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Journal of Chinese Society for Corrosion and protection  2015, Vol. 35 Issue (6): 479-487    DOI: 10.11902/1005.4537.2015.024
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Review of Irradiation Assisted Stress Corrosion Cracking of Core Structural Materials
Ping DENG1,Chen SUN2,Qunjia PENG1(),En-Hou HAN1,Wei KE1
1 Key Laboratory of Nuclear Materials and Safety Assessment, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016, China
2 State Nuclear Power Research Institute, Beijing 102209, China
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Abstract  

The current status of research on irradiation assisted stress corrosion cracking (IASCC) of core structural materials for light water reactors was reviewed with focuses on influencing factors and mechanism of IASCC. Challenges and perspectives for the research of IASCC in the future were also briefly addressed.
Key words:  light water reactor      core structural material      irradiation damage      IASCC     
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Ping DENG
Chen SUN
Qunjia PENG
En-Hou HAN
Wei KE

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Ping DENG,Chen SUN,Qunjia PENG,En-Hou HAN,Wei KE. Review of Irradiation Assisted Stress Corrosion Cracking of Core Structural Materials. Journal of Chinese Society for Corrosion and protection, 2015, 35(6): 479-487.

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https://www.jcscp.org/EN/10.11902/1005.4537.2015.024     OR     https://www.jcscp.org/EN/Y2015/V35/I6/479

Fig.1  Dose dependence of radiation induced segregation for several 300-series austenitic stainless steels irradiated at a temperature of about 300 ℃[7]
Fig.2  Summary of reported defect structures in 300-series austenitic stainless steels as a function of irradiation dose and temperature[7,20]
Fig.3  Percentage of IGSCC vs DO for 304 stainless steel[69,70]
Fig.4  Schematic drawing showing how localized deformation by irradiation enhances IASCC[7]
Fig.5  Schematic drawing showing how deformation at the grain boundary leads to rupture of the oxide film[7,76,78]
Fig.6  Schematic drawing showing how expanded cha-nnels enhance a grain boundary slipping[7,76]
Fig.7  Schematic drawing showing how steps are formed by local strain[7,76,80]
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