Keywords: Alloy 600, control rod drives, drive mechanism nozzles, modelling, pressurised water reactors, PWR, primary water stress corrosion cracking, PWSCC, nuclear power plants, nuclear energy, crack propagation, strain rate, Brazil
Preliminary results on modelling of primary water stress corrosion cracking at control rod drive mechanism nozzles of PWR nuclear plants
One of the main deterioration modes that cause risks to pressurised water reactors is the Primary Water Stress Corrosion Cracking (PWSCC) at the Control Rod Drive Mechanism (CRDM) nozzles in the reactor pressure vessel. These cracks can cause accidents that reduce nuclear safety, and/or leakage of primary water. In this paper, preliminary modelling to predict these failures is proposed. The potential-pH diagram for Alloy 600 on primary water at high temperature is assumed. Over it is marked the region where the PWSCC cracks can initiate and propagate. Later, a comparative model is superimposed based on strength fraction to PWSCC, a strain rate damage model and a semi-empirical one that can describe the time of failure. Some preliminary results are presented and discussed. These models are adequate for using experimental data to be obtained from Slow Strain Rate Testing (SSRT) at the CDTN-Development Center of Nuclear Technology, Belo Horizonte, Brazil.