Zircaloy-4, which is used widely as a core structural material in pressurized water reactors (PWRs), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and zirconium hydride phases precipitate after the Zircaloy-4 lattice becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4, degrading its mechanical performance as a structural material. Because hydrogen can move rapidly through the Zircaloy-4 lattice, the potential exists for large concentrations of hydride to accumulate in local regions of a Zircaloy component remote from its point of entry into the component. Much has been reported in the literature regarding the long-range migration of hydrogen through Zircaloy under concentration gradients and temperature gradients. Relatively little has been reported, however, regarding the long-range migration of hydrogen under stress gradients. This paper presents experimental results regarding the long-range migration of hydrogen through Zircaloy in response to both tensile and compressive stress gradients. The importance of this driving force for hydrogen migration relative to concentration and thermal gradients is discussed.

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A study of the oxide layers formed in 360 °C and 316 °C water on Zircaloy-4 samples has been performed in an attempt to help answer fundamental questions about oxide protectiveness, growth mechanisms, and the nature of oxide growth during autoclave corrosion. Two different oxide thicknesses-12 and 39.5 microns-were investigated. Microbeam synchrotron radiation diffraction and fluorescence techniques with an x-ray beam size of 0.2 microns were used to characterize oxide in cross sections to determine the oxide phase content, grain size, texture, and orientation relationships as a function of through-thickness from the oxide-metal interface. The results confirm that the oxide is comprised primarily of monoclinic ZrO 2 , with tetragonal ZrO 2 present in small amounts. The observed diffraction peaks are consistent with monoclinic phases having a strong fiber texture with the 200 m plane aligned with the oxide-metal interface, and with the 011 m plane closely aligned with the transversenormal (T) plane. The fraction of bulk tetragonal phase increased in the region located within one transition thickness near the oxide-metal interface. A strong periodicity was seen in oxide intensity from both the monoclinic and tetragonal phases corresponding to an oxide transition thickness of 1.8 to 1.9 microns. The grain size of the tetragonal phase was determined to be smaller than the monoclinic phase, and the grain size for the monoclinic phase decreased starting at a distance of approximately one transition layer from the oxide-metal interface. The relative amounts of monoclinic peak broadening due to strain and grain size were calculated, the former being approximately constant, while the latter decreased with increasing distance from the oxide-metal interface, corresponding to an increase in grain size. These findings are compared to previously performed microbeam diffraction experiments on Zircaloy-4 and other Zr-based alloys.

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Long-time Advanced Test Reactor (ATR) corrosion weight-gain data from nonfueled α-processed Zircaloy-4 cladding coupons are presented herein. The data were generated in pressurized loops that simulated pressurized water reactor (PWR) chemistry, temperature, and neutron flux conditions. High neutron flux and high-fluence post-transition corrosion rates were accelerated over out-of-flux post-transition corrosion rates by more than 40 times at 270°C and more than 30 times at 310°C. We conclude in large part that the acceleration was caused by irradiation-induced changes to the base metal. The effect that irradiation has on the base metal and its effect on corrosion has been shown in the past to be carried over into post-irradiation tests in autoclave environments, but its impact was shown in this study to be accentuated in the in-reactor environment, in which sustained corrosion accelerations were much greater than observed in the postirradiation tests (in nonirradiation environments). This suggests some aspect of the radiation environment potentially acts synergistically with the corrosion film formed from the irradiation-damaged base metal to further accelerate in-reactor corrosion rates. We also present results from detailed microstructural examinations of thick (more than 25 µm) oxide films formed on α-processed Zircaloy-4 grown in a 360°C autoclave and in the reactor. The examinations consisted of optical microscopy, field emission gun scanning electron microscopy, focused ion beam microscopy, analytical electron microscopy, and X-ray diffraction, with a focus on the evolution of the microstructure of the intermetallic precipitates in the oxide films during exposure to autoclave water or to the corroding environment under neutron irradiation. The film microstructure suggests that the water environment was closer to the oxide-metal interface in the films grown in the reactor than in the autoclave-grown corrosion films examined, but it was not obvious from the microstructural examinations how the radiation environment facilitated that effect.

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