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Journal of Chinese Society for Corrosion and protection  2021, Vol. 41 Issue (4): 417-428    DOI: 10.11902/1005.4537.2020.101
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Research Progress on Stress Corrosion Cracking of Stainless Steel for Nuclear Power Plant in High-temperature and High-pressure Water
JIAO Yang, ZHANG Shenghan(), TAN Yu
Department of Environment Science and Engineering, North China Electric Power University, Baoding 071003, China
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Abstract  

Stress corrosion cracking behavior (SCC) of typical nuclear structural materials stainless steel in high-temperature and high-pressure water was reviewed. The effect of material type, mechanical properties, water chemistry on the sensitivity, initiation and propagation of cracks were discussed. The development of the zinc injection technology in the coolant system and its inhibitory effect on the primary side stress corrosion (PWSCC) were introduced. Finally, by taking various factors and actual operating conditions into consideration, the method of reducing the sensitivity of SCC was summarized, and the problems that should be focused on in future research were proposed.

Key words:  nuclear power plant      stainless steel      SCC      zing injection technology     
Received:  12 June 2020     
ZTFLH:  TG174  
Corresponding Authors:  ZHANG Shenghan     E-mail:  shenghan_zhang@126.com
About author:  ZHANG Shenghan, E-mail: shenghan_zhang@126.com

Cite this article: 

JIAO Yang, ZHANG Shenghan, TAN Yu. Research Progress on Stress Corrosion Cracking of Stainless Steel for Nuclear Power Plant in High-temperature and High-pressure Water. Journal of Chinese Society for Corrosion and protection, 2021, 41(4): 417-428.

URL: 

https://www.jcscp.org/EN/10.11902/1005.4537.2020.101     OR     https://www.jcscp.org/EN/Y2021/V41/I4/417

Fig.1  Effects of yield strength on crack growth rate of stainless steels[15,29]
Fig.2  Variations of crack growth rates of several stainless steels in oxygenated pure water at high temperature with stress intensity factor[32,33]
Fig.3  Effects of temperature, heat treatment and surface state on stress corrosion cracking of 304 stainless steel[44]
Fig.4  Effects of pH on SCC of 304L, 316L and France Z6CND17.12 austenitic stainless steel[47-49]
ParameterAP1000Qinshan nuclear power Plant
Conductivity1~40 μS/cm (25 ℃)1~40 μS/cm (25 ℃)
pH4.2~10.5 (25 ℃)5.4~10.5 (25 ℃)
O20.1 mg/L0.1 mg/L
Cl-0.15 mg/L0.1 mg/L
F-0.15 mg/L---
H225~50 cm3 (STP)/kg H2O25~35 cm3 (STP)/kg H2O
Suspended solids0.2 mg/L1.0 mg/L
LiOH---0.22~2.2 mg/L
H3BO30~4000 mg/L (calculated as B)0~2400 mg/L (calculated as B)
Table 1  Water quality requirements for PWR primary water
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