Crack Growth Tests on a Ferritic Reactor Pressure Vessel Steel under the Simultaneous Influence of Simulated BWR Coolant and Irradiation

摘要

Comprehensive work was performed worldwide on stress corrosion cracking assessment of ferritic reactor pressure vessel (RPV) steels of boiling water reactors (BWR) as well as discussions about the environmental effects of BWR reactor water on crack growth rates (CGR). A possible influence should be evaluated from several points of view reflecting the complexity of in-pile environment in a B WR-RP V. The corrosion potential of a low alloy steel can be increased due to the presence of oxygen and other species generated by radiolysis (species like H{sup}+, OH{sup}-, H{sub}2O{sub}2, (HO{sub}2){sup}-), [1]. In this investigation the focus is on oxygen and hydrogen peroxide. The influences of water impurities have been investigated as well as the influences of metallurgical characteristics of low alloy steels (LAS) [2]. It has been shown that there is no sustained technical relevant crack growth up to a stress intensity factor of 60 MPam{sup}(1/2). Only at stress intensity factors higher than 60 MPam{sup}(1/2) stable crack growth may occur [3].