Studies On Corrosion And Radiation Damage In Materials In Nuclear Environment

The effect of the surface conditions, i.e.,1200-grit ground surface and electro-polished surface, on the corrosion behavior of alloy 600 was studied in this paper. Supercritical water and subcritical water were used in this study to simulate the corrosion environment in the pressurized water reactor (PWR). The oxides formed in the nickel based alloys 600 and 690 were studied systematically with respect to the morphology, the structure and the chemical composition, using scanning electron microscopy (SEM), transmission electron microscopy (TEM) combined with energy dispersive X-ray spectrum (EDS). The gravimetry results showed that the ground samples exhibited higher weight gain and therefore more susceptible to oxidation. It was found that there was a layer of recrystallization under the surface for ground sample and the internal oxidation and corrosion along grain boundary in this layer were dominant in this layer, indicating different corrosion behavior occurred on alloy 600 with different surface conditions. The corrosion behavior of alloys 600 and 690 showed similar mechanism in both supercritical and subcritical water. Similar oxides morphology, structure and composition were observed. Both alloys exhibited dual-layered oxides structure with outer layer composed of granular NiFe2O4 and NiO, while the inner layer of comparatively fine oxides. The discrepant content of Cr between the two alloys gave rise to the different components of the inner layer. The inner oxide layer in alloy 600 was composed of NiO, NiFe2O4 and Cr2O3, and there was no monolithic Cr2O3 layer formed; while for alloy 690, it was composed of Cr2O3 and spinels containing Cr. The diffusion of oxygen in spinels and Cr2O3 was rather slow and will retard the corrosion rate in the following corrosion, thus providing an explaination for superior corrosion resistance for alloy 690 than alloy 600.Regarding the radiation damage, high voltage electron microscope (HVEM) was used to study the nature and bias value of dislocation loops formed in hydrogen ion irradiated pure iron at ambient temperature. It’s demonstrated that both interstitial and vacancy loops formed simultaneously in samples when annealing at temperatures from 400℃ to 490℃. Vacancy loops started to form at 400℃ and accounted for an increasing proportion with increasing annealing temperature. The bias factor for dislocation loops was demonstrated by the size change of the dislocation loops. The bias factor for interstitial loops was higher than that for vacancy loops.Besides, the microstructure of the vanadium irradiated by hydrogen ion at ambient temperature was completely investigated by TEM. High density of small clusters (dislocation loops) formed after hydrogen ion irradiation. Almost all the dislocation loops were characterized with Burgers vector of a/2<111> type and interstitial in nature, with less than 10% of a/2<110> type. Neither a<100> type nor vacancy loops were identified here and the existence of vacancy loops in pure vanadium was discussed.Finally, the effect of general grain boundaries and twin boundaries on the cavities distribution was studied in Cu followed by helium ion irradiation at ambient temperature. High density of small cavities formed after irradiation. Cavities-depleted zone formed readily near the general grain boundaries. Meanwhile, high density of cavities was observed within the grain boundary plane. However, for the twin boundaries, no cavities-depleted zone was observed. We concluded that the role of twin boundaries as point defects sinks was far less significant compared with general grain boundaries.