fuel assemblies (LUAs), which are designed for a maximum assembly burnup of 55 GWd/t in Japan, have been examined after irradiation in a commercial BWR to conrm the reliability of the current safety evaluation methodology and to accumulate data for judging the adequacy of its application to the future higher burnup fuel. The irradiation performance of 9×9 LUAs for two diŠerent designs, types A and B, is generally on the extended trend of 8×8 fuel, but some newˆndings in terms of fuel performance have been addressed after 5 cycle irradiations. Accelerated corrosion of cladding for the corner rods in Type B fuel assemblies and spacers in both types is observed after 5 cycle irradiations. The increasing trend of high hydrogen concentration seems to be an issue, which should be paid much attention with respect to fuel integrity during high-burnup irradiation. The large diŠerence inˆssion gas release rate between two types of fuel is conˆrmed after 3 and 5 cycle irradiations, and the release rate of Al Si O doped pellets is particularly higher than the others.
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