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首次应用控制电位慢应变速率试验方法(SSRT)研究了600MW核电站蒸汽发生器(SG)传热管材Inconel-690合金在模拟二次侧水化学限值工况260℃、10~(-6)Cl~-、未除气条件下的应力腐蚀破裂(SCC)行为与外加电化学电位(ECP)之间的关系,确定了产生SCC的临界破裂电位。
(SSRT) was used for the first time to study the effects of the Inconel-690 alloy heat transfer tube (600MW nuclear power plant steam generator) on the simulation of the secondary chemical resistance of 260 ℃, 10 ~ (-6) Cl The relationship between SCC behavior and applied electrochemical potential (ECP) under non-de-aeration conditions was determined. The critical rupture potential of SCC was determined.