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-ray irradiation on pitting corrosion of fuel cladding in water containing sea saltMotooka, Takafumi; Tsukada, Takashi
Proceedings of 2014 Nuclear Plant Chemistry Conference (NPC 2014) (USB Flash Drive), 9 Pages, 2014/10
In Fukushima Daiichi Nuclear Power Station (1F), seawater was injected into spent fuel pools in March 2011. Zircaloy-2 is adopted for the fuel cladding at 1F. Zirconium alloys including Zircaloy-2 are susceptible to pitting corrosion in oxidizing chloride solutions. In this study, we investigated the effect of oxide film formed during -ray irradiation on pitting corrosion of fuel cladding in water containing sea salt. The pitting potentials of Zircaloy-2 were measured using the water containing artificial sea salt. Changes in the composition of water containing sea salt were analyzed before and after -ray irradiation. The characteristics of the oxide films formed on Zircaloy-2 were evaluated by X-ray photoelectron spectroscopy. Solution analyses for water containing sea salt showed that hydrogen peroxide was generated by the irradiation. The pitting potential of Ziracloy-2 with oxide film formed under -ray irradiation was higher than that with oxide film formed without irradiation. X-ray photoelectron spectroscopy indicated that the oxide film was composed of zirconium oxide and the growth of oxide film was enhanced during the irradiation. It could thus be explained that the enhanced growth of oxide film under -ray irradiation caused the higher pitting potential.
Kawamura, Hiroshi; *; *; *;
Journal of the Ceramic Society of Japan, International Edition, 97, p.1403 - 1408, 1989/00
no abstracts in English
; Tachikawa, Enzo; ; ; ; ; ; Yamamoto, Katsumune
Journal of Nuclear Science and Technology, 23(6), p.511 - 521, 1986/00
Times Cited Count：4 Percentile：48.02(Nuclear Science & Technology)no abstracts in English
; *;
JAERI-M 82-155, 28 Pages, 1982/11
no abstracts in English
JAERI-M 7349, 28 Pages, 1977/10
no abstracts in English
Ioka, Ikuo; Kato, Hitoshi; Ogawa, Hiroaki
no journal, ,
When the function of cooling system for a spent fuel pool is lost, the spent fuel pin is exposed to steam and air environment. In addition, oxidation behavior of the cladding may be changed due to axial temperature gradient and induced stress gradient during the process of dry out of the spent fuel pool. The oxidation behavior of the Zircaloy-2 cladding under axial temperature gradient was investigated in this study. The axial temperature gradients was about 100 
C/cm. The oxidation test was carried out at 600 C in the saturated steam flow with Ar of 0.5 l/min as a career gas. Little difference was seen in the configuration of the surface cracks and the oxide thickness of specimens oxidized with different temperature gradients. Consequently, the high-temperature oxidation of Zircaloy-2 cladding was hardly changed by the steep axial temperature gradient of about 100 C/cm in this study.
Shizukawa, Yuta; Sekio, Yoshihiro; Yamagata, Ichiro; Akasaka, Naoaki; Maeda, Koji
no journal, ,
The upper plug of a fuel rod used for the spent fuel pool (SFP) of Unit 4 in Fukushima Daiichi Nuclear Plant (1F) consists of Zircaloy-2 bolt and SUS304L nut, and it forms the screw-gap structure. Therefore, when seawater left in this gap structure by a seawater injection after an accident, crevice corrosion might occur. It may have an influence on the integrity of a fuel assembly stored in common pool. In this study, the systematic determining the repassivation potential for crevice corrosion of SUS304L and Zircaloy-2 samples. From the result, SUS304L/SUS304L clearance specimen indicates the tendency of corrosion to progress at 50C, Chloride ion concentration 10 ppm. SUS304L/SUS304L clearance specimen and SUS304L/Zircaloy-2 clearance specimen show the tendency not to corrode at 80C, Chloride ion concentration 10 ppm.
