Effects of surface grinding for scratched alloy 690TT tube in PWR nuclear power plant: Microstructure and stress corrosion cracking

2022 ◽  
Author(s):  
Bin Wu ◽  
Hongliang Ming ◽  
Fanjiang Meng ◽  
Yifeng Li ◽  
Guangqing He ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Corrosion Cracking ◽  
Surface Grinding

scholarly journals Factors Involved in Stress Corrosion Cracking of Tubes from a Nuclear Power Plant Feedwater Heater

2019 ◽  
Vol 70 (11) ◽  
pp. 3926-3930
Author(s):  
Manuela Fulger ◽  
Dumitra Lucan ◽  
Maria Mihalache ◽  
Lucian Velciu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
High Water ◽  
Corrosion Cracking ◽  
304 Stainless Steel ◽  
Final Conclusion ◽  
X Ray

This study has been carried out to identify the operating factors involved in the failure of two tubes made of 304 stainless steel removed from a high pressure feedwater heater working in a nuclear power plant. The samples cut from tubes have been analyzed by different methods: visual examination, optical (metallographic) microscopy, scanning electron microscopy, and energy dispersive X-ray spectroscopy. Following analyses, on the surface of tubes was highlighted the presence of many pits in which cracks have started. Inside the pits, impurities of the type S, Cl, K, Ca were detected too. The branched cracks most likely have occurred as a result of precipitation of small amounts of chlorides deposited and concentrated on the surface of the tubes over a long period of operation. The stresses that favored this type of corrosion cracking were both residual stresses and stresses occurring at the torsion and bending of the tubes, while the high water temperature from feed heater was also a favorable factor for corrosion. The final conclusion was that the tubes failed by a chloride induced stress corrosion cracking mechanism.

scholarly journals Factors Involved in Stress Corrosion Cracking of Tubes from a Nuclear Power Plant Feedwater Heater

2019 ◽  
Vol 70 (11) ◽  
pp. 3926-3930
Author(s):  
Manuela Fulger ◽  
Dumitra Lucan ◽  
Maria Mihalache ◽  
Lucian Velciu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
High Water ◽  
Corrosion Cracking ◽  
304 Stainless Steel ◽  
Final Conclusion ◽  
X Ray

This study has been carried out to identify the operating factors involved in the failure of two tubes made of 304 stainless steel removed from a high pressure feedwater heater working in a nuclear power plant. The samples cut from tubes have been analyzed by different methods: visual examination, optical (metallographic) microscopy, scanning electron microscopy, and energy dispersive X-ray spectroscopy. Following analyses, on the surface of tubes was highlighted the presence of many pits in which cracks have started. Inside the pits, impurities of the type S, Cl, K, Ca were detected too. The branched cracks most likely have occurred as a result of precipitation of small amounts of chlorides deposited and concentrated on the surface of the tubes over a long period of operation. The stresses that favored this type of corrosion cracking were both residual stresses and stresses occurring at the torsion and bending of the tubes, while the high water temperature from feed heater was also a favorable factor for corrosion. The final conclusion was that the tubes failed by a chloride induced stress corrosion cracking mechanism.

STRESS CORROSION CRACKING RESISTANCE OF CF8A STEEL USED IN THE NUCLEAR POWER PLANT

2008 ◽  
Vol 22 (11) ◽  
pp. 1031-1037
Author(s):  
DONG HO BAE ◽  
GYU YOUNG LEE ◽  
KWANG JIN PARK ◽  
SUN KYU PARK
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Corrosion Cracking ◽  
Solution Temperature ◽  
Pipe Material ◽  
Primary Stage ◽  
Degraded Material

In this investigation, stress corrosion cracking characteristics due to pipe material degradation in the primary stage environment of the nuclear power plant was evaluated. Firstly, after artificially degrading the CF8A steel during 2, 4, and 6 months at elevated temperature, 400°C, assessed the corrosion susceptibility of the degraded material following ASTM G5 standard. And next, stress corrosion cracking (S.C.C.) tests using smooth specimens fabricated with degraded material were conducted in the 2wt.% H 2 BO 3+ Li 7 OH solution at 60°C, and static stress of 0.8 σ y (σ y : yield strength of artificially degraded material). From the results, corrosion rates linearly increased with increasing of degradation period and solution temperature. And both the raw material and the degraded materials were not failed during the S.C.C. test period. In spite of long time test (about 3,900 hrs) under S.C.C. test condition, surface pits or surface corrosion by the electro chemical reaction were not observed. Consequently, CF8A steel using as one of the pipe material in the primary stage of the nuclear power plant showed good S.C.C. resistance in 2wt.% H 2 BO 3+ Li 7 OH solution.

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