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Report No.
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Integrity assessment of zircaloy fuel cladding tube experienced transient environmental history of spent fuel pool in Fukushima Dai-ichi Nuclear Power Plant

Sekio, Yoshihiro ; Yamagata, Ichiro ; Yamashita, Shinichiro  ; Sasaki, Shinji ; Ogawa, Ryuichiro; Mashiko, Shinichi; Hayashi, Takehiro; Inoue, Toshihiko ; Inoue, Masaki ; Maeda, Koji 

Corrosion and mechanical property tests utilizing spent fuel cladding made of zircaloy-2 were performed as a tentative test of the project for the purpose of simulating an environment at the very early stage after the accident in the SFP of unit 4 in the Fukushima Dai-ichi Nuclear Power Plant. The result of metallurgical investigation after corrosion test showed that no obvious changes in oxide film formed on the outer surface of cladding such as stripping occurred. In addition to that, the ring-tensile test results of samples after corrosion test was obtained and compared with that of samples before corrosion test, indicating that no significant degradation in mechanical property was confirmed. These results would have indicated that integrity of FAs was kept to be high as same as what it was before conducting corrosion test.

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