Long-term Corrosion of Zircaloy-4 and Zircaloy-2 by Continuous Hydrogen Measurement under Repository Condition
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Long-term Corrosion of Zircaloy-4 and Zircaloy-2 by Continuous Hydrogen Measurement under Repository Condition Tomofumi Sakuragi, Hideaki Miyakawa, Tsutomu Nishimura1 and Tsuyoshi Tateishi2 Repository Engineering and EBS Technology Research Project, Radioactive Waste Management Funding and Research Center, 1-15-7 Tsukishima, Chuo City, Tokyo 104-0052, Japan 1 Kobe Steel, Ltd., 4-7-2 Iwaya-Nakamachi, Nada-ku, Kobe 657-0845, Japan 2 Kobelco Research Institute, Inc., 1-5-5 Takatsukadai, Nishi-ku, Kobe 657-2271, Japan ABSTRACT Corrosion behavior is a key issue for the waste disposal of irradiated metals, such as hulls and endpieces, and is considered to be a leaching source of radionuclides including C-14. However, little information about Zircaloy corrosion in anticorrosive conditions has been provided. In the present study, long-term corrosion tests of Zircaloy-4 and Zircaloy-2 were performed in assumed disposal conditions (dilute NaOH solution, pH 12.5, 303 K) by using the gas flow system for 1500 days. The corrosion rate, which was determined by measuring gaseous hydrogen and the hydrogen absorbed in Zircaloy, decreased with immersion time and was lower than the value of 2×10-2 ȝm/y used in performance assessment (1500-day values: 5.84×10-3 and 5.66×10-3 ȝm/y for Zircaloy-4, 1000-day values: 8.81×10-3 ȝm/y for Zircaloy-2). The difference in corrosion behavior between Zircaloy 4 and Zircaloy-2 was negligible. The average values of the hydrogen absorption ratios for Zircaloy-4 and Zircaloy-2 during corrosion were 91% and 94%, respectively. The hydrogen generation kinetics of both gas evolution and absorption into metal can be shown by a parabolic curve. This result indicates that the diffusion process controls the Zircaloy corrosion in the early corrosion stage of the present study, and that the thickness of the oxide film in this stage is limited to approximately 25 nm and may therefore be in the form of dense tetragonal zirconia. INTRODUCTION Spent fuel claddings after reprocessing are expected to be disposed in a deep underground repository. Corrosion of the irradiated metals raises a concern in the safety assessment of gas generation and the source term of activated radionuclides (e.g. C-14) [1]. However, the corrosion rate of Zircaloy under the anticorrosive repository conditions (low oxygen, high alkaline, and low temperature) is extremely slow. Therefore, the available data on Zircaloy corrosion are very limited. On the basis of numerous out-pile studies, it is widely accepted that the corrosion kinetics of Zircaloy follows a cubic rate law before transition (“breakaway”) in the temperature range of approximately 561 K to 673 K [2]. Although some disagreement remains, this empirical kinetic behavior is considered not only a simple diffusion process through the zirconia as a corrosion product, but also one of cracks and stress effects of the oxide on mass transport. One study has extrapolated the cubic law data to the corrosion rate at low temperatures [3], but the sufficiency of the extrapolation to corrosion b
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