STRESS CORROSION CRACKING BEHAVIOR OF ALLOY 690 IN BWR ENVIRONMENT

摘要

Alloy 690 (UNS N06690) is a high-Cr Nickel-base alloy that contains 27.0-31.0 % Cr, as specified in the ASTM standard. It is reported that alloy 690 is a highly corrosion-resistant material and widely used in pressurized water reactor (PWR) systems as the control rod drive mechanism (CRDM) and the heat-exchange tube for the steam generator (SG). However, there are few reports on SCC and intergranular corrosion resistance of alloy 690 in the boiling water reactor (BWR) environment. So, in this study, to clarify the intergranular corrosion resistance and SCC resistance of alloy 690 in the BWR environment, an intergranular corrosion test and SCC crack growth test were performed. To consider the influence of heat input of welding process, the alloy 690 of MA (mill annealed) and HAZ (heat affected zone) specimen were used for tests. In the simulated BWR water environment, the alloy 690 (MA) specimen shows a crack growth rate of less than 10~(-12)m/s at stress intensity factor (K) = 21.4 MPa√m. In the other cases (MA (K = 32.9 MPa√m) and HAZ (K = 21.2 MPa√m and 34.5 MPa√m)), the crack growth rates were in the order of 10~(-12) m/s. Thus, it is considered that alloy 690 shows a very low crack growth rate in the BWR environment. Furthermore, intergranular corrosion was not observed in the alloy 690 (MA) and alloy 690 (HAZ). In addition, there was no significant difference in intergranular corrosion rate between alloy 690 (MA) and alloy 690 (HAZ). The intergranular corrosion rate of alloy 690 (MA) and alloy 690 (HAZ) was 7.9 g/m~2/h.