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http://dx.doi.org/10.14773/cst.2021.20.4.210

Review of Factors Affecting IASCC Initiation of Stainless Steel in PWRs  

Hwang, Seong Sik (Korea Atomic Energy research Institute, Material Safety Technology Development Division)
Choi, Min Jae (Korea Atomic Energy research Institute, Material Safety Technology Development Division)
Kim, Sung Woo (Korea Atomic Energy research Institute, Material Safety Technology Development Division)
Kim, Dong Jin (Korea Atomic Energy research Institute, Material Safety Technology Development Division)
Publication Information
Corrosion Science and Technology / v.20, no.4, 2021 , pp. 210-229 More about this Journal
Abstract
To safely operate domestic nuclear power plants approaching the end of their design life, the material degradation management strategy of the components is important. Among studies conducted to improve the soundness of nuclear reactor components, research methods for understanding the degradation of reactor internals and preparing management strategies were surveyed. Since the IGSCC (Intergranular Stress Corrosion Cracking) initiation and propagation process is associated with metal dissolution at the crack tip, crack initiation sensitivity was decreased in the hydrogenated water with decreased crack sensitivity but occurrence of small surface cracks increased. A stress of 50 to 55% of the yield strength of the irradiated materials was required to cause IASCC (Irradiation Assisted Stress Corrosion Cracking) failure at the end of the reactor operating life. In the threshold-stress analysis, IASCC cracks were not expected to occur until the end of life at a stress of less than 62% of the investigated yield strength, and the IASCC critical dose was determined to be 4 dpa (Displacement Per Atom). The stainless steel surface oxide was composed of an internal Cr-rich spinel oxide and an external Fe and Ni-rich oxide, regardless of the dose and applied strain level.
Keywords
Nuclear power plant; Stainless steels; IASCC initiation; Dissolved hydrogen; Threshold fluence;
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