Cyclic cracking behavior of low-alloy pressure vessel steel in simulated BWR water
Author:
Publisher
Elsevier BV
Subject
Nuclear Energy and Engineering,General Materials Science,Nuclear and High Energy Physics
Reference27 articles.
1. Corrosion behaviour of reactor coolant system materials in nuclear power plants
2. On the mechanisms of environment sensitive cyclic crack growth of nuclear reactor pressure vessel steels
3. The effect of temperature on fatigue crack growth behaviour of a low alloy pressure vessel steel in a simulated BWR environment
4. Status of Research on Environmentally Assisted Cracking in LWR Pressure Vessel Steels
5. Fatigue strength correction factors for carbon and low-alloy steels in oxygen-containing high-temperature water
Cited by 3 articles. 订阅此论文施引文献 订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献
1. Effects of dynamic strain aging on mechanical properties of SA508 class 3 reactor pressure vessel steel;Journal of Materials Science;2009-06
2. Influence of hydrogen on the toughness of irradiated reactor pressure vessel steels;Journal of Nuclear Materials;2006-12
3. Effect of dynamic strain aging on fatigue crack growth behaviour of reactor pressure vessel steels;Materials Science and Technology;2006-08
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