CZ alloys were developed by China General Nuclear Power Group (CGN) for pressurized water reactor. To improve the corrosion resistance, the impacts of annealing on corrosion of CZ1 alloy were studied. The fuel cladding tubes of CZ1 alloy was fabricated by different annealing processes, with intermediate annealing in the range of 560 °C to 700 °C. Then the corrosion properties of CZ1 alloy and low Tin Zr-4 specimens were investigated by autoclave tests. The matrix and second phase particles along with the oxide characteristics were analyzed. The experimental results showed that the corrosion resistance of CZ1 alloy is superior to that of low Tin Zr-4,. Besides, Lower corrosion weight gain is obtained in the cladding tubes with decreasing annealing temperature, which correlates well with the smaller size of second phase particles. The underlying mechanism was discussed. The results in this study will be used as a guide for the material processing design of CZ1 alloy.

Volume Subject Area:
Nuclear Fuel and Material, Reactor Physics and Transport Theory
Topics:
Alloys, Annealing, Corrosion, Cladding systems (Building), Corrosion resistance, Particulate matter, Zirconium, China, Design, Fuels, Materials processing, Nuclear power, Pressurized water reactors, Temperature, Weight (Mass)
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