Development of stress corrosion susceptibility assessment methodology at nuclear reactor environment

This paper presents the stress corrosion cracking (SCC) primary tests, performed to develop a SCC susceptibility assessment methodology of metallic alloys, at nuclear reactor ambients such as PWR and BWR. The behavior of stainless steel AISI 321, under BWR conditions and alloy 600 MA, under PWR conditions, were evaluated. Slow strain rate test, a SCC susceptibility comparative method, employable for various materials and ambients, was used. For stainless steel 321 the test parameters were: demineralized water; temperature: 288 deg C; pressure: 8 MPa; strain rate: 3x10-7 s-1. For alloy 600 MA the ambient conditions were: water containing 1200 ppm of boron; 2.2 ppm of lithium; 35 cm3 H2/kg H2O; 5 ppb of O2; temperature: 303 deg C; pressure: 10 MPa; strain rate: 3x10-7 s-1. A second test was performed with the alloy 600 MA, using N2, pressure: 0.2 MPa, under the same temperature and strain rate. The results were obtained and compared as stress deformation curves, showing the SCC behavior in all ambients. Additionally, the fracture surface specimens were analyzed by scanning electron microscopy - SEM. (author)