Study on the Corrosion Behavior of 690TT Alloy Heat Transfer Tubes with Scratch Defects in a Simulated High-Temperature and High-Pressure Water Environment of the Pressurized Water Reactor Secondary Circuit

In order to evaluate the effect of scratch defects on the stress corrosion performance of 690TT alloy heat transfer tubes in a simulated high-temperature and high-pressure water environment of the secondary circuit， a nuclear power steam generator heat transfer tube was chosen as the research subject.A scratch device was used to create C-ring specimens with scratch defects on the surface of the tube.The stress corrosion behavior of the scratched 690TT alloy heat transfer tubes was investigated in the simulated secondary circuit high-temperature and highpressure water environment.The surface defect morphology， oxidation morphology， distribution of oxide elements and surface profile of the specimens before and after corrosion were analyzed using scanning electron microscopy (SEM)， transmission electron microscopy (TEM)， energy dispersive spectrometer (EDS)，Raman spectroscopy and white light interferometer.Results showed that in the simulated secondary circuit high-temperature and high-pressure water environment， spinel oxide and chromium-rich oxide were formed on the surface of the C-ring scratch groove under stress loading.Additionally， the particle size and quantity of the spinel oxide on the outer surface increased with the test duration.The scratching process led to the formation of mechanical microcracks in the scratch groove area， but after long-term immersion， these microcracks did not show significant propagation， indicating that the scratched 690TT alloy has a high resistance to stress corrosion cracking in the secondary circuit simulated environment.