Nuclear Plant Operation and Maintenance Code has been developed and is going to be applied for nuclear power system components in Japan. If a crack is detected in a component, the evaluation of crack growth due to stress corrosion cracking (SCC) is required. In recent years, the components in BWR primary systems made of low carbon stainless steel, such as core shroud and PLR piping, have suffered from SCC and it is necessary to prepare the crack growth rate reference curves for the materials. In this paper, the development of the SCC growth rate database for low carbon stainless steel in BWR water and the proposed reference curves in Japan are described.

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