Abstract
The purpose of this work was to examine the effect of tensile stress on the oxide properties of a nickel-based Alloy 600 that was exposed to simulated nuclear steam generator water at 340 °C for 1000 h. The size of the outer oxide particles increased, and the chromium content of the inner oxides decreased under tensile stress. Electrochemical measurements revealed that the charge carrier density increased, and the charge transfer resistance and film resistance were reduced under the tensile stress condition. These changes in the oxide properties are attributed to the formation of short diffusion paths such as line and surface defects due to tensile deformation.
Highlights
Nickel-based Alloy 600 materials have been used extensively to fabricate the major components of pressurized water reactors (PWRs) such as steam generator (SG) tubes, control rod drive mechanisms, and instrument penetrations
The size of the surface particles on the specimen with the tensile stress markedly increased to approximately 2 μm, indicating that the corrosion from the alloy substrate was significantly accelerated by the tensile stress
When comparing the stressed with the stress-free condition, important observations obtained under the tensile condition can be summarized as follows: the size of the outer oxide particles increased remarkably (Figure 4), the chromium content in the inner oxide layer was halved (Figures 5 and 6), the anodic polarization current density increased significantly (Figure 7), the Electrochemical impedance spectroscopy (EIS) resistance of the film was decreased (Table 2), and the defect density in the film was doubled (Table 3)
Summary
Nickel-based Alloy 600 materials have been used extensively to fabricate the major components of pressurized water reactors (PWRs) such as steam generator (SG) tubes, control rod drive mechanisms, and instrument penetrations. Due to the susceptibility of such materials to corrosion damage [1,2,3], in particular, stress corrosion cracking (SCC) in both primary and secondary water environments of PWRs, Alloy 690 materials containing higher chromium content are specified for new plants and for replacement components of operating plants. SCC is the most critical among various modes of damage that occur in SGs. SCC is the most critical among various modes of damage that occur in SGs This is because cracks in SG tubes may cause an abrupt tube rupture accident, resulting in a radioactive coolant leak from the primary to the secondary system. Heat flux from the hotter primary side to the colder secondary side causes compressive hoop stress on the inner surface and tensile hoop stress on the outer surface [7,8,9,10]. SG tubes suffer from mechanical stresses developed by plastic deformation such as U-bending [11,12,13,14], tube expansion at the top of the tubesheet [14,15,16,17], and denting corrosion [1,14,18,19]
Sign-in/Register to view highlights & summary Sign-in/Register to access full text options 
Talk to us
Join us for a 30 min session where you can share your feedback and ask us any queries you have
Schedule a call
