Cracking of a nuclear waste container material by irradiation in a simulated groundwater

Abstract

Fatigue-crack propagation tests were conducted on a candidate container material (ASTM A27 steel) tested in Hanford groundwater at 150°C for application in a potential basalt repository. Tests were run at a single value of stress intensity factor on groups of identical specimens undergoing gamma irradiation and control specimens not exposed to irradiation. The gamma flux levels (approx. 123 rad/hour) were prototypic of the maximum levels expected at the outer surface of the waste container. A statistical evaluation suggested that there were no significant differences between crack growth rates in the unirradiated and irradiated specimens.