Alloy 600 specimens were exposed to a steam-hydrogen environment considered to simulate exposure to primary water conditions in nuclear power plants. Ni nodules were expelled to the surface indicating the likelihood of internal oxidation. A cross-section of the exposed specimen was prepared using a focused ion beam and a composition profile across the cross-section was generated to confirm internal oxidation of Cr. It was revealed that internal intergranular oxidation occurred, and embrittlement was confirmed by the appearance of cracks in the thin section. From these results, it can be concluded that in primary water-type environments Alloy 600 can undergo internal intergranular oxidation of Cr causing embrittlement of grain boundaries.
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机译：合金600标本暴露于被认为模拟核电站初级水条件暴露的蒸汽氢环境。 Ni Nodules被驱逐到表明内部氧化可能性的表面。使用聚焦离子束制备暴露样本的横截面，并产生横截面的组成曲线以确认Cr的内部氧化。揭示了内部晶间氧化发生，并且通过薄截面中的裂缝的出现确认了脆化。从这些结果中，可以得出结论，在初级水型环境中，合金600可以经历内部晶间氧化Cr导致晶界的脆化。
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