Abstract
A comprehensive model is proposed to predict the plateau corrosion fatigue crack growth rate, (CFCGR) d a/d N p, and the threshold stress intensity factor range, Δ K pth, for reactor pressure vessel (RPV) steels which display stress corrosion-type corrosion fatigue in high temperature water. The controlling factors have been evaluated and their interactive effect on the prediction has been established. The modelling was based on the experimental data obtained from A533B Class 1 steel plate in stagnant PWR primary water for a range of steel sulphur contents, temperature and corrosion potential values. The mechanistic background of the model is supported by fractographic observations, being consistent with a crack tip environment controlled process related to sulphur chemistry. The model is presented by mathematical descriptions. The applicability of this model in the assessment of critical corrosion fatigue situations for the safe operation of nuclear power plants is discussed.
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