英高鎳600於高溫水質環境中，具備優良的抗蝕性及良好的機械性質，因此廣泛地被使用於核能電廠高溫高壓環境做為壓力界面組件及反應器結構組件。然而，經過多年的實際運轉，在各種核能電廠水質環境下，皆被發現會產生晶間應力腐蝕被裂的現象。本文闡述各種核能電廠高溫水質環境的電化學意義，包括沸水式反應器正常水質化學(BWR NWC)及加氫水化學(BWRHWC)，壓水式反應器一次側及二次側水質，及其他可能發生之異常水質環境電化學意義的不同，並討論英高鎳600在各種不同核能環境產生晶間應力腐蝕破裂機構的異同，尤其對迄今仍不明確之PWSCC機構做現象性分析。核能電廠各種電化學環境，電位是決定英高鎳600或不銹鋼產生晶間應力腐蝕破裂機構的關鍵，比pH扮演著更重要的角色。

Alloy 600 has been broadly used as pressure boundary and vessel structure component materials in high temperature high pressure nuclear reactor power plants, since it possesses excellent corrosion resistant and good mechanical properties in high temperature aerated or deaerated water and steam. However, in practical operations, alloy 600 has been observed to suffer Intergranular Stress Corrosion Cracking in many nuclear water environments which causes great industry concern. This paper illustrates the electrochemical meaning in pH-Ediagram for various nuclear reactor water environments, including Boiling Water Reactor Normal Water Chemistry (BWR NWC), BWR Hydrogen Water Chemistry (BWR HWC), Pressurized Water Reactor Primary and Secondary side water chemistries and other abnormal water chemistries when resin or other impurities intrusion occur. The mechanism of intergranular stress corrosion cracking of alloy 600 in various nuclear reactor water systems will be discussed, especially the phenomenological analysis of PWSCC mechanism which, so far, is still unambiguously clear. Potential, which is more important than pH, plays the dominant role in determining cracking mechanism of alloy 600 or stainless steel in various electrochemical environments of nuclear reactor water systems.