Embrittlement of zircaloy-4 due to oxidation in environment of stagnant steam.
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概要
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With a view to examining the embrittlement behavior of Zircaloy due to inner surface oxidation occurring in an LWR loss-of-coolant accident, ring-like Zircaloy-4 cladding specimens were heated at the isothermal oxidation temperature ranging 8901, 194°C in an environment of stagnant steam, which simulated the atmospheric condition inside the ruptured cladding.<BR>The embrittlement of the specimen due to oxidation in an environment of stagnant steam is influenced primarily by the amount of hydrogen absorbed by the Zircaloy-4. Ring compression tests conducted at 100°C on oxidized ring-like cladding specimen showed that Zircaloy containing more than about 500 wt. ppm of hydrogen had become brittle.<BR>The results of the present experiment support the idea supposed in the previous paper that the hydrogen absorption on the inner surface of the ruptured cladding, causing severe embrittlement, is strongly related with the atmospheric condition inside the ruptured cladding.
- 一般社団法人 日本原子力学会の論文
著者
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KAWASAKI SATORU
Division of Cardiology, National Akashi Hospital
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UETSUKA Hiroshi
Division of Reactor Safety, Japan Atomic Energy Research Institute
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UETSUKA Hiroshi
Division of Nuclear Safety Research, Japan Atomic Energy Research Institute
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FURUTA Teruo
Division of Reactor Safety, Japan Atomic Energy Research Institute
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KAWASAKI Satoru
Division of Reactor Safety, Japan Atomic Energy Research Institute
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FURUTA Teruo
Division of Nuclear Safety Research, Japan Atomic Energy Research Institute
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