Hydride Reorientation and Delayed Hydride Cracking of Spent Fuel Rods in Dry Storage
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ODUCTION
THE most likely failure mechanisms of spent fuel rods are known to be cladding creep during vacuum drying[1,2] and their rupture by radial hydrides during handling after long-term storage.[3] The latter should be prevented to secure their retrievability. Due to these concerns, most of the studies conducted so far have focused on evaluation of creep ductility of spent fuel rods and determination of the critical stress for hydride reorientation. Experimental creep results indicate that the spent fuel rods have sufficiently high creep ductility even at 400 C, which is the recommended maximum temperature for drying, transfer, and dry cask storage.[4] Thus, it seems that thermal creep deformation of spent fuel rods will not affect their mechanical integrity. However, given that prior plastic deformation would promote nucleation of hydrides,[5] the creep deformation during vacuum drying would affect hydride reorientation, which thus far has not been appreciated. Furthermore, little attention has been paid to delayed hydride cracking (DHC) of spent fuel rods. The reason is based on the invalid hypothesis that DHC will not occur in the spent fuel rods due to limited stresses and slow diffusion of hydrogen at low temperatures below 200 C.[2,6] However, Simpson and Ells[7] reported a failure of
unirradiated Zr-2.5Nb fuel rods after their long-term storage at room temperature, the cause of which was recognized to be DHC. Therefore, it is clear that, in contrast to the belief that no DHC occurs in the spent fuel rods in dry storage, they would fail as long as stress raisers such as surface flaws or the weld region are present inside the cladding tube. Especially, high burnup fuel rods may have incipient cracks on the inside cladding surface due to an interaction of the fuel and the cladding during reactor operation.[3] The aim of this work is to understand the effect of creep on nucleation of reoriented hydrides in the radial direction (or radial hydrides) and to evaluate DHC susceptibility of spent fuel rods in dry storage. To these ends, we analyzed Tsai’s creep results[4] of the spent fuel rods in dry storage for 15 years, for the former, and Simpson and Ells’ observation[7] where a zirconium alloy cladding tube failed during long-term storage at room temperature, for the latter. As concrete evidence for the effect of prior plastic deformation on hydride reorientation, a model experiment was conducted using a cold-worked Zr-2.5Nb tube with 60 ppm H.[5]
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EXPERIMENTAL
[4]
YOUNG S. KIM, Leader of Zirconium Team, is with the Nuclear Material Research Division, Korea Atomic Energy Research Institute, Yuseong, Daejeon 305-353, Korea. Contact e-mail: yskim1@kaeri. re.kr This article is based on a presentation given in the symposium ‘‘Materials for the Nuclear Renaissance,’’ which occurred during the TMS Annual Meeting, February 15–19, 2009, in San Francisco, CA, under the auspices of Corrosion and Environmental Effects and the Nuclear Materials Committees of ASM-TMS. Article published online September 11, 2009 METALLURGIC
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