Correlation Between Characteristics of Oxide Films Formed on Zr Alloys in BWRs and Corrosion Performance

Nanikawa, S; Etoh, Yoshinori; Shimada, Satoshi; Kubo, Takashi; Ito, Kazuyuki; Harada, Hideto

doi:10.1520/stp14329s

Cited by 13 publications

(2 citation statements)

References 22 publications

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“…Corrosion of Zr-2.5Nb in low and high oxygen water at 570 K [43]. Nb-containing alloys was explained by the low electric resistance of the oxide on Nb-containing alloys [44].…”

Section: Effect Of Coolant Conditionmentioning

confidence: 99%

Irradiation Effects on ZR-2.5NB in Power Reactors

Song

2016

CNL Nuclear Review

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Zirconium alloys are widely used as structural materials in nuclear applications because of their attractive properties such as a low absorption cross-section for thermal neutrons, excellent corrosion resistance in water, and good mechanical properties at reactor operating temperatures. Zr-2.5Nb is one of the most commonly used zirconium alloys and has been used for pressure tube materials in CANDU (Canada Deuterium Uranium) and RBMK (Reaktor Bolshoy Moshchnosti Kanalnyy, “High Power Channel-type Reactor”) reactors for over 40 years. In a recent report from the Electric Power Research Institute, Zr-2.5Nb was identified as one of the candidate materials for use in normal structural applications in light-water reactors owing to its increased resistance to irradiation-induced degradation as compared with currently used materials. Historically, the largest program of in-reactor tests on zirconium alloys was performed by Atomic Energy of Canada Limited. Over many years of in-reactor testing and CANDU operating experience with Zr-2.5Nb, extensive research has been conducted on the irradiation effects on its microstructures, mechanical properties, deformation behaviours, fracture toughness, delayed hydride cracking, and corrosion. Most of the results on Zr-2.5Nb obtained from CANDU experience could be used to predict the material performance under light water reactors. This paper reviews the irradiation effects on Zr-2.5Nb in power reactors (including heavy-water and light-water reactors) and summarizes the current state of knowledge.

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“…Corrosion of Zr-2.5Nb in low and high oxygen water at 570 K [43]. Nb-containing alloys was explained by the low electric resistance of the oxide on Nb-containing alloys [44].…”

Section: Effect Of Coolant Conditionmentioning

confidence: 99%

Irradiation Effects on ZR-2.5NB in Power Reactors

Song

2016

CNL Nuclear Review

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“…Details about the measurement system and cell have been reported previously [12,13]. The EIS spectra were obtained at the open-circuit potential (OCP) of the working electrode in 25 o C 0.05M H2SO4 solution, with an amplitude of 10 mV.…”

Section: Electrochemical Impedance Spectroscopy (Eis) Measurement Of mentioning

confidence: 99%

Correlation between the oxide impedance and corrosion behavior of Zr−Nb−Sn−Fe−Cu alloys

et al. 2004

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The correlation between the oxide impedance and corrosion behavior of two series of Zr-Nb-Sn-Fe-Cu alloys was evaluated. Corrosion tests were performed in a 70 ppm LiOH aqueous solution at 360 o C for 300 days. The results of the corrosion tests revealed that the corrosion behavior of the alloys depended on the Nb and Sn content. The impedance characteristics for the pre-and post-transition oxide layers formed on the surface of the alloys were investigated in sulfuric acid at room temperature. From the results, a pertinent equivalent circuit model was preferably established, explaining the properties of double oxide layers. The impedance of the oxide layers correlated with the corrosion behavior; better corrosion resistance always showed higher electric resistance for the inner layers. It is thus concluded that a pertinent equivalent circuit model would be useful for evaluating the long-term corrosion behavior of Zr-Nb-Sn-Fe-Cu alloys.

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Direct observation of hydrogen and deuterium in oxide grain boundaries in corroded Zirconium alloys

et al. 2015

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Atom probe tomography has been used to the study the distribution of hydrogen and deuterium in the oxide scale of two common zirconium alloys after autoclave testing in H2O and D2O, respectively. Comparison between hydrogen and deuterium in the mass spectra allows for separation of hydrogen as a corrosion product from adsorbed H2 gas from the vacuum chamber. Enrichment of hydrogen and deuterium, as OH + and OD + , was observed in grain boundaries. The grain boundaries were identified through segregation of iron. This lends experimental support to existing theories for the mechanism of hydrogen pickup in zirconium alloys.

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Correlation Between Characteristics of Oxide Films Formed on Zr Alloys in BWRs and Corrosion Performance

Cited by 13 publications

References 22 publications

Irradiation Effects on ZR-2.5NB in Power Reactors

Irradiation Effects on ZR-2.5NB in Power Reactors

Correlation between the oxide impedance and corrosion behavior of Zr−Nb−Sn−Fe−Cu alloys

Direct observation of hydrogen and deuterium in oxide grain boundaries in corroded Zirconium alloys

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