Comparison of the Long-Time Corrosion Behavior of Certain Zr Alloys in PWR, BWR, and Laboratory Tests

Abstract

Laboratory corrosion tests have always been an important tool for Zr alloy development and optimization. However, it must be known whether a test is representative for the application in-reactor. To shed more light on this question, coupons of several Zr alloys were exposed under isothermal conditions in all or most of the following environments: In-Reactor: (1) PWR core at 300 to 340°C up to six years. (2) BWR core with a low sensitivity to nodular corrosion up to four years. (3) BWR core with a high sensitivity to nodular corrosion up to two years. Ex-Reactor (in Autoclave): (1) 350°C/pressurized water up to three years. (2) 400°C/100-bar steam up to two years. (3) 350°C/0.01 M LiOH water up to two years. (4) 500 to 515°C/high-pressure steam 16 to 24 h. In addition, the material condition of several of the examined Zr alloys was varied over a wide range. For evaluation of the in-PWR tests and for comparison of out-of-pile and in-pile tests, the different temperatures and times were normalized to a temperature-independent “normalized time” by assuming an activation temperature (Q/R) of 14 200 K. Comparison of in-PWR and out-of-pile corrosion behavior of Zircaloy shows that corrosion deviates to higher values in PWR if a weight gain of about 50 mg/dm2 is exceeded. In the case of the Zr2.5Nb alloy, a slight deviation of corrosion as compared to laboratory results starts in PWR only above a weight gain of 100 mg/dm2. In BWR, corrosion of Zircaloy is enhanced early in time if compared with out-of-pile. Zr2.5Nb exhibits higher corrosion results in BWR than Zircaloy-4. Alloying chemistry and material condition affect corrosion of Zr alloys. However, several of the material parameters have shown a different ranking in the different environments. Nevertheless, several material parameters influencing in-reactor corrosion like the second phase particle (SPP) size or in-PWR behavior as the Sn and Fe content can be optimized by out-of-pile corrosion tests.

Effects of Sn on the oxidation and oxide characteristics of Zr alloys in water and a LiOH solution

“…Takeda [11] reported that an increase of Sn content stabilized the monoclinic ZrO2 in the oxide, and thus the corrosion resistance decreased in high Sn content alloys. It has also been reported [12] that the effect of Sn on the corrosion is different depending on the corrosion environment; a low corrosion rate of high Sn content alloys was observed in simulated BWR conditions and LiOH solution.…”

Investigating the Effect of Zirconium Oxide Microstructure on Corrosion Performance: A Comparison between Neutron, Proton, and Nonirradiated Oxides

“…The extensive transformation from tetragonal to monoclinic oxide during growth leads to disruption of the protective oxide and the development of extensive networks of transformation twin boundaries [7], [10], [16], [17]. However, as the corrosion behaviour under irradiation differs from that observed during autoclave exposure [18], many of the mechanisms derived from observations on non-irradiated oxides may no longer be as relevant in the complex reactor environment. As the industry is pushing towards extended burn up, where the corrosion behaviour of the cladding deviates further from out-of-pile tests [19], [20], it is vitally important that the underlying mechanisms of corrosion under irradiation are understood.…”

Influence of light ion irradiation of the oxide layer on the oxidation rate of Zircaloy-4

“…A second assumption assigned the accelerated corrosion to the amorphisation and the dissolution of the Zr(Fe,Cr) 2 precipitates under irradiation [2,[8][9][10][11][12]. Others authors suggest that the kinetic acceleration is due to the tin distribution in the alloy in service [13][14][15]. Indeed, the reduction of tin content in Zircaloy-4 seems to be correlated with a lowering of the corrosion rate of the alloy in PWRs.…”